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039492320 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention concerns nuclear oil well-logging in general. More specifically, it relates to an improvement applicable to a high-voltage neutron generator as used in a logging tool. It particularly deals with a detector for sensing and controlling arcing conditions. 2. Description of the Prior Art It has been found in connection with well-logging instruments of the type which use a neutron generator, that the exceedingly high voltage employed tends to create conditions such as to subject the generator to a high-voltage discharge, or breakdown. Furthermore, such conditions are unstable, and the arcs involved are frequent and large in amplitude. Consequently, if such arcs continue without control, the glass envelope of the neutron tube can be punctured. Clearly, such results are troublesome and expensive and may involve considerable loss of valuable equipment. Consequently, it is an object of this invention to provide a means for detecting arcing conditions so that protective steps may be taken to avoid destructive conditions. SUMMARY OF THE INVENTION Briefly, this invention concerns a high-voltage arc detector for protecting against breakdown of well-logging equipment where such equipment has a high-voltage supply and a neutron generator tube connected to said high-voltage supply. The invention comprises in combination a high-voltage connector for carrying said high voltage from said supply to said tube, and insulation surrounding said connector. It also comprises capacitance means associated with said connector for producing a signal when an arc occurs. Again, briefly, the invention concerns a high-voltage arc detector for protecting against breakdown of well-logging equipment, which has a high-voltage supply and a neutron generator tube connected to said voltage supply. It comprises in combination a cylindrical housing for said equipment enclosing said high-voltage supply and said neutron generator tube. It also comprises a section of said housing which encloses an axially located high-voltage connector comprising a metallic cylindrical conductor for carrying said high voltage from said supply to a target in said tube. It also comprises high-voltage insulation filling said housing section surrounding said cylindrical conductor, and a thin relatively narrow electrically conductive strip located in a circumferential groove in said insulation and located longitudinally between said high-voltage supply and said neutron generator tube, and radially spaced from said conductor in order to form an electrical capacitor therewith. It also comrpises a resistor connected between said strip and circuit ground, and an amplifier having an input and an output as well as a relay having a switch for controlling energization of said high-voltage supply. It also comprises first circuit means for connecting said strip to said amplifier input, and second circuit means for connecting said amplifier output to said relay in order to activate said relay whenever an arc occurs so that the high-voltage supply will be deenergized. |
claims | 1. A fastening and loosening device having a plurality of stud bolts screwed into an object to be fastened and a plurality of nuts respectively screwed onto the stud bolts, the device configured to rotate a respective nut, causing tension to act on a respective stud bolt, in a shaft center direction, away from the object to be fastened, to perform fastening or loosening, the fastening and loosening device comprising:a device main body having an upper portion movably supported along an arrangement direction of the stud bolts;a bolt tensioner supported by the device main body and being freely movable along the shaft center direction of the stud bolts;a guide device provided in a lower portion of the device main body and having right and left guide members contactable with an outer periphery of the nut from both right and left sides of a moving direction of the device main body; anda guide position adjustment device capable of moving at least one of the right and left guide members in a horizontal direction intersecting with the moving direction of the device main body. 2. The fastening and loosening device according to claim 1, wherein the arrangement direction of the stud bolts is a circumferential direction along an outer periphery of the object to be fastened, and the guide device allows the guide member positioned at an inner side of the circumferential direction to be movable by the guide position adjustment device. 3. The fastening and loosening device according to claim 2, further comprising a positioning device having a positioning member inserted between adjacent nuts from an outside of the circumferential direction. 4. The fastening and loosening device according to claim 3, further comprising an insertion position adjustment device capable of adjusting an insertion position of the positioning device. 5. The fastening and loosening device according to claim 1, wherein the guide position adjustment device includes a support shaft supporting one end portion of the guide member in a freely rotatable manner, and an eccentric mechanism allowing the other end portion of the guide member to be movable. 6. The fastening and loosening device according to claim 1, wherein the guide device is constructed of the right and left guide members fixed to a lower portion of a box body with a predetermined interval, and the guide position adjustment device is capable of horizontally swinging the box body. 7. The fastening and loosening device according to claim 1, wherein the right and left guide members are provided in front and behind the bolt tensioner in the moving direction of the device main body, and the right and left guide members are contactable with the outer peripheries of two or more of the nuts. 8. A fastening and loosening method for a fastening and loosening device having a plurality of stud bolts screwed into an object to be fastened and a plurality of nuts respectively screwed onto the stud bolts, the device configured to rotate a respective nut, causing tension to act on a respective stud bolt, in a shaft center direction, away from the object to be fastened, to perform fastening or loosening, the method comprising:providing a device main body having an upper portion movably supported along an arrangement direction of the stud bolts;providing a bolt tensioner supported by the device main body and being freely movable along the shaft center direction of the stud bolts; providing a guide device provided in a lower portion of the device main body and having right and left guide members contactable with an outer periphery of the nut from both right and left sides of a moving direction of the device main body; and providing a guide position adjustment device capable of moving at least one of the right and left guide members in a horizontal direction intersecting with the moving direction of the device main body. |
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041586069 | summary | BACKGROUND OF THE INVENTION The present invention relates to stainless steel alloys modified to withstand radiation damage attributable to fast neutrons. In a method sense, the invention to be described provides an alloy design formula wherein existing alloys may be modified or new alloys may be synthesized which exhibit enhanced resistance to density and dimensional changes resulting from exposure to fast neutrons at high temperature, i.e., greater than about 300.degree. C. In still another sense, the invention resides in stainless steel clad nuclear fuel elements intended for use in fast neutron environments wherein the stainless steel cladding is an austenitic stainless steel containing void suppressing concentrations of Si and Ti and the nuclear fuel is an oxide such as UO.sub.2, a nitride such as UN or U.sub.2 N.sub.3, a carbide of uranium (such as UC or UC.sub.2) or mixed with an oxide, nitride, or carbide of Pu or Th. The core components of a thermal or fast nuclear reactor are known to undergo a variety of stresses during their service life. For example, the fuel cladding will experience thermal and mechanical stress due to such factors as fission gas pressure, fuel-cladding interactions, and differential thermal swelling due to development of thermal gradients in the core. Nuclear transmutations, particularly (n,.alpha.) reactions, play an important role in radiation behavior of alloys, such as stainless steels. Experience with stainless steel has shown that the combination of thermal and irradiation effects lead to hardening and embrittlement of fuel core materials and supporting structural elements. As materials research extended to the study of irradiation effects in fast breeder reactors, a rew radiation phenomenon was discovered. In 1967, Cawthorne and Fulton of Dounreay Experimental Reactor Establishment, UKEA, reported that stainless steel fuel cladding exposed to neutrons developed extensive internal porosity in the form of small cavities or voids. The British finding stimulated activity in the field of irradiation damage as reported in "Radiation-Induced Voids in Metals" dated April 1972, a publication of U.S. Atomic Energy Commission, Office of Information Services. Irradiation-induced swelling results from precipitation of vacancies into voids and interstitials into dislocation loops. Creation of the vacancies and interstitials is the result of collision between a neutron and a lattice atom. In such a collision, a portion of the neutron energy is imparted to a lattice atom sufficient to tear away from its lattice site. The result of this collision is the production of vacant sites and the atoms rejected from their former positions end up in interstitial sites. The dominant features of void swelling can be described as a phenomenon characterized in that it occurs in a fast neutron environment at elevated temperatures in the range 350 -600.degree. C. Swelling increases approximately linearly with fluence after a threshold does is exceeded. The swelling does not appear to saturate. As voids form, density decreases, and hence volume increases. In stainless steels the amount of swelling varies with temperature, with most swelling occuring between 400.degree.-600.degree. C. Operating temperatures of LMFBR cladding and structural components are 350.degree. to about 700.degree. C. and this encompass the temperature of maximum swelling of 300 series stainless steels. Reactor design engineers who have studied the economical implications involved with the phenomenon of radiation-induced swelling have estimated that cost savings of the order of billions of dollars could be realized if void swelling could be reduced only a few percent in the 300 series stainless steel alloys contemplated for use as fuel cladding in fast breeder reactors under design or contemplated for construction. SUMMARY OF THE INVENTION It is, accordingly, a general object of this invention to provide a process or alloy design formula for suppressing the formation of voids induced in metals and alloys as a result of interaction with fast neutrons at elevated temperatures. A principal object of this invention is to provide a low-swelling alloy which can be used as a cladding in a nuclear fuel for a fast breeder reactor or as a supporting structure for the core. A specific object of this invention is to provide a low-swelling austentitic stainless steel. Another object of this invention is to provide a formula or method whereby minor modifications in alloy composition lead to a major reduction in void swelling due to fast neutrons. Another specific object of this invention is to provide a type 316 stainless steel that exhibits very low swelling in comparison to previously known type 316 steels. Still another object of this invention is to provide a solid solution stainless steel alloy suitable for use as a cladding for a fast breeder reactor fuel. A further object of this invention is to provide nuclear fuel elements having a stainless steel cladding which contains void suppressing quantities of Si and Ti. |
claims | 1. A reactor structural member comprising:a surface adapted to be located in a reactor water of a nuclear reactor; anda corrosion potential reducing substance provided on the surface, the corrosion potential reducing substance being formed as particles having a surface on which at least one of Pt, Rh, Ru and Pd is partially attached and being selected from the group consisting of a photocatalytic substance which produces an electromotive force under an irradiation of a light or a radioactive ray in the nuclear reactor and a metal or a metal compound which forms the photocatalytic substance under a condition specified by a temperature and a pressure in the nuclear reactor,wherein the corrosion potential reducing substance is a compound. 2. The reactor structural member according to claim 1, wherein the light in the nuclear reactor is a Cherenkov ray produced in a water-cooled nuclear reactor. 3. The reactor structural member according to claim 1, wherein the photocatalytic substance has a property of an n-type semiconductor. 4. The reactor structural member according to claim 1, wherein the corrosion potential reducing substance at least one of adheres to or is part of a film on the surface of the reactor structural member. 5. The reactor structural member according to claim 1, wherein a mass or a thickness of the corrosion potential reducing substance is such that a current produced by the photocatalytic substance under the irradiation of the light or the radioactive ray is not lower than a sum of threshold current densities of an oxygen and a hydrogen peroxide contained in the reactor water. 6. The reactor structural member according to claim 1, wherein the photocatalytic substance is one or more compound selected from the group consisting of TiO2, ZrO2, PbO, BaTiO3, Bi2O3, ZnO, WO3, SrTiO3, Fe2O3, FeTiO3, KTaO3, MnTiO3 and SnO2. 7. The reactor structural member according to claim 1, wherein the corrosion potential reducing substance is an oxide of Ti or Zr, metal Ti, metal Zr, or a hydrate of Ti or Zr. 8. The reactor structural member according to claim 1, further comprising a corrosion oxide film formed on the surface of the reactor structural member;wherein an adhesiveness of the corrosion potential reducing substance to the corrosion oxide film formed on the surface of the reactor structural member is enhanced by providing a hydrophilic property or by mixing a binder substance. |
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description | The present invention relates to a process and a system for creating isotopes using laser beams in a plasma medium. A “plasma” re tiers to a partially or completely ionized medium, composed of ions and electrons, with no presupposition regarding temperature and/or equilibrium. The isotopes created may be stable isotopes or unstable isotopes, which are then referred to as radioisotopes, in their fundamental energy state or excited energy state, which are then referred to as nuclear isomers. In the remainder of the text, they will be grouped together under the term “isotopes”. Isotopes are in particular used in medicine, in the context of diagnostics and therapies. They may also be used for other scientific or industrial applications, for example for tracing products. Currently the isotopes are generally produced by circular particle accelerators (cyclotrons) or linear particle accelerators, or in nuclear reactors. However, due to the cost and size of these facilities, production must be carried out at a dedicated site, far from the places of use. This involves organizing rapid and secure transport between the places of manufacture and of use. This also prevents the use of a certain number of isotopes, the lifetimes of which are too short to go through this process, but which would exhibit advantages from the point of view of the applications. Document U.S. Pat. No. 6,909,764 describes a process for creating isotopes, in which a target is bombarded by particles generated using a laser beam. Isotopes are created by nuclear reactions produced by the interaction between the target and the particles. The use of a laser beam makes it possible to reduce the size and the cost of the system for creating isotopes. It is thus possible to install the system for creating isotopes in the vicinity of their place of use, eliminating the problem of transport, which is particularly advantageous for isotopes with a short lifetime. The implementation of the process described in document U.S. Pat. No. 6,909,764 made it possible to measure an activity of 2×105 Bq with the Vulcan laser facility (Rutherford Laboratory, UK) for the production of carbon-11 (11C) with a laser of 1020 W/cm2, a pulse duration of 750 femtoseconds, a single laser beam and a single pulse during the interaction of the beam of protons with a solid nitrogen-14 (14N) target. However, positron emission tomography (PET) imaging requires an activity of around 5×108 Bq. The process described in document U.S. Pat. No. 6,909,764 does not therefore make it possible to obtain, with the current lasers, sufficient isotope production level. There is therefore a need for smaller and less expensive systems for creating isotopes, that may be installed in the vicinity of the place of use of the isotopes, and that make it possible to obtain a sufficient production level. The present invention improves the situation. For this purpose, the invention proposes a process for creating isotopes using laser beams, comprising steps of: /1/ converting a target to the plasma state, /2/ bombarding the target in the plasma state with particles generated using a set of laser beams, the set of laser beams being synchronized with the conversion to the plasma state, the fuel and the particles being selected so that the interaction between the target in the plasma state and the particles produces nuclear reactions, /3/ recovering isotopes generated by the nuclear reactions. Step /2/ may be repeated once or several times on the same target. A characteristic duration of the pulses produced by the set of laser beams is, for example, between 50 femtoseconds and 300 picoseconds. Step /2/ may comprise an operation for production of particles by irradiation of a second solid, structured solid, gaseous or liquid target by the set of laser beams. According to one embodiment of the invention, the set of laser beams used to bombard the target is a first set of laser beams, the target being converted to the plasma state by using a second set of laser beams synchronised with the first. A characteristic duration of the pulses produced by the second set of laser beams for example, between one picosecond and twenty nanoseconds. According to another embodiment of the invention, the target is converted to the plasma state by using a Z-pinch machine. According to one embodiment of the invention, the target comprises a hollow, the particles bombarding the target inside the hollow. According to another embodiment of the invention, the target is positioned in an envelope comprising an opening, the particles bombarding the target through the opening. The fuel and the particles may be selected so that the interaction between the target in the plasma state and the particles produces nuclear chain reactions. The invention also proposes a computer program comprising instructions for the implementation of the process when this program is executed by a processor. The invention also proposes system for creating isotopes using laser beams, comprising: /1/ ionization means configured in order to convert a target to the plasma state, /2/ a set of laser beams configured in order to irradiate the target in the plasma state with particles, the set of laser beams being synchronized with the ionization means, the fuel and the particles being selected so that the irradiation of the target in the plasma state with the particles produces nuclear reactions, /3/ isotope recovery means configured in order to recover isotopes generated by the nuclear reactions. FIG. 1 represents a system for creating isotopes using laser beams. The isotopes may be stable isotopes, radioisotopes, or nuclear isomers. The system comprises a first set of laser beams 1 configured in order to allow the irradiation of a target 2 in the plasma state with a beam of particles 3. The target 2 may have various shapes. According to the embodiment of the invention represented in FIG. 1, the target 2 is positioned in an envelope 4 comprising an opening. According to another embodiment of the invention, the target 2 comprises a hollow 8. The particles comprise ions and electrons. The first set of laser beams 1 may comprise one or more laser beam(s). In FIG. 1, three beams have been represented. A characteristic duration of the pulses produced by the laser beams 1 is, for example, between 50 femtoseconds and 300 picoseconds. The intensity, the wavelength, the duration and the shape of the pulses produced by the laser beams 1 are in addition determined so that the bombardment particles have an energy which is close to, or greater than, that of the resonances of the effective cross section of the nuclear reaction in question. A higher energy makes it possible to take into account the energy losses linked to passing through the plasma surrounding the target 2. The intensity of the laser beams 1 is for example of the order of, or greater than, 1018 W/cm2. The system also comprises ionization means 5, configured in order to place the target 2 in the plasma state. According to the embodiment of the invention represented in FIG. 1, the ionization means comprise a second set of laser beams 5. The second set of laser beams 5 may comprise one or more laser beam(s). In FIG. 1, two laser beams have been represented. The pulses that are produced by the second set of laser beams 5 have a characteristic duration of between one picosecond and twenty nanoseconds. The intensity of the laser beams 5 is for example of the order of 1012-1015 W/cm2. According to another embodiment of the invention, the ionization means 5 comprise an axial necking (Z-pinch) machine 9. The system also comprises synchronizing means 7, configured in order to synchronize the first set of laser beams and the ionization means 5. Thus, the production of the particles is synchronized with the production of the plasma, so that the target 2 is irradiated while it is in the plasma state. The system also comprises isotope recovery means 10 configured in order to recover isotopes generated by nuclear reactions. Described below, with reference to FIG. 2, are the steps of a process for creating isotopes using laser beams according to one embodiment of the invention. The process may be implemented by the system described above. The process comprises: a step S1 of initializing the synchronization, a step S2 of converting a target 2 to the plasma state, a step S3 of generating a beam of particles 3, a step S4 of bombarding the target 2 with the particles 3, and a step S5 of recovering isotopes. In step S1, the synchronization means 7 are actuated, so as to control the times for carrying out the steps S2 to S4. Indeed, the creation of the plasma and its bombardment by the particles 3 must be synchronized. In the embodiment of the invention represented, this may be carried out by the synchronization of the first and second sets of laser beams 1, 5. In step S2, the target 2 is converted to the plasma state. The target 2 may be solid, structured solid, gaseous or liquid. In step S3, the particles 3 are generated by irradiation of a second target 6 by the first set of laser beams 1. The initial state of the target 6 may be solid, structured solid, gaseous or liquid. The target 6 is for example a metal sheet of limited thickness. In step S4, the target 2 in the plasma state is bombarded by the particles 3. The fuel and the particles are selected so that the interaction between the target 2 in the plasma state and the beam of particles 3 produces nuclear reactions. According to embodiments of the invention, the fuel and the particles are selected so that the interaction between the target 2 in the plasma state and the beam of particles 3 produces nuclear chain reactions. The production of nuclear chain reactions makes it possible to increase the production of isotopes. Moreover, due to the use of a target 2 in the plasma state, the electrons of the beam of particles 3 interact with the target 2, at the same time as the interaction between the ions of the beam of particles 3 and the target 2. This double interaction also makes it possible to increase the production of isotopes. When the target 2 comprises a hollow 8, the particles 3 bombard the target 2 inside the hollow 8. When the target 2 is positioned in an envelope 4 comprising an opening, the particles 3 bombard the target 2 through the opening. The use of a hollow 8 or of an envelope makes it possible to confine the isotopes produced inside the target 2 or the envelope 4. Step S4 may be repeated once or several times on the same target 2. The accumulation of laser strikes on the same target 2 makes it possible to increase the production of isotopes. The repetition rate is, for example, of the order of 103 Hz. In step S5, isotopes generated by the nuclear reactions are recovered. The isotopes may be recovered directly in the target, in particular when they have been confined in the target 2 or in the envelope 4. The recovery is thus facilitated. According to other embodiments of the invention, an isotope recovery device 10 is positioned in the vicinity of the target 2. The calculations and the first experimental results show a great increase in the rates of reaction when the target 2 is in the plasma state, resulting in isotope production yields that are much higher than the laser methods currently proposed. Furthermore, owing to the process, the emission zone is denser and smaller, which facilitates the recovery of the isotopes. The isotopes created may be stable isotopes, radioisotopes, or nuclear isomers, depending on the applications under consideration. This process makes it possible in particular to produce the carbon-11 (11C) isotope from the 14N(p,α)11C nuclear reaction produced by a beam of protons (p) bombarding a target 2 containing nitrogen-14 (14N) or from the 11B(p,n)11C nuclear reaction by bombarding target containing boron 11B with protons. Other isotopes, such as fluorine-18 (18F), nitrogen-13 (13N) and oxygen-15 (15O) may be produced from the following reactions: 18O(p,n)18F, 20Ne(d,n)18F, 16O(p,α)13N, 13C(p,n)13N, 14N(d,n)15O and 15N(p,n)15O. The isotopes created depend on the fuel 2 and on the particles 3 used. Described below is an example of the implementation of the process for creating isotopes. The first set of laser beams comprises, in this example, a laser beam that produces a laser pulse delivering 20 J in 1 ps at the wavelength of 0.53 μm. The laser beam 1 is focused on a sheet of aluminum 6 having an initial thickness of 20 μm. The beam of particles 3 generated is a beam of energetic protons. Protons having an exponentially decreasing energy spectrum with a maximum energy of around 12 MeV are sent to a target 2 of natural boron (20% 10B and 80% 11B) converted into plasma just before the arrival of the beam of protons. The conversion to the plasma state is carried out by another laser beam 5 delivering 300 J in 1.5 ns at the wavelength of 0.53 μm. The carbon-11 (11C) produced on a boron target by the 11B(p,n)11C reaction is measured after striking by the activation of the target 2. The 11C produced is then measured in the target 2. In order to evaluate the improvement in the production yield, a solid boron target is also bombarded by a beam of protons under the same conditions. A great increase in the production of 11C was observed when the boron target is in plasma form compared to the solid. The process and the system described above thus enable the creation of facilities that are less expensive, more efficient and may operate on site, in particular for the production of isotopes for medicinal diagnostic and therapy purposes. Of course, the present invention is not limited to the embodiments described above by way of example; it extends to other variants. |
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054065941 | summary | The present invention relates to an injection system of cryogenic pellets formed of solid slate hydrogen isotopes in a temperature range between about 4 and 19K to be supplied to machines for the magnetic confinement of plasma (for example F.T.U.). As it is known, plasma of the above machines should be supplied during the discharge for maintaining a predetermined density. The inlet, of hydrogen isotopes into the gas phase was proved to be ineffective due to the fast ionization of the gas preventing the plasma from being penetrated. A solution of this problem is the injection of the above machines with a solid state reactant at such a rate as to assure the deposition of the fusionable material at the deepest, layers of the plasma column. Such method further allows the material removed from the pellets to be deposited more or less deeply and then the density profile of plasma to be shaped as the dimensions and the rate of the pellets change. A general improvement of the confinement parameters has been noticed in plasma having a marked density at the centre so that injection rates higher than 2 km/s are being requested for such machines today, and even higher rates (5-10 km/s) shall be reached in the future, an impossible performance for the injection systems used today. Such systems can be divided into two classes: centrifugal injection systems; PA1 pneumatic single-stage injection systems. PA1 1) the pressure at the base of the pellet falls off very quickly by dynamic effect so that a high acceleration cannot be maintained for the requested time; PA1 2) the maximum acceleration is limited by the mechanical characteristics of the pellet (about 10.sup.7 m/s.sup.2); PA1 3) difficulty of providing a sufficient propelling gas stock at high temperature (500.degree.-600.degree. C.). PA1 1) the dynamic pressure drop at the base of the pellet is partially compensated by the leading edge of the pressure pulse generated at the input of the barrel; PA1 2) the pulsing pressure allows much more higher peak values (about 1200 bars) to be reached without damaging the pellet; PA1 3) the adiabatic compression allows the gas temperature to be increased up to several thousands of degrees for a time of about 100 microseconds, which also permits the sound propagation rate in the propelling gas and then the maximum rate of the pellet to be increased. PA1 a pneumatic system operating with hydrogen and/or helium and formed of one or more two-stage or multi-stage propulsion systems, the relative inlet circuits, and one or more decompression chambers; PA1 a cryogenic device formed of a Dewar flask containing liquid helium, a circuit for transferring and recovering the cooling fluid, and one or more conventional (in situ) or alternative cryostats like that described afterwards provided each with one or more launching barrels, in which the cryogenic pellets are solidified; PA1 a vacuum system including electrovalves, electropneumatic valves, rotating and turbomolecular pumps; and PA1 a set of equipment for the automatic remote control of the whole system and for collecting and supplying diagnostic data to the central processing unit. Centrifugal injection systems are characterized by a high injection frequency (number of pellets per second double as high as the rotation frequency of the system), however, by a low rate (.ltoreq.800 m/s) and a poor shot. These features are cause of drawbacks as the low rate does not allow the plasma column to be deeply penetrated, and the poor shot causes a large proportion of pellets not to pass through the entrance window of the plasma. Pneumatic injection systems, to which the system of the present invention belongs, are on the contrary characterized by a better shot and a higher rate (1600 m/s for single pellets, 1200 m/s for multiple injections up to a maximum of 6 pellets per second in two seconds). With the injection systems of the status of art based upon single stage propulsion systems the above rates are the highest limit which can be reached due to the inherent ineffectiveness of the propulsion systems essentially depending upon the following reasons: In case of conventional systems the sonic or infrasonic launching rate causes very cumbersome decompression chambers to be used (about 800/1000 l). The present invention seeks to avoid the drawbacks and the limits of the present systems and to provide a system for producing, accelerating and letting into magnetic confinement machines one or more cryogenic pellets at higher rates than those reached nowaday with the pneumatic single-stage propulsion systems. This has been provided according to the invention by using one or more two-stage or multi-stage propulsion systems provided with a pair of special valves, better described afterwards, one of which is a control valve, the other a cutoff valve. The use of pneumatic two-stage or multi-stage propulsion systems allows the injection rate to be increased up to about, 3 km/s (with non-protected pellets) because: It should be pointed out, that the above mentioned upper limit rate is determined in the present invention by the fragility of the pellet which cannot stand accelerations beyond the limit; which is imposed by the mechanical properties of the solid and not by the limits originating from the propulsion system itself having on the contrary a higher capacity than that requested for such application. The above mentioned quick control valve of the propulsion system is operatively different from those used nowadays. Actually the quick control valves used in similar systems are essentially electropneumatic valves in which the pressure-containing member is opened by a coil and closed by the differential pressure so as to reduce the flow of the propelling gas following the pellet. In the present case, both the opening and the closure of the pressure-containing member are based upon the principle of the differential pressure and then are executed very rapidly so as to improve the efficiency of the propulsion system and to limit the flow of the propelling gas. The above mentioned cutoff valve seeks to separate the propulsion system from the launching barrel during the condensation of the pellet, and to contribute to the limitations of the flow of the propelling gas following the pellet. Its characteristic is of assuring a good vacuum sealing with a considerably reduced dead volume and with an orifice having a diameter corresponding to the inner diameter of the launching barrel. In order to couple the injection system to an user a system is provided for removing and evacuating the propelling gases, said system being formed of one or more decompression chambers which may have considerably reduced dimensions (about 100/200 l) with respect to those used up to now because of the supersonic rate of the pellets. According to another advantageous feature the cryogenic device of said system has been miniaturized in order to reduce to the minimum the consumption of liquid helium, and has been provided with a particular pipe for transferring the cooling fluid (cold vapours of helium), thus allowing a better operation and preventing thermal oscillations during the cooling. According to a preferred embodiment the injection system according to the invention includes: |
description | The present patent application is a continuation of application Ser. No. 10/637,240 filed Aug. 7, 2003 now U.S. Pat. No. 6,949,476, which is a Continuation of patent application Ser. No. 09/821,323 filed Mar. 28, 2001, now U.S. Pat. No. 6,696,369, which is a divisional of patent application Ser. No. 09/540,072 filed Mar. 31, 2009, now U.S. Pat. No. 6,400,015. 1. Field of the Invention The present invention relates to the field of semiconductor device fabrication, and more specifically to a method and structure for constructing a structure using semiconductor device fabrication methods that shields semiconductor devices. 2. Discussion of Related Art Today integrated circuits are made up of literally millions of active and passive devices such as transistors, capacitors, and resistors. In order to improve overall chip performance, some devices may need to be shielded from the electromagnetic interference (EMI) from adjacent devices, from heat, and from light. A novel device structure and method for shielding a region on a semiconductor is described. In the following description numerous details are set forth such as specific materials and processes in order to provide a thorough understanding of the present invention. In other instances, well known semiconductor processing techniques and machinery have not been set forth in detail to avoid obscuring the present invention. The present invention is a novel device structure and method for shielding individual or a selection of semiconductor devices from conductive and/or radiated energy. Such as, for example, electromagnetic interference (EMI) from radiation originating outside the semiconductor or from adjacent devices on the semiconductor. The present invention may be also used to direct thermal energy relative to a semiconductor, or to shield a semiconductor from light. In an embodiment, a Faraday cage is constructed on a silicon substrate, and encloses one or more semiconductor devices within a structure of metal. The semiconductor devices having input/output leads or pass-thrus that pass through the Faraday cage walls at one or more insulated locations. The embodiment provides that interconnects and vias both inside and outside the disclosed Faraday cage may also be constructed in the same layers used to construct the Faraday cage. To construct the Faraday cage with insulated pass-thrus, vias and interconnects, alternating layers of tungsten (W) and aluminum (Al) are used. The use of tungsten (conductive metal) will fill in via openings between interconnects to create plugs or filled vias. This tungsten layer will add metal layers to the Faraday cage wall(s) at the same time. Alternating with the tungsten layers, interconnects are etched from layers of aluminum or aluminum alloy. As with the tungsten layers, each aluminum layer will also add a layer in constructing the walls of the Faraday cage. Left within layers of the metal Faraday cage walls are pieces of dielectric material that will make up the insulation framework (frame) around each pass-thru. This insulation surrounds or frames each pass-thru lead at the point where it passes through the metal Faraday cage wall. For each pass-thru insulation or frame construction, one metal layer will have one horizontal block (base frame), the next metal layer will have one pair of vertical bars (vertical frame pairs), and finally the insulation will be complete with the next layer addition of another horizontal bar (top frame). However it is possible for multiple insulated pass-thrus to share common frame components. The process of depositing layers of dielectric and metal, positive or negative photoresist and etching the layers to form vias and interconnects is well understood. In addition, at the same time for this embodiment, trenches (slots) will be etched in dielectric around the semiconductor device(s) to be enclosed by the Faraday cage. The slots will be filled with tungsten. Alternating the tungsten layers are layers of aluminum. These will be etched to add more layers to the Faraday cage walls and at the same time construct the interconnects. Finally a last metal layer will completely cover the area enclosed by the Faraday cage wall(s) to act as a roof or lid over the walls. The following embodiment will describe the process of joint fabrication of the Faraday cage, enclosing with metal, one or more semiconductor devices having insulated pass-thrus (input/output leads or conductors) and layers of interconnects joined by vias to the semiconductor devices. Referring to FIG. 1, cross-sections A-A and B-B will be shown as figure a and figure b designations respectively in later illustrations. These cross-sections (A-A & B-B) appear throughout many of the figures to show a simultaneous construction of the Faraday cage walls 102 (B-B) with the vias (shown after FIG. 2) (A-A) and the interconnects 108 (A-A). Although FIG. 1 shows the construction of two pass-thru leads (pass-thrus) 108 and insulators 106 at the front and back walls 102, the later figures only illustrate construction of a single pass-thru 108 and a single insulator 106. This is done for clarity, however it is to be understood that any number of pass-thrus 108 and insulators 106 may be fabricated in a Faraday cage 100 at different levels. As shown in FIG. 2, prior to beginning depositions for Faraday cage 100 (FIG. 1) construction, the semiconductor device 104 such as an MOS transistor having a gate 103 with a gate oxide beneath 106, and a pair of source and drain regions 105, have been constructed on a wafer substrate (substrate) 101. The substrate may be made from such materials as silicon (Si), gallium arsenide (GaAs), or one of the silicon-on-insulator (SOI) materials such as silicon-on-sapphire (SOS) or silicon-on-diamond. The transistor may link with other transistors to function in a variety of tasks such as a resister, capacitor, memory storage device, sense amp, or an input/output buffer. Turning to FIGS. 3a & b, a first coating of the dielectric 120 (first dielectric layer) is deposited as an insulative layer over the substrate 101 and the previously fabricated semiconductor device 104. The dielectric material for this embodiment is silicon dioxide (SiO2) but may also be silicon nitride (Si3N4), phosphorus-doped silicon oxide (PSG), or boron/phosphorus-doped silicon oxide (BPSG). A process known as patterning is next performed. This involves applying a photoresist coating over the substrate and then using well known photolithography steps such as masking, exposing, and developing, to form a patterned photoresist layer. The underlying material is then etched in alignment with the patterned photoresist layer. As shown in FIGS. 4a & b, the first dielectric layer 120 is coated with the photoresist layer 123 within which is formed a pattern 123. The pattern 123 in the photoresist is reacted and the non-reacted photoresist material is then removed. The next step is an etch of the first dielectric layer 120 that follows the shape of the photoresist pattern 123. With this etch, the photoresist layer protects the dielectric layer 120 beneath from the etch operation. Referring to FIG. 5a, first via openings 126 are etched within the first dielectric layer 120. These first via openings 126 are etched through the first dielectric layer 120 exposing a portion of the semiconductor 104 surface. Turning now to FIG. 5b, at the same time a first slot 124 is etched in the first dielectric layer 120, and surrounds the semiconductor device(s) 104 (FIG. 5a) to be EMI shielded. The first slot 124 begins the formation of the Faraday cage walls 102 (FIG. 1). This etch and subsequent etches may be accomplished by a variety of methods such as with a wet chemical (wet-chem) or by one of the plasma etches such as a reactive ion etch. Next, but not shown, a barrier coating may be applied to the etched dielectric 120 surface to improve adhesion between a metal coating to be next applied and the dielectric 120. This coating may be titanium or titanium nitride material. This barrier coating may be used on any dielectric surface when a metal coating will be applied over the dielectric. Now, a fill layer of a material (first conducting layer) 125 is deposited as shown in FIGS. 6a & b. Turning now to FIGS. 7a & b, there is seen the first conducting layer 125 after it has been polished back to the first dielectric layer 120. This polish is accomplished by a chemical etch and a chemical-mechanical polish (CMP) may be used prior to the chemical etch. The first conducting layer 125 has filled in the via openings 126 (FIG. 5a) forming vias 127 (filled vias, via plugs, or plugs) and filled in the first slot 124 (FIG. 5b) to form a first layer of wall 128 in constructing the Faraday cage walls 102 (FIG. 1). The conducting material used to fill in the vias for this embodiment is tungsten (W) but may be another metal such aluminum (Al) or a non-metal such as polysilicon (Si). Referring now to FIGS. 8a & b, a first metal layer or metal one (M1) 130 of aluminum (Al) is deposited over the dielectric top surface 228. While the metal layers for this embodiment are made of aluminum, other well known metals used for interconnects, such as copper, may be used. Turning to FIGS. 9a & b are displayed the after-patterning results. A first layer of interconnects (first interconnects) 225 are formed in the M1 130. At the same time with M1 130, a second layer of the wall 224 is placed over the first layer of the wall 125 that is forming the overall wall structure 102 (FIG. 1). Referring now to FIGS. 10a & b, there is seen a deposit of a second dielectric (SiO2) layer 220. This second dielectric layer 220 fills in around the second layer of the wall 224 construction and the first layer of interconnects 225. The second dielectric (SiO2) layer 220 is patterned (photoresist+etch) as described above but not shown here. The results of the patterning is displayed in FIGS. 11a & b. Via openings 226 are etched until surfaces on the first interconnects 225 are exposed. In addition, a second slot 324 is constructed within the second dielectric layer 220 and positioned above the first and second layers of wall 125, 224. At a selected location, the second slot 324 construction leaves a base frame 350 within, of dielectric (from second SiO2 layer 220), to begin construction of the pass-thru insulation 106 (FIG. 1). Referring now to FIGS. 12a & b, a second fill layer of tungsten (second conducting material) is deposited and then polished back to the second dielectric layer 220. After polish, a tungsten filled third layer of the wall 326 remains over the previously constructed walls 125, 224. At the same time, vias 227 are created. In addition, the tungsten layer 326 fills in around the base frame 350. FIGS. 13a & b show a deposit of a second metal layer or metal two (M2) 230 of aluminum. After deposition, the M2 230 is patterned as described above. Turning now to FIGS. 14a & b, after etching, a second layer of interconnects 325 (second interconnects) are formed from M2. At the same time, a fourth layer of the wall 424 is formed from M2 that is positioned over the previously constructed layers of wall 125, 224, 326. Additionally, within the fourth layer of the wall 424 there remain two vertical spaces (vertical frame slots) 360 over each base frame 350. Above the base frame 350 and between the two vertical frame slots 360 passes the pass-thru 380 from the interconnects 325 to circuitry outside the partially constructed wall 102 (FIG. 1). At this point (FIGS. 14a & b), there is constructed in alternating tungsten and aluminum, four layers of the partially constructed wall 125, 224, 326, 424. The metal (Al) pass-thru lead 380 connects from the second interconnect 325 and passes through the partially constructed insulator 350, 360 to outside circuitry (not shown). Referring now to FIGS. 15a & b, a third dielectric (SiO2) layer 320 is deposited. The third SiO2 layer 320 fills in the vertical frame slots 360 (FIG. 14b) to form the vertical frame pairs 361 and later the top frame 550 (shown in FIG. 16b later) of the insulator 106 (FIG. 1). The next patterning operation is not shown but uses the techniques described above with the results shown in FIGS. 16 a & b. The third dielectric 320 layer is patterned to form a third slot 524 above the previous layers of wall 125, 224, 326, 424. Within the third slot 524 is formed the top frame 550 (SiO2) over the vertical frame pairs 361 (SiO2). The pass-thru is now enclosed with insulation (SiO2) at the wall 125, 224, 326, 424. In addition, via openings if needed may be created that expose surfaces on the interconnects 325 beneath. Turning now to FIGS. 17a & b, a third fill layer of tungsten (third conducting layer) (not shown) is deposited and then polished back to the third dielectric layer 320. The third fill layer fills in the third slot 524 (FIG. 16b) to form the fifth layer of the wall. Referring now to FIGS. 18 a & b, a metal layer (M3) 330 of aluminum is deposited to form the lid 560 to complete the enclosure of the semiconductor(s) (not shown). The ML3 330 may be patterned (not shown) to shape the lid 560 or add other interconnect circuitry (not shown) and completes the basic construction of the Faraday cage 100 (FIG. 1). The lid 560 now covers the walls 102 (FIG. 1) and the entire area contained within the walls 102 (FIG. 1). Afterward, a last coating of dielectric 510 may be deposited to place a barrier coating or sealant on the lid 560. Referring to FIGS. 19a & b, there is shown an embodiment having more layers added to create another interconnect layer 425 and vias not shown here may be fabricated connecting the lid 460 to other interconnects below. There is also seen two pair of insulated pass-thrus 400. Here, the two pair of insulated pass-thrus 400 are constructed on differing layers or levels. In this illustration, the individual insulated pass-thrus 400 and pass-thru pairs 400 are separated from each other by dielectric material 402. Turning to FIG. 20 is shown a pair of insulated pass-thrus 502 separated by both dielectric material 504 and metal material 506. Referring to FIG. 21 is seen two pair of insulated pass-thrus 604 each on a different level and separated within each insulated pass-thru pair 604 and between pass-thru pairs 604 by both dielectric 604 and metal 606 material. In FIG. 22 is illustrated an insulated pass-thru 702 in which the pass-thru lead 702 is supported by a dielectric material 706 but the vertical frame pairs 703 and the top frame 704 are spaces (voids) filled with air. It should also be understood that any number of layers of insulation or metal layers, M1, M2, M3 (metal four, metal five, etc.) may be used to construct multiple pass-thrus on a single level and multiple levels of interconnects and pass-thrus. In addition, for other embodiments, the metal layer deposited to form the lid in the disclosed embodiment may be patterned into interconnect circuitry for devices outside the Faraday cage. Further, for other embodiments, there may be subsequent layers deposited above a Faraday cage lid to add interconnects, vias, and other Faraday cage walls to circuitry stacked outside and/or higher than a given Faraday cage. This method of forming the Faraday cage could be employed to construct structures for other applications such as to redirect electrostatic discharge, to distribute thermal energy, or to shield light sensitive devices such as, for example, might be used in optical switching. |
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abstract | Apparatus and method are provided for the treatment of uranium-contaminated soil by using comprehensive joint technology. The apparatus include the pumping system, the electrokinetic remediation system, elution system, remediation-separation system and recharge system. The remediation technologies (i.e. chemical, photolysis and electrokinetic) are used to remedy the uranium-contaminated soil. First, extract uranium from the contaminated areas and make the ionized uranium extract from the soil phase to the solution phase. Then, use the electrokinetic remediation technology to drive uranium enrichment electromigrate to near the anode. Finally, return the repaired-soil and groundwater back to anode area and recharge well, respectively. This comprehensive joint apparatus can reduce the uranium volume in the contaminated soil or water, and recycle the obtained uranium, which are cleaning processes and have no secondary pollution. |
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051704250 | claims | 1. An x-ray diagnostics installation comprising: an x-ray tube which generates an x-ray beam in beam path; a primary radiation diaphragm disposed in said beam path and having a plurality of beam-interacting elements; an x-ray image intensifier video chain means for processing an image of an examination subject which attenuates said x-ray beam and for producing a visible image of said x-ray image on a display, said image intensifier video chain including an image memory; a plurality of sensor means for generating electrical signals corresponding to the respective positions of said beam-interacting elements of said primary radiation diaphragm; and processing means connected to said image intensifier video chain means and to said sensors for acting on signals stored in said image memory for simulating the effect of said beam-interacting elements on said x-ray beam. blanking circuit means connected to said image memory and to said control computer for generating blanking signals corresponding to at least one of the contour, position or transparency of said beam-interacting elements of said primary radiation diaphragm for correspondingly blanking the signals stored in said image memory; and summing means, to which an output of said blanking means and an output of said simulation means are supplied for adding said outputs and thereby superimposing the output of said blanking circuit means on the output of said simulation means. 2. An x-ray diagnostics installation as claimed in claim 1 wherein said processing means includes a control computer connected to said sensor means and simulation means connected to said image memory and to said control computer for simulating the effect of said primary radiation diaphragm. 3. An x-ray diagnostics installation as claimed in claim 2 wherein said image intensifier video chain means generates a video signal, wherein said beam-interacting elements of said primary radiation diaphragm have respective beam absorption factors, and wherein said simulation circuit includes means for amplitude-matching of the video signals for attenuating said video signal to an extent corresponding to the absorption factors of said beam-interacting elements. 4. An x-ray diagnostics installation as claimed in claim 2 wherein said image intensifier video chain means generates a video signal, and wherein at least one of said beam-interacting elements blocks a portion of said x-ray beam, and wherein said simulation circuit includes blanking circuit means for blanking a portion of said video signal corresponding to the position of said beam-interacting element which blocks said x-ray beam. 5. An x-ray diagnostics installation as claimed in claim 2 wherein said image intensifier video chain means generates a video signal, and wherein said simulation means includes means at an input of said simulation means for logarithmizing said video signal to produce a delogarithmized signal and means at an output of said simulation means for delogarithmizing said logarithmized signal. 6. An x-ray diagnostics installation as claimed in claim 2 further comprising: |
description | The present patent application/patent claims the benefit of priority of U.S. Provisional Patent Application No. 61/786,749, filed on Mar. 15, 2013, and entitled “COMPUTED RADIOGRAPHY IMAGING PLATES AND ASSOCIATED METHODS OF MANUFACTURE,” the contents of which are incorporated in full by reference herein. The U.S. Government has rights to the present disclosure pursuant to Contract No. AC05-00OR22800 between the U.S. Department of Energy and Babcock and Wilcox Technical Services Y-12, LLC. The present disclosure relates generally to the medical and industrial imaging fields. More specifically, the present disclosure relates to computed radiography (CR) imaging plates and associated methods of manufacture. Computed radiography (CR) is a reusable digital radiography modality. It utilizes reusable imaging plates incorporating a relatively thin, flexible construction. These imaging plates include a photostimulable phosphor layer, which converts radiation into a latent image. This image is retrieved using a laser-catalyzed photoemission phosphor layer. A CR scan produces a digital image. The imaging plate is erased with light to remove any remaining latent image that was not retrieved by the laser, thereby allowing the imaging plate to be reused. From a technician's standpoint, there is little difference between film and CR radiography, and general handling and use precautions are very similar, the biggest differences being related to the operation of the CR scanner, as opposed to the film processor. Much of the advancement of radiographic technologies comes from the medical sector. Industrial non-destructive testing (NDT) markets are significantly smaller. The medical sector primarily uses energies of in the range of about 25-150 KeV for imaging. Advancements in the medical sector's technologies, equipment, and protocols emphasize lowering the radiation received by a patient undergoing a given test. Most conventional imaging plates are constructed in a similar manner, and the following is not an exhaustive list of construction options, but covers the basic construction and functionality of the major layers. Starting from the back of the imaging plate and moving to the front (i.e., source side), the imaging plate includes: A backing layer that provides structural support. This backing layer is relatively thick and made of plastic or the like. A group of layers designed to reflect some wavelengths of light, but not others; in most plate constructions, the light emitted by the phosphor layer is reflected to decrease dose to the patient with only an acceptable sacrifice of resolution. Imaging plates designed for higher resolutions can be specifically designed to not reflect light emitted by the phosphor layer. Most imaging plates can be designed to not reflect light from the laser used to stimulate the phosphor layers. The phosphor layer that reacts with x-rays to produce a latent image. The phosphor layer releases light in proportion to the radiation absorbed when stimulated by a laser. A clear plastic protective layer used to protect the phosphor layer yet allow light to enter and leave the imaging plate. This layer is normally thinner on higher resolution plates. Binding layers may also be present between other layers.The imaging plates are built by or for film manufacturers and have a very similar construction to film. However, film does not have the reflective/antireflective layers, and film replaces the phosphor layer with a layer of light and x-ray sensitive silver salts. Commercially available CR imaging plates are optimized for use in the KeV energy range. The typical setup for CR is as follows. The imaging plate is placed inside a light-tight container, typically a film/CR cassette or sleeve, where the test object itself can serve as the light-tight container, or the x-ray area can be darkened. Between the object and the source, a collimator(s) can be used to lessen the effect of scatter by decreasing the size of the primary beam. At the source, some filtration can be used to preferentially remove lower energy radiation from the primary beam, decreasing scatter and effectively increasing the average energy of the primary x-ray beam. In medical radiography, great care is taken to minimize patient radiation exposure, and reduced radiation exposure is one focus of the present disclosure. U.S. Pat. No. 4,712,011 is related to the present disclosure: An X-ray image intensifier tube which includes a luminescent layer with an absorption material having a high absorption for secondary X-rays which are generated in the original luminescent material and which are intercepted to only a very small extent by the original luminescent material. The absorption material may be included in the layer of luminescent material in homogeneous form as well as in recesses in said layer. In addition to improved resolution, a higher efficiency can be achieved by ensuring that, upon interception of the secondary radiation, the absorption material generates luminescent or secondary radiation which is intercepted by the original luminescent material. U.S. Patent Application Publication No. 2010/0034351 is also related to the present disclosure: Disclosed are a radiation image conversion panel, which provides high luminance, an image without white or black defects, an image free from cracks and an image with reduced unevenness, and its manufacturing method. Also disclosed is an X-ray radiographic system employing the radiation image conversion panel. The radiation image conversion panel of the invention comprises a substrate and provided thereon, a reflection layer, a phosphor layer and a protective layer in that order, wherein the phosphor layer is composed of a phosphor crystal in the form of a column, and the reflection layer is formed by vapor phase deposition of two or more kinds of metals. In the present disclosure, a novel screen (also referred to herein as a layer, filter, grid, and/or intensifier) is coupled to the phosphor layer, allowing electrons and/or low energy x-rays to impart their energy on the phosphor layer, while decreasing internal scattering and increasing resolution. The radiation dose needed to perform radiography can also be reduced as a result. Thus, the present disclosure provides a novel process for manufacturing and design of a CR imaging plate that provides significant improvement in resolution for CR systems. This design incorporates a metallic screen or other metallic structure as the phosphor layer's substrate backing, as opposed to a conventional plastic substrate. This design places the phosphor layer in intimate contact with the intensifying screen. As an added advantage, the metallic substrate can be designed to act as a secondary scatter filter for the phosphor layer. A second, and potentially mutually non-exclusive, mechanism for accomplishing intensification (i.e., resolution enhancement/exposure reduction) is to intersperse finely divided metal or metallic compounds into the phosphor layer to achieve the intensification at the grain level of the phosphor. Previous to the present disclosure, the conventional plastic substrate on commercially-available products has its own luminescence that interferes with the images. This problem is especially apparent in higher-energy x-ray applications. With current imaging plate designs, the use of intensifying screens is extremely difficult due to proximity limitations inherent in the current imaging plate designs, which incorporate a poly substrate on one side of the phosphor layer with a protective coating on the other side of the phosphor layer, thereby preventing the proper placement of such intensifying screens. With the use of intensifiers, as in the present disclosure, and with an increase in resolution, there is the potential to greatly reduce specimen exposure in low-energy (medical) x-ray operations, for example. In one exemplary embodiment, the present disclosure provides an imaging plate for use in radiography applications, including: a phosphor layer; and a metallic or metallic compound intensifying layer coupled to the phosphor layer. Optionally, the imaging plate also includes one of a reflective layer or an anti-reflective layer disposed between the phosphor layer and the metallic or metallic compound intensifying layer. Optionally, the imaging plate further includes a clear protective layer coupled to the phosphor layer opposite the metallic or metallic compound intensifying layer. Optionally, the imaging plate still further includes a structural support layer. Optionally, the imaging plate still further includes one or more binding layers. Preferably, the metallic or metallic compound intensifying layer primarily includes a metal or metal alloy—one or more of lead, copper, tungsten, tantalum, steel, stainless steel, brass, aluminum, nickel, cobalt, silver, gold, platinum, osmium, ruthenium, niobium, hafnium, zinc, cadmium, bismuth, tin, iridium, molybdenum, manganese, titanium, vanadium, scandium, thorium, and uranium. In another exemplary embodiment, the present disclosure provides an imaging plate for use in radiography applications, including: a phosphor layer; and a plurality of metallic or metallic compound intensifying particles disposed one or more of on a surface of and within an interior portion of the phosphor layer. Optionally, the imaging plate also includes one of a reflective layer or an anti-reflective layer disposed between the phosphor layer and the metallic or metallic compound intensifying particles. Optionally, the imaging plate further includes a clear protective layer coupled to the phosphor layer opposite the metallic or metallic compound intensifying particles. Optionally, the imaging plate still further includes a structural support layer. Optionally, the imaging plate still further includes one or more binding layers. Preferably, the metallic or metallic compound intensifying particles primarily include a metal or metal alloy—one or more of lead, copper, tungsten, tantalum, steel, stainless steel, brass, aluminum, nickel, cobalt, silver, gold, platinum, osmium, ruthenium, niobium, hafnium, zinc, cadmium, bismuth, tin, iridium, molybdenum, manganese, titanium, vanadium, scandium, thorium, and uranium. Optionally, the phosphor layer is coated or doped with the metallic or metallic compound intensifying particles. Referring specifically to FIG. 1, in one exemplary embodiment of the CR imaging plate 10 of the present disclosure, a metallic or other similar high density material layer 12 (also referred to herein as the intensifying layer 12) is coupled to the phosphor layer 14. Reflective or anti-reflective layers or surfacing 16 optionally may be disposed between the intensifying layer 12 and the phosphor layer 14, depending on desired use of the imaging plate 10. The thickness and material composition of the intensifying layer 12 will affect the effective quantum efficiency of the detector (not illustrated) and change the dose response curves and the effective resolution. Different thicknesses and material compositions of the intensifying layer 12 will be better suited for different energies of x-rays and applications. The conventional backing layer (not illustrated) may be able to be removed and its construction is less vital to the present disclosure. It may still be needed to provide backing if a sufficiently thin or otherwise structurally unsound or non-robust intensifying layer 12 is used. Since it is no longer being used to provide a mounting surface for the phosphor layer 14, the options available for the backing layer's construction are greater, and carbon fiber composites or other fibrous or non-fibrous materials may provide a more optimal backing than previously possible. The clear protective layer 18 remains roughly the same as in conventional applications, and serves the same purposes. Typically, in conventional CR plate designs, the clear protective layer 18 is the first layer that the radiation beam encounters. In the present disclosure, however, it could be the first layer or the last layer, depending upon the optimization required. Binding layers (not illustrated) may be present between other layers. Depending upon the construction chosen, some sealant around the imaging plate 10 may be necessary to maintain water/moisture tightness around the phosphor layer 14. A layer or other features to facilitate imaging plate handling by radiographers or the CR scanner are not strictly needed, but may improve performance and/or usability. Increases to resolution and decreases to radiation exposure (i.e., dose) are the outcome of the use of the imaging plate 10 of the present disclosure. The imaging plate 10 of the present disclosure is radiographed with the intensifying layer 12 typically towards the object being imaged 20. The clear protective coating 18 is facing the laser beam during scanning. The advantages of the imaging plate 10 of the present disclosure include better portability, better efficiency, and better resolution. It should be noted that, as used herein, “layer” may also refer to a layer, a coating, or any other comparable structure. Thus, stated differently, the present disclosure affixes the phosphor (active) layer 14 of the plate 10 to a metallic substrate 12, as opposed to a conventional polymer substrate. By eliminating the polymer substrate, the present disclosure eliminates all florescence and adverse secondary energy deposition into the phosphor layer 14 emanating from within the polymer substrate. Such image interferences are common with the use of polymer substrates. This action is observed more at higher primary energy sources. By reducing the interference from the polymer substrate, the present disclosure is able to achieve both a higher contrast-to-noise ratio and a better modulation transfer function, while reducing specimen exposure. With the phosphor layer 14 in intimate contact with the metallic substrate layer 12, there will be an efficient and coherent interaction of secondary energy emanating from the metallic substrate layer 12 and being deposited in the phosphor layer 14. The structure of the phosphor layer 14 and the metallic substrate layer 12 could be formed by vapor deposition, electroplating, a mechanical coating process, a mechanical process (i.e. a vacuum/presser/clamping process), a chemical bonding process, and/or the like, depending upon final plate usage requirements. The size and shape of the phosphor micro structure can be modified to fit the final usage requirements. The metallic substrate 12 performs the following functions, depending on material and material thickness. It provides a durable backing material that protects the phosphor layer 14 from mechanical damage. It provides a strong substrate that maintains order and coherence of the phosphor layer 14. A notable function is to provide an intensifying screen for the phosphor layer 14 that increases efficiency of the phosphor (active) layer 14, such that a lower level of primary beam exposure to an object can be used to achieve an equivalent or better image resolution. This can be used in the medical field to achieve lower patient exposure levels for equivalent image qualities. This metallic intensifying media 12 can also be combined with the phosphor layer 14 as a finely divided metal or metallic compound, for example. In this granular embodiment, the metallic intensifying media 12 can be modified to be a doping agent for the phosphor layer 14 to alter and optimize the efficiency of the phosphor layer 14. This allows for optimization of the imaging plate 10 for a given application. A secondary function of the metallic substrate 12 is to be an energy filter, or X-ray beam hardener. Another function of the metallic substrate 12 is to absorb secondary scatter radiation, mostly emanating from the specimen 20. Variance in the material used in the metallic substrate 12 changes the quantum efficiency of the phosphor layer 14 by changing the energy and abundance of the secondary energy emanating from the metallic substrate 12 and being deposited in the phosphor layer 14. The metal substrate thickness can be modified to also act as a low energy (scatter radiation) filter for the phosphor layer 14 to increase the signal-to-noise ratio, thus improving the image quality. The thickness of the metallic substrate 12 is dependent on the energies utilized and the amount of intensification desired, as well as the scatter absorption desired. For designs where the intensifying substrate thickness requirements are too thin to provide a robust enough backing for the phosphor layer 14, the intensifying substrate 12 can be plated or bonded to a separate substrate (not illustrated) for strength. This strengthening substrate can be an ordered or non-ordered fibrous backing or some other design. Some of the metals that have been used as the metallic substrate 12 include: lead, copper, tungsten, and tantalum. This does not preclude the option of using any metallic alloy, steel, stainless steel, brass, aluminum, nickel, cobalt, silver, gold, platinum, osmium, ruthenium, niobium, hafnium, zinc, cadmium, bismuth, tin, iridium, molybdenum, manganese, titanium, vanadium, scandium, thorium, or uranium, for example. Again, there can be a visible light barrier 16 that is also non-reflective between the substrate 12 and the phosphor layer 14. This light barrier's thickness and material will depend upon several factors. Some of these factors include, but are not limited to, the following: primary energy used, metal substrate material and alloy, metal substrate thickness, phosphor composition, phosphor layer construction, specimen material, and specimen construction. This visible light barrier 16 is tuned to absorb the light wavelength of the excitation laser, while either reflecting or absorbing the wavelength of the light emitted by the phosphor layer 14. If specimen exposure is of concern, then a visible light barrier 16 that reflects the light emitted by the phosphor layer 14 is utilized. If maximum resolution is desired, then a visible light barrier 16 that absorbs the light emitted from the phosphor layer 14 is utilized. Although the present disclosure has been illustrated and described herein with reference to preferred embodiments and specific examples thereof, it will be appreciated by those of ordinary skill in the art that other embodiments and examples may perform similar functions and/or achieve like results. All such equivalent embodiments and examples are within the spirit and scope of the present disclosure, are contemplated thereby, and are intended to be covered by the following claims, which should be given the benefit of all reasonable equivalents. |
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claims | 1. A system for accommodating a solid target in an accelerator, comprising:a rotating target changer having at least one feed through slot dimensioned for passage of the solid target;an insert having a cavity with a radial circumference, the cavity radial circumference in communication with the feed through slot and passing the solid target therethrough;a piston slidably received and retained within the insert cavity before the target is loaded into the insert;a cylinder coupled to the piston, which displaces the piston in one three positions within the cavity; anda bracket coupled to the insert, cylinder and target changer. 2. The system of claim 1, wherein the first position comprises the piston being in a load position. 3. The system of claim 1, wherein the second position comprises the piston being in an extended position. 4. The system of claim 1, wherein the third position comprises the piston being in an extracted position. 5. The system of claim 1, wherein the target changer has respective beam and service rotational positions. 6. The system of claim 1, wherein the target changer feed through slot is in communication with at least one insert slot defined within the cavity radial circumference which conveys the solid target between the insert cavity and feed through slot. 7. The system of claim 6, wherein the insert defines a second insert slot which conveys the solid target out of the insert cavity while the piston remains retained therein. 8. The system of claim 7, wherein the piston includes a tab translatable over the second insert slot which prevents passage of the solid target out of the insert cavity. 9. The system of claim 1, wherein the piston includes a face and an o-ring for abutment against the solid target. 10. The system of claim 1, wherein the insert includes an o-ring interposed between the cavity and piston. 11. A system for accommodating a solid target in an accelerator, comprising:a rotating target changer having at least one feed through slot dimensioned for passage of the solid target;an insert for passage of the solid target therethrough having:a cavity with a radial circumference and first and second axial openings;a piston slidably received within the first opening during target passage through the insert; andat least one insert slot defined within the cavity between the piston and second opening that is in communication with the feed through slot and receiving the solid target therein when the target changer is in a load position;a cylinder coupled to the piston, which displaces the piston within the cavity; anda bracket coupled to the insert, cylinder and target changer. 12. The system of claim 11, wherein the piston includes a tab translatable over a second insert slot which prevents passage of the solid target. |
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041749995 | abstract | Positioning means for locating inspection apparatus used to volumetrically examine a nuclear reactor vessel is disclosed. The positioning means is provided with a support ring having an annular key positioned longitudinally about its periphery. Three support legs are attached to the support ring by brackets adapted to fit the annular key. In addition, the support ring also carries three guide stud bushings which are movably mounted thereon by clamps adapted to engage the support ring key. Prior to lowering the inspection apparatus into the vessel, the guide stud bushings are each moved to a point of alignment with one of three guide studs extending upwardly from the vessel. After alignment has been verified, the guide stud bushings are clamped in position.. The inspection apparatus is now lowered towards its fully seated position within the vessel and is coarsely circumferentially positioned with respect thereto by the engagement of the guide studs within the guide stud bushings. A fine degree of circumferential positioning is achieved by providing a specially configured shoe for one of the support legs. With the core barrel internals in, the special shoe is adapted to key onto a core barrel pin the exact location of which is known. With the core barrel internals removed, the special shoe is adapted to place a locating key into a notch in a vessel flange, the location of which is also exactly known. Thus, as the inspection apparatus is lowered into its fully seated position, exact circumferential positioning thereof with respect to the vessel is achieved. The other support legs rest on an inner circumferential flange so that no portion of the inspection apparatus touches or threatens the vessel's top flange. |
claims | 1. A method of tuning an ion beam in an ion implanter, wherein the ion beam is dependent upon a plurality of operational parameters of the ion implanter, the method comprising:retrieving a set of operational parameters at least some of which are stored in a dynamic database;configuring the ion implanter according to the retrieved set of operational parameters thereby to provide an ion beam;optimizing the ion beam by varying one or more of the operational parameters; andupdating the operational parameters stored in the dynamic database that changed during optimization with rolling averages calculated using the values obtained for those operational parameters at the end of the optimization. 2. The method of claim 1, comprising updating the operational parameters only if a particular operational parameter changed within a predefined limit. 3. The method of claim 2, comprising updating the operational parameters only if a particular operational parameter changed by less than a predefined fraction. 4. The method of claim 1, comprising retrieving at least one of the set of operational parameters from those stored in a principal database. 5. The method of claim 4, wherein, if an operational parameter is stored in both the principal database and the dynamic database, the method comprises retrieving the operational parameter from the principal database in preference to the dynamic database. 6. The method of claim 4, comprising optimising the ion beam by varying one or more of the operational parameters and storing one or more operational parameters so established in the principal database. 7. The method of claim 4, wherein, even if an operational parameter that changed during optimisation was retrieved from the principal database, the method comprises updating that operational parameter stored in the dynamic database. 8. The method of claim 1, comprising retrieving a set of operational parameters associated with a particular implant recipe from a plurality of such sets associated with different implant recipes. 9. The method of claim 8 comprising, when an implant recipe does not have an associated set of operational parameters, retrieving a set of operational parameters stored in an initialisation database, configuring the ion implanter according to the retrieved set of operational parameters thereby to provide an ion beam, optimising the ion beam by varying one or more of the operational parameters, and storing the operational parameters in the dynamic database as determined during optimisation. 10. The method of claim 1, further comprising using an ion source of the ion implanter to generate the ion beam. 11. The method of claim 10, comprising retrieving a set of operational parameters from a plurality of such sets associated with different ages of the ion source. 12. The method of claim 11, comprising retrieving a set of operational parameters from a plurality of such sets associated with total ion beam current delivered by the ion source. 13. The method of claim 11, comprising retrieving a set of operational parameters from a plurality of such sets associated with the product of the age of the ion source and the total ion beam current delivered by the ion source. 14. The method of claim 11, further comprising resetting the operational parameters stored in the dynamic database after the ion source has been serviced or replaced. 15. The method of claim 11, further comprising keeping the operational parameters stored in the dynamic database after the ion source has been serviced or replaced, and subsequently retrieving a set of operational parameters from the plurality of such sets associated with different ages of the ion source according to the age of the ion source since servicing or replacement. 16. An ion implanter and a computer program comprising computer code instructions that, when executed, cause the ion implanter to operate in accordance with the method of claim 1. 17. A programmable controller arranged to retrieve from memory the computer program of claim 16 and to execute the computer program. 18. An ion implanter and a computer readable medium having recorded thereon a computer program that, when executed, causes the ion implanter to operate according to the method of claim 1. 19. An ion implanter comprising a controller configured to implement the method of claim 1. |
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047643323 | description | DETAILED DESCRIPTION The present pipe end sealing element is usable in sealing the end of a plain end pipe, i.e. without threading, and enables hydrotesting of the pipe. The device is especially useful in a method for hydrotesting of an open-end pipe that is situated in a nuclear reactor, so as to preclude the need for hydrotesting the entire reactor vessel, in order to determine the integrity of the open-end pipe. Referring now to FIG. 1, the pipe end sealing element 1 comprises a hollow tubular member 3, hollow conical bushing 5, a sealing plug 7 and a threaded bolt 9. The sealing element is usable to seal a pipe 11 having an outer wall 13, inner wall 15 and end wall 17 that surrounds an open end 19 of the pipe 11. The hollow tubular member 3 is of a size that the same will encircle the open end of the pipe 11, and has an inner diameter d slightly larger than the outer diameter d' of the pipe 11. The wall 21 of the inner tubular member 3, has at one end 23 an inwardly directed flange 25 that extends inwardly from the inner face 27 thereof. The inwardly directed flange 25 terminates at an end 29 which provides the distance d, between facing portions of the flange, and has an inner bevelled surface 31, the bevelled surface diverging from the end 29 towards the inner face 27 of the inner tubular element 3. Preferably, a shoulder 33 is provided about the inner face 27 where the diverging bevelled surface 31 terminates, spaced from the end 29 thereof. At the other end 35 of the hollow tubular member 3, a threaded portion 37 is provided on the inner face 27, having threads 39. An aperture 41 is provided through the wall 21 of the hollow threaded member for use in applying the sealing element to the open end of a pipe, as hereinafter described. The hollow conical bushing 5, of variable interior diameter, in the nature of a hollow inverted truncated cone, has a wall 43 that forms an apical end 45 and base end 47. At the apical end 45, the inner diameter d.sup.2 of the hollow conical bushing 5 is of a value between d and d', while the outer diameter d.sup.3 is of a value that is larger than d.sup.2, but still with a value between d and d'. The inner surface 49 of the hollow conical bushing 5 is annular, while the outer surface 51 has a bevelled portion 53, the bevel of surface 53 being complementary to the inner bevelled surface 31 of the hollow tubular member 3. A slot 55 is provided in the wall 43 of the hollow conical bushing, extending from the base end 47 to the apical end 45 thereof, such that upon external pressure being applied to the exterior of the bushing, the hollow conical bushing will decrease in interior diameter size. Sealing plug 7, has a cylindrical portion 57 and flange portion 59, flange portion 59 forming a shoulder 61. The cylindrical portion 57 has an outer diameter which is slightly smaller than the inner diameter d.sup.4 of the pipe 11, but has a close fit with the inner wall of the pipe 11 when inserted into the open end thereof. A groove 63 is formed in the outer wall 65 of the cylindrical portion 57 for engagement therein of an O-ring 67. The threaded bolt 9 has threads 69 thereon, which are engageable with the threads 39 on the threaded portion 37 of the hollow tubular member 3, and a head portion 71 for use in turning of the bolt. Preferably, a raised portion 73 is provided on the threaded bolt 9 on the end opposite the head portion 71. The engagement of the pipe end sealing element 1 with the end of pipe 11 is illustrated in FIG. 2, which illustrates the various components just prior to actual sealing of the open end 19 of pipe 11. In such a position, the hollow tubular member 3 is placed over the open end 19 of the pipe 11 and encircles the same, with the threaded portion 37 thereof adjacent and beyond the open end 19. The conical bushing 5 is next inserted into the hollow tubular member 3 and surrounds the outer wall 13 of the pipe 11, with the bevelled surface 53 on the outer surface 51 of the conical bushing in contact with the bevelled surface 31 on the flange 25 of the hollow tubular member 3. The sealing plug 7 is then placed on the open end 19 of the pipe 11 with the cylindrical portion 57 thereof inserted into the open end 19, and the shoulder 61 of flange 59 seated on the end wall 17 of the pipe 11. The threaded bolt is next engaged by threadedly engaging the threads 69 of the threaded bolt 9 with the threads 39 of the threaded portion 37 of the hollow tubular member 3. The threaded bolt 9 is advanced towards the sealing plug 7 by use of a wrench 75, while the hollow tubular member 3 is prevented from rotation by use of a stop member or rod 77 inserted into the aperture 41 thereof. In FIG. 2, the raised portion 73 of the threaded bolt 9 has not yet contacted the sealing plug 7. With continued threading, however, the threaded bolt will advance and contact the sealing plug 7, and force the flange 59 thereof into sealing relationship with the end wall 17 of the pipe 11. Upon continued rotation of the threaded bolt 9, the hollow tubular element 3 will be drawn, in an axial direction, towards the bolt 9, and the flange 25 of the hollow tubular element, along bevelled surface 31, will slide along the complementary bevelled surface 53 of the conical bushing 5, as indicated by the arrow in FIG. 3. This sliding movement will cause the conical bushing 5 to decrease in interior diameter and tightly grip the outer wall 13 of the pipe 11, while the sealing plug 7 is held firmly in sealing contact with the end wall 17 of the pipe 11. Upon such securement, as illustrated in FIG. 3, the open end 19 of pipe 11 is completely sealed. By torquing of the threaded bolt 9, the conical bushing 5 applies contact pressure on the pipe 11, so that a sealing relationship is achieved by friction between the conical bushing 5 and the pipe 11. The torque required will be dependent upon the hydrostatic pressure to be exerted on the pipe 11, and is calculated in order to prevent high external stresses on the pipe, as well as to have a sufficient margin below the load that would tend to separate the sealing device from its sealing relationship. The pipe end sealing element may be produced from various materials depending upon the application of its use. The components may, for example, be formed of carbon steel, stainless steel, other metals, or hard plastics, depending on the application, temperature and pressure to which they will be exposed. The O-ring is normally of a rubber or plastic material, but may be of a metallic substance where high temperatures are present. The pipe sealing element may, of course, be fabricated in various sizes so as to be usable with a variety of sizes of pipe. The pipe end sealing element is especially useful in the hydrotesting of pipes present in a nuclear reactor, such as instrument tubes in a pressurized water reactor. As illustrated in FIG. 6, a pressurized water reactor 81 typically includes a cylindrical pressure vessel 83 that comprises an outer pressure resistant wall 85, closed at the bottom by a bottom wall 87 of a hemispherical contour. The vessel is closed at the top by a flanged, dome-shaped head 89, which is secured, such as by bolts, to the top edge 91 of the pressure resistant wall 85, preferably seated in a channel 93 about the wall 85. The pressure resistant wall 85 has at least one inlet nozzle 95 and at least one outlet nozzle 97 distributed about its periphery, with a pair of each of such nozzles usually provided. A nuclear core 99 is supported in the lower region of the vessel 83, the core being supported in spaced relationship to the bottom wall 87 by a core barrel 101, the core barrel 101 having a flange 103 which rests on a ledge 105 in the inner surface of the pressure resistant wall 85. The core includes a series of fuel assemblies 107 and thimbles 109 for receiving control rods, not shown, with at least one such thimble 111 adapted for insertion therein of an instrument for monitoring the operation of the core. The fuel assemblies and thimbles are mounted between a lower core plate 113 and an upper core plate 115. The control rods, as is known, may contain rod clusters of high or low absorption cross-section for neutrons, and water displacement rod clusters, and serve to reduce the thermal power of the reactor, or otherwise control the same, through monitoring by use of the instrument in the dedicated thimble therefore, or to shutdown the reactor. In the upper region of vessel 83, the upper internals 117 may have an associated calandria structure 119. The upper internals 117 include vertical guides 121 for control rods and vertical guides 123 for water displacement rods, while the calandria structure 119 is disposed thererabove. The core 99, upper internals 117, and calandria structure 119 are mounted generally coaxially within the vessel 83. An annulus 125 between the core barrel 101 and the pressure resistant wall 85 provides for communication between the inlet nozzles 95 and the lower end of the core 99. Drive rods 127 from the control rods extend through head penetrating adaptors 129 in the dome-shaped head and then through the calandria structure 119. Drive rod mechanisms 131 are used to properly position the control rods. Coolant enters through inlet nozzles 95 and flows downwardly through annulus 125 to the bottom wall 87 and then upwardly through the core 99, upper internals 117 and into the calandria structure 119, from which it flows transversely to, and outwardly from, the outlet nozzles 97. The instrument leads to instruments in thimbles 111 extend through a plurality of adapters 133 in the bottom wall 87, through tubular elements or pipes 135 which cooperate with apertures (not shown) in the lower core plate 113, such that readings of the operation of the core may be made externally of the reactor vessel. When testing of pipes in a nuclear reactor for integrity, such as testing of the tubular elements or pipes 135 after the lower core plate has been removed, the method of the present invention may be used. The present pipe end sealing element is placed over the open end of the pipe to seal the same, as hereinbefore described, and water injected under pressure through the other end of the pipe to a predetermined pressure. If the pipe withstands the pressurized water, the pressure is then released, and the pipe end sealing element removed. The pipes are then suitable for intended use. |
summary | ||
summary | ||
description | Currently, various countries are enacting or enforcing regulations for restricting use of a specific constituent. Such the regulation include the RoHS (Restriction on Hazardous Substances) directive or the ELV directive (End of Life Vehicles Directive) in the EU. In this situation, management of constituents contained in a product is urgent necessity for product manufacturers. In particular, a method for certificating non-containment of a prohibited constituent has become an issue for each of the manufactures. Assemblers and the like generally do not manufacture components in their own factories and employ process of purchasing the components from component manufacturers and the like, and then assemblers assemble a product with the purchased components. In this case, for example, the assembler requests the suppliers to provide with constituent information of the components constituting the product so as to obtain constituent information of the product which is manufactured by the assembler. In response to such a requirement, the green procurement movement has recently been promoted in manufacturers so as to urge component suppliers (hereinafter, referred to simply as “suppliers”; including component venders and component manufacturers) to investigate and provide constituent information of components. The provided information is accumulated in a component information database of the product manufacturer. In order to confirm constituent information of components, a designer in the manufacturer or the like searches for information of a target component from the component information database to confirm the component information so as to decide on a component as a component for constituting a product. In many cases, however, not all the component information that the designer wants to know is provided by the suppliers. In addition, even if the component information is registered, the registered contents are insufficient, information quality differs between the suppliers, or some suppliers provide the component information without conducting full investigation thereof. Moreover, since each supplier has a different idea about constituent information, the quality of provided constituent information frequently varies depending on the suppliers which provide the constituent information. However, since there is no means of examining such component information quality at present, it is impossible to fully examine whether or not a target component conforms to regulations. Accordingly, the conformity of a component to environmental regulations (hereinafter, referred to simply as “conformity”) cannot be precisely evaluated. In view of such a situation, the assemblers try to ensure the credibility of information by a method of obliging the suppliers to submit non-containment certificates or the like. However, an amount of information available at early stages of design is limited in many cases. Therefore, the assemblers are required to evaluate the credibility of components based on a limited amount of constituent information provided by the suppliers at early stages of design to confirm the conformity of the components to the environmental regulations. As a technology of evaluating the credibility of components, Japanese Patent Laid-open Publication No. 2004-038656, “Product design support device and method”, proposes a product design support device which has functions of designating a range of a specification value to search for a component satisfying the range of the specification value as a recommended component and of accepting the registration made by a designer in terms of a recommended component to give the recommended component higher priority to be selected as a recommended component for the next component search. Japanese Patent Laid-open Publication No. H 11-238069, “Component selection device”, discloses a component selection device which has functions of searching a database for alternative components for a certain component designated by an item code and of deriving a recommendation ranking of the alternative components based on data such as reliability, cost, an inventory condition of the alternative components for the certain component. Japanese Patent Laid-open Publication No. 2002-049649, “Environment information simulation system, device, and method, and recorded medium”, describes an environment information simulation system which has the function of evaluating a conformity status of a certain component, even if whose constituent information has not been registered to environment regulations, by using a weighted average value of constituent information of similar components as constituent information of the certain component. Japanese Patent Laid-open Publication No. 2003-203105, “Product check system, information setting check device, product-related information providing device, and product check support device”, discloses a product check system and the like having the following function. In the product check system, a database is created which includes information on components most likely to be contained by each type of component. Upon input of a constituent that should be reported, a supplier also inputs the type of the component. With respect to the type of the component thus input, a constituent that is highly likely to be contained in the component is extracted from the database to be compared with the input constituent content. When no constituent corresponding to the report is extracted from the database, a message is sent to the reporter so as to prompt reconfirmation. Throughout this specification, the term “component” denotes both a “part” which constitutes a product and demonstrates a part of a function of the product and a “material” which signifies a raw material necessary for manufacturing the part. The term “constituent” denotes a physical constituent element necessary for producing the material. For example, a chemical substance is one of the constituents. The systems and devices described in the above-cited documents are for estimating component information of a target component based on information of components similar to the target component (Japanese Patent Laid-open Publication No. 2002-049649) or extracting a recommended component from the specifications designated by a designer (Japanese Patent Laid-open Publication No. 2004-038656 and Japanese Patent Laid-open Publication No. H 11-238069). However, the systems and devices cannot evaluate the accuracy of the similar component information, that is, the accuracy of the component information provided by the supplier to output the result of evaluation. Moreover, in the case where the constituent information of the component information to be evaluated is estimated using the similar component information to verify the conformity of the target component, there is no means of verifying whether the result of evaluation of the conformity is credible or not. Therefore, even if the result of evaluation of the conformity is obtained, it is sometimes difficult to perfectly avoid the risk of inconformity to regulations, because the credibility of the result of evaluation is not definite. Furthermore, even if the result of evaluation of the conformity of the target component is obtained, there is no means of ensuring that the target component can avoid the risk of inconformity to regulations. Moreover, in the above-cited related art, if the target component turns out not to conform to the regulations, the designer is required to select a target component again. Therefore, a large number of steps are needed for the designer to select a component, which impairs efficiency. It is therefore an object of the present invention to provide a method for supporting a designer in confirming constituent information of a component at early stages of design and a device capable of evaluating constituent information of a target component and conformity of the target component to environmental regulations by using the method. In order to attain the above object, the present invention provides a system for evaluating constituent information, which evaluates credibility of an evaluation result of conformity of a component to environmental regulations, the system including: an input unit; an output unit; a processing unit; and a memory unit, in which the memory unit includes: a component basic information table which retains identification information that distinguishes each item; a constituent information table which retains constituents contained in the component and a content of each of the constituents; and an environmental qualification information table which retains constituents contained in the component for each item and a evaluation result of conformity of the component to environmental regulations, and in which the processing unit includes: means which accepts selection of a target component through the input unit; means which calculates constituents contained in the target component and a content of each of the constituents by referring to the constituent information table based on identification information of the target component and evaluates conformity of the constituent to environmental regulations based on the content; means which extracts a component similar to the target component by referring to the environmental qualification information table and calculates constituents contained in the similar component and a content of each of the constituents; and means which evaluates credibility of the evaluation of the conformity to environmental regulations based on a difference between the content of the constituent in the target component and the content of the constituent in the similar components, the constituent being used for the evaluation of the conformity to environmental regulations among the constituents contained in the target component. Furthermore, the present invention provides a method and a program for evaluating reliability on constituent information used for the system for evaluating reliability on constituent information. According to the present invention, the processing unit may further include: means which accepts selection of a status re-investigation, adoption and non-adoption, of the target component, through the input unit; means which makes a re-investigation request attached with the result of investigation of the target component to a manufacturer of the component when the re-investigation is selected as the status; means which makes a qualification request to a component final confirmation personnel when adoption is selected as the status; and means which returns to the input unit of basic information of the target component when non-adoption is selected as the status. Specifically, according to the present invention, conformity of the evaluated target component, credibility of the component information, and a result of verification of the target component by the designer are stored as a part of the component information. From then on, when information of the target component is found in the search, the result of verification is output. Moreover, for the target component not qualified by the designer as being not conforming to the environmental regulations, the supplier is requested to provide information again so as to assure the conformity. As a result, it is possible to support the operations regarding component information. Also, according to the present invention, the component similar to the target component can be extracted by referring to the component basic information table and by searching for the similar component based, on the identification information of the target component according to a definition of the similar component. Also, it is desirable to calculate each of the constituents contained in the target component and the content of the constituent based on constituent information of the component similar to target component, which is obtained from the constituent information table, the constituent information being weighted based on environmental qualification information of the similar component obtained from the environmental qualification information table. Further, according to the present invention, it is possible to provide the processing unit with: means which searches for a component similar to the target component, obtains constituent information of the target component and the similar component from any one of the constituent information table and the result of calculation of the constituent information, and estimates the environmental qualification information of the target component by using at least a part of the obtained constituent information as an explaining variable and environmental qualification information of the target component and the similar component extracted from the environmental qualification information table as an explained variable; and means which calculates an accuracy of the estimation. Also, it is possible to further provide the processing unit with: means which searches for a component similar to the target component and calculates a trust of each of suppliers based on at least a part of data quality of a component provided by each of the suppliers and information about quality of the component, for the suppliers of the target component and the similar components; and means which determines recommendation ranking of the target component and the similar components from the result of calculation of the trust of each of the suppliers and the result of estimation of the environmental qualification information of the target component and the similar components. In this manner, based on the conformity of the component information of the target component evaluated by the present invention, a registration rate of component information provided by the supplier, and a conformity rate of the component provided by the supplier, recommendation ranking of the target component and the similar components are calculated to thereby support the designer in selecting a component. According to the present invention, the designer can search for candidate components based on abstract information such as a specification value of the component, the designer can evaluate the conformity of the candidate components to environmental regulations even if an amount of information provided by the supplier is small. Moreover, according to the present invention, the recommendation ranking of a plurality of candidate components is clearly defined, which allows the designer to easily select a component to be used. Furthermore, according to the present invention, the operation of requesting the supplier to collect information of the candidate components or the operation of requesting a quality assurance monitor to qualify the component can be provided seamlessly. Hereinafter, an embodiment of the present invention will be described. However, the present invention is not limited thereto. In this embodiment, the details of an evaluation system of credibility of constituent information, which supports a designer in evaluating the conformity of a target component to select a component having the least risk of inconformity to environmental regulations, will be described. (1) Overall Configuration A hardware configuration of a constituent information credibility evaluation system 100 according to this embodiment is shown in FIG. 17, while its functional configuration is shown in FIG. 1. The system 100 according to this embodiment is, as shown in FIG. 17, an information processing device (computer) including a main memory unit 1701, a central processing unit (processing unit) 1702, an external storage unit (storage unit) 1703, and an input/output device (an input unit and an output unit) 1704. The external storage unit 1703 retains a component information database 112 and a material information database 113. Each of means 101 to 107 constituting the system 100 is realized by the execution of a program by the central processing unit 1702, the program being retained in the external storage unit 1703 and being read by the main memory unit 1701. The result of execution of the program by the central processing unit 1702 is stored in the external storage unit 1703 or is output through the input/output device 1704. Although each of the means 101 to 107 is realized by the execution of software with such a general information processing device in this embodiment, the present invention is not limited thereto. Each of the means 101 to 107 may be realized by the combination of special-purpose hardware or the like. A designer of a product wants to obtain candidate components having a low environmental load for the selection of a component to be used for the product. The designer inputs a specification value, a component number unique to a component, and the like as identification information of the component to the system 100 to search through the component information database 112. When selecting a target component from the components found in the search in a step 102, the designer is required to check constituent information of the components so as to evaluate the conformity to environmental regulations. However, the component information database does not always contain constituent information. In some cases, constituent information is not even provided by a supplier. Therefore, the constituent information credibility evaluation system 100 in this embodiment refers to material composition information of a target component and the material information database 113 or refers to constituent information of components similar to the target component to estimate the constituent information of the target component in a step 103 so as to evaluate the conformity of the target component in a step 104. Moreover, if constituent information is not approved by anyone even though the constituent information of the target component is provided by a supplier, it is unclear whether the constituent information is credible or not. Therefore, the constituent information credibility evaluation system 100 according to this embodiment compares the constituent information of the target component with the constituent information of similar components that have already been qualified among the components similar to the target component to estimate the credibility of the constituent information of the target component, and outputs the result of evaluation. The designer can confirm the result of evaluation in a step 105 to determine whether or not to use the target component. If the designer selects re-investigation of the component information, the constituent information credibility evaluation system 100 accepts the selection and transmits a request for re-investigation attached with the result of evaluation (step 108) to the supplier in a step 106. If the designer selects the adoption of the target component, the constituent information credibility evaluation system 100 transmits a request for component qualification (step 110) to a quality assurance monitor in a step 107. If the designer selects non-adoption of the component, the constituent information credibility evaluation system 100 accepts the selection, searches for a component again in a step 114 and then returns the processing to the step 102. In the case of the re-investigation (step 106), the component information database 112 accepts and stores component information 109 corresponding to the result of re-investigation by the supplier. In the case of the component adoption (step 107), the component information database 112 accepts and stores information containing the result of component qualification by the quality assurance monitor. The component information 109 and qualification information 111 are in cooperation with the constituent information credibility evaluation system 100 so as to be used by the designer in real time for the component search. (2) Search for Candidate Components First, means which searches for candidate components will be described. The system 100 first accepts identification information of a component to be searched for. The identification information accepted by the system 100 is exclusively an item code at the supplier. However, information to be input may be a specification value of a target component, an item code of each of the components constituting the target component, or a general appellation of the component. Herein, the term “specification value” denotes a value quantitatively representing the performance of a component. The “specification value” corresponds to, in the case of a capacitor, for example, a capacitance, a size, a temperature characteristic, or the like. The term “item code” denotes an identification number unique to a component corresponding to a component name officially announced by a supplier or an identification number unique to the component described in drawings and the like by a designer. Upon input of identification information, the system 100 searches the component information database 112 for a corresponding component based on the identification information. When a search key is a supplier name, an item name, or the like, the identification information is searched for by perfect match or partial match of the supplier name or the item name. When a search key is composed of symbols as in the case of a supplier code or an item code, a so-called “fuzzy search” for a search with spaces ignored or a case-insensitive search is implemented (a screen image at this time is shown in FIG. 2). Next, the system 100 outputs a list of components found in the search to a search result screen. At this time, in addition to output on the screen, the list of the components may also be output in the form of another file such as that of spreadsheet software. On the search result screen, the system 100 accepts the selection of a target component whose conformity and credibility of the constituent information are desired to be evaluated from the plurality of components found in the search. On the other hand, the component information database is searched for similar components mainly based on information that does not specify components such as a specification value of a component and a component code. A search result screen image of the target components and the similar components is shown in FIG. 3. (3) Evaluation of Conformity to Environmental Regulations Next, means which evaluates the conformity of the target component to environmental regulations will be described. The component information database includes at least a constituent information table which retains constituent information of each of the components and a material composition information database. For example, if a chemical substance is used as a constituent, the constituent information herein is associated with a chemical substance code, a chemical substance name, or a content mass or ratio of the chemical substance contained in the component with the use of an item code of the component as a key. The chemical substance code is an identification number which serves to specify a chemical substance, and a generally-used code can be used as the chemical substance code. In this embodiment, a CAS Registry Number defined by American Chemical Society is used. The material composition information is associated with a material code, a material name, or a material mass or ratio of the material constituting the component with the use of an item code of the component as a key. The material code is an identification number for specifying the material, and a generally-used material code can be used as the material code. In this embodiment, a JIS code is used as the material code. As the first evaluation means of the conformity, means which evaluates the conformity of the target component in the case where constituent information of the target component is provided by a supplier will be described with reference to FIG. 4. The system 100 searches a component information table 407 constituting the component information database, for information of a target component or similar components. If material composition information is that of the component which has already been registered, constituent information 530 is estimated based on material composition information 510 of the target component or the similar components by referring to the material information database 520. The material information table 408 constituting the material information database is registered with content ratio information of the material for the material code. An example of the content of the material information database is shown in FIG. 6. A content ratio of each constituent may be fixed or may be the range defined by a lower limit and an upper limit as shown in FIG. 6. The system 100 uses a material code 512 of the material composition information associated with the item code 511 as a key to search through the material information database. Then, the system 100 relates a constituent code 522 and a content ratio 523 of the constituent with the item code 521, and multiplies a material mass constituting the target component or the similar component with a content ratio of each material to calculate (estimate) each constituent and its content in constituent information 533 of the target component. The result of estimation of the constituent information is used for comparing the estimated constituent information with the constituent information provided by the supplier in a step 406 to examine the credibility of the component information provided by the supplier when the constituent information of the target component is independently provided by the supplier. The result of estimation of the constituent information is also used for comparing the result of estimation of the constituent information with information of regulations in a step 405 to evaluate the conformity of the constituent to environmental regulations. In this case, it is not necessary to consider whether the constituent information is provided by the supplier or not. The effect of this means is that the material composition information can be converted into the constituent information to provide the means which examines the constituent information to support the designer in rapidly adopting a component. Next, means which estimates constituent information from a similar component in the same classification of the target component to evaluate the conformity of the target component even if the constituent information of the target component is not provided will be described (FIG. 7). In this case, the constituent information or the material composition information of the target component is not required. First, the system 100 accepts the input of the component information by the designer in a step 702. Then, in a step 703, the system 100 searches for components similar to the target component by the above-described similar component search method to find constituent information of the similar components from a constituent information table 708 constituting the component information database 112 in a step 704 and to temporarily store the found constituent information in the system. Subsequently, in a step 705, the system 100 searches an environmental qualification information table 709 constituting the component information database 112, for environmental qualification information of the similar components whose constituent information are found, and in step 706, estimates the constituent information of the target component based on the result of the search. An example of the estimating method will be described below. As shown in FIG. 8, a plurality of qualification information are present for the same component. Therefore, the number of records with the result of investigation being not “not investigated” is set as the number of establishments that have investigated the component (in the example shown in FIG. 8, three establishments) and is reflected on weighting of the constituent information of the similar component. Specifically, if the number of similar components is i, a content mass of a constituent j in the similar component i is qij, and the number of establishments that have investigated the similar component i is xi, a content mass Qj of the constituent j contained in the target component is calculated as the following formula (Formula 1). Q j = ∑ i q ij ( x i + 1 ) ∑ i ( x i + 1 ) ( Formula 1 ) An example of the result obtained by estimating the constituent information of the target component by the following formula based on the data shown in FIG. 8 is shown in FIG. 9. A pass rate Rp of an estimated value of a content mass of a constituent of the target component is evaluated as follows. By using the number xi of establishments that have already investigated the similar components i and a total number yi of the similar components with the result of qualification “pass”, the pass rate Rp is obtained by the following formula (Formula 2). R p = ∑ i y i ∑ i x i ( Formula 2 ) The effect of this means is that the designer can evaluate the risk of inconformity to environmental regulations from the specifications of the component without waiting for the provision of information by the supplier. As a result, the selection of a component at the early stage of design can be facilitated. (4) Evaluation of Environmental Qualification Information Next, means which extracts the constituent information of the target component and the similar components to evaluate the environmental qualification information of the target component will be described with reference to FIG. 10. It is desirable that the constituent information of the target component be originally registered by the supplier, but the constituent information of the target component may be obtained by estimation based on the material composition information or the constituent information of the similar components by the above-described means. In this embodiment, substance group information is used as the constituent information. The substance group information will now be briefly described. The substance group information is a kind of constituent information. However, the constituent information is data of a content mass of each specific constituent, while the substance information is containment information for each category of the constituents, determined for each constituent element (for example, lead and a lead monoxide both fall in the category of lead and its chemical substances). Examples of a method which creates the categories include a method which classifies constituents containing a specific element in the same category as in the case of “lead and its chemical substances”, a method which classifies constituents based on the property of the constituents as in the case of “ozone-depleting substances”, a method which classifies constituents based on regulations as in the case of “substances to be restricted by the EU RoHS directive”. Although the containment information is normally in the form of content mass or rate, the containment information may also be discontinuous information such as the existence of a substance. A flow of the process of evaluating the environmental qualification information in the constituent information credibility evaluation system 100 will be described with reference to FIG. 10. First, the system 100 searches for components similar to the target component. Since the means which searches for the similar components has already been described, the description thereof is herein omitted. Next, in a step 1004, the system 100 obtains the substance group information of the similar components from a substance group information table 1011 constituting the component information database 112. The substance group information used herein is information which can have three variables, i.e., “contained”, “not contained”, and “under investigation” for the existence of a certain substance group. Next, in a step 1005, the system 100 extracts ×(×is an arbitrary number) substance groups in the order from the smallest ratio of “under investigation” except prohibited substance groups and stores substance group information of the extracted target component and similar components in the system. Furthermore, in a step 1006, the system 100 obtains the environmental qualification information of the similar components (in this case, the environmental qualification information can have either “pass” or “fail” as a value, but the environmental qualification information may also have a value that does not correspond to either of the values, for example, “limited pass”). Furthermore, the system 100 obtains the substance group information of the target component (step 1009). Examples of a method which obtains the substance group information include a method which extracts the substance group information provided by the supplier from the substance group information table 1011 and a method which creates the substance group information based on the constituent information estimated in FIGS. 4 and 7. When information as shown in FIG. 11 is collected by the above-described process, the system 100 obtains an estimate value of the environmental qualification information of the target component (step 1007). As an example of a method which estimates the environmental qualification information used herein, there is analysis using the quantification method II. According to this quantification method II, if an explaining variable (the substance group information in this case) and an explained variable (the environmental qualification information in this case) are both qualitative variables, the environmental qualification information of the target component is estimated based on the substance group information of the similar components and the substance group information of the similar components and the target component. The detailed description of the calculation method is herein omitted. Subsequently, the system 100 obtains the accuracy of the thus obtained environmental qualification information of the target component by a calculation of a discriminant efficiency according to the quantification method II (step 1008). The accuracy is a value indicating the credibility of the result of evaluation. The effect of this means is that an indefinitely large number of constituents are classified into categories to limit the number of types of constituent information. As a result, a load on an information provider is reduced to further facilitate the analysis. The constituent information is evaluated based on a broader concept of the constituents such as the substance group or discontinuous data such as the existence of a substance is used to allow the designer to evaluate the conformity to environmental regulations even if the range of information provided by the supplier is small. As a result, a load on the supplier is reduced to allow the designer to avoid the risk of inconformity to environmental regulations at the early stage of design. (5) Evaluation of a Possibility of Qualification of Constituent Information Next, means which extracts the substance group information of the target component and the qualified similar component and calculates a similarity of the similar component to the target component to evaluate a possibility of qualification of the constituent information of the target component will be described (FIG. 12). For a similarity of constituent information between components, there is a method which compares content masses of a specific component. In many cases, however, a function inherent in the component is not only a characteristic of the component itself but also that of the elements constituting the component. Therefore, it is desirable to collectively examine the similarity for each substance group. A process in the means which searches for the components similar to the target component is as described above. At this time, since the similar components used for the evaluation of a similarity are limited to those qualified, the system 100 in this embodiment searches for an environmental qualification information table 709 in a step 1204 subsequent to the search for the similar components. Next, the system 100 extracts ×(×is an arbitrary number) substance groups having a small ratio of “under investigation” except the prohibited substance groups (step 1205) and calculates a similarity from the substance group information of the target component and the similar components (step 1206). The system 100 regards the similarity obtained in this step as the credibility of the constituent information of the target component. The system 100 outputs the thus obtained similarity through the input/output device 1704 so as to provide the similarity to a user as a reference value for selection of a component. An example of the method which calculates the similarity will be described. Assuming that the extracted substance group is represented by i (i is a natural number from 1 to ×), a possibility that a qualified component contains the substance group is Pi, a value indicating the existence of the substance group in the target component (for example, a value quantified in 1 in the case where the target component contains the substance group, 0.5 under investigation, and 0 in the case where the target component does not contain the substance group) is yi, a similarity R of the target component is obtained by the following formula (Formula 3). R = ∏ i = 1 x P i - y i ( Formula 3 ) Taking the case shown in FIG. 13 as an example, the similarity is 64.8% for a component A and is 16.2% for a component B. The effect of this means is as follows. Even if data is not provided by the supplier or the amount of data provided by the supplier is small, the possibility that the target component is qualified is estimated based on the qualification information of the similar component. As a result, the designer can avoid the risk of inconformity to environmental regulations at the early stage of design. (6) Calculation of Recommendation Ranking Means which calculates recommendation ranking of the target component and the similar components based on the pass rate of the constituent information of the target component, a registration rate of the component information provided by the supplier, and the similarity of the component provided by the supplier to the qualified component to support the designer's selection of a component will be described with reference to FIG. 15. Although the similarity denotes that obtained by the process shown in FIG. 12, any value representing the possibility that the component is qualified such as the credibility obtained in FIG. 10 may be used. The system 100 in this embodiment searches for components similar to the target component in a step 1401. The system 100 refers to a component basic information table 1410 to search for similar components at each supplier (step 1402). Then, the system 100 obtains a pass rate at each supplier in a step 1403 and obtains a registration rate in a step 1404 and obtains a multiple of the pass rate and the registration rate as a trust of each supplier (step 1405). An example of calculation of the trust of a supplier A is shown in FIG. 15. Furthermore, the system 100 calculates the number of qualifying establishments, a pass rate, and a similarity for each component (steps 1406 to 1408). The number of qualifying establishments is the number of establishments which have qualified the component, and the pass rate is a pass rate on the results of qualification (passing statuses such as pass or fail; a status that is not “not investigated”). An example of the calculation of the number of qualifying establishments and the pass rate is shown in FIG. 8. The similarity is obtained by the above-described method shown in FIG. 12. In a step 1409, recommendation ranking is determined for the target component and the similar components. An example of the method which calculates the recommendation ranking will be described below. The number of qualifying establishments, a pass rate, a credibility, and a trust are provided for each component. In determining the recommendation ranking, the first priority is given to the components with a larger number of qualifying establishments. If the recommendation ranking cannot be determined based on the number of qualifying establishments, the recommendation ranking is determined based on the pass rate, then, based on the credibility, and finally based on the trust. An example of the calculation is shown in FIG. 16. The effect of this means is that the recommendation ranking is determined for the components to facilitate the designer's selection of a component. (7) Output The constituent information credibility evaluation system 100 outputs the result of evaluation of the conformity/credibility of the target component obtained by the processing in the step 104 through the input/output device 1704. According to this embodiment, the designer can use the result of output as a reference value for selecting a component. Next, the system 100 accepts the input of the content of confirmation of the result of evaluation by the designer (step 105). In accordance with the content of confirmation, the system 100 requests for re-investigation of the target component (step 106), determines not to adopt the target component and makes a new search (step 114), or determines to adopt the target component and transmits a component qualification request (step 107). |
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047909765 | claims | 1. In combination, a reactor pressure vessel of a boiling water reactor wherein reactor water is received, and a device for flushing out radioactive deposits accumulated in a lance-housing tube at least partly disposed in the reactor pressure vessel and for aligning in the lance-housing tube a dry power distribution detector lance which partly protrudes with a pressure-tight lance passthrough from an end flange on the lance-housing tube of the reactor pressure vessel and is supported in a seat on the end flange, comprising a tubular housing surrounding from below a part of the lance protruding from the reactor pressure vessel, said tubular housing being sealed by a lance protection tube and being fastened to the end flange; and a piston arranged in said tubular housing underneath the lance sealed in said tubular housing, and being vertically displaceable and rotatable so as to entrain and lift the lance out of said seat in which it is supported so that reactor water received in the reactor pressure vessel is admitted to the lance-housing tube, said piston being formed with a flow passage for the reactor water admitted to the lance-housing tube and including a valve disposed in said flow passage for controlling the flow of reactor water out of the reactor pressure vessel. 2. Device according to claim 1 wherein said piston is formed with an entrainer pin cooperatively engaging in a slot formed in said lance protection tube. 3. Device according to claim 1 wherein said lance protection tube is formed at an upper end in the interior thereof with a spherical member engageable in a marker slot formed in the lance passthrough. 4. Device according to claim 1 including a cap screw supporting said piston and cooperatively secured with a thread formed on an outside surface of said tubular housing. 5. Device according to claim 1 including a folding lever disposed on said piston. |
description | This is a continuation application of application Ser. No. 12/659,751, filed Mar. 19, 2010 now U.S. Pat. No. 8,039,796, which is, in turn, a continuation application of U.S. patent application Ser. No. 11/717,201, filed Mar. 13, 2007 (now U.S. Pat. No. 7,741,602), and claims priority of German patent application No. 10 2006 011 615.1, filed Mar. 14, 2006, the entire contents of which are incorporated herein by reference. The invention relates to a phase contrast electron microscope and especially a phase contrast transmission electron microscope. Electron microscope specimens are, as a rule, phase objects which generate only a very slight amplitude contrast in a transmission electron microscope because of the high electron energy in the range of 100 keV and higher. For this reason, in a conventional transmission electron microscope, objects of this kind generate a contrast only when utilizing the phase-shifting effect of the spherical aberration of the transmission electron microscope and are therefore imaged with correspondingly little contrast in a conventional transmission electron microscope. The introduction of a phase plate into the back focal plane of the objective of the transmission electron microscope therefore provides a large increase in contrast in a manner similar to the generation of a phase contrast in phase objects according to Zernicke in the optical microscopy. However, the dimensions required in a transmission electron microscope are problematic. Especially when the so-called unscattered ray (that is, only the ray which is undiffracted at the specimen) is to be shifted in phase, but the ray, which is diffracted at the specimen into the first order or higher orders, is intended to be uninfluenced by the phase plate, the small diameter of the unscattered ray of less than 1 μm imposes considerable requirements on the technology because, for the phase plate, freedom from contamination, freedom of charging and dielectric strength are required. For generating the phase-shifting effect, basically two starting points are known, namely: for the first starting point, the phase plate is realized as a correspondingly small configured electrostatic lens which imparts a phase shift only to the unscattered ray but leaves the higher diffracting orders entirely or substantially uninfluenced. For the second starting point, a thin foil is used which is substantially transparent for electrons of the electron energy used in the transmission electron microscope and which has the required structure. For the second starting point, the inherent electrostatic potentials of the material are used in order to impart the desired phase shift onto the unscattered ray or onto the scattered electron rays. The first starting point has the disadvantage that the small electrostatic lens perforce requires outer holding structures which interrupt regions wherein the paths of the electrons run which are diffracted into high diffraction orders whereby important information is lost for the generation of images. With respect to the latter, the second starting point has the disadvantage that the higher orders of diffraction, which are anyway very weak compared to the unscattered ray, are additionally weakened by the unavoidable material absorption of the foil. Because of these technological problems, no phase contrast electron microscopes with phase-shifting elements directly in the back focal plane of the objective lens could, up to now, be successfully established in the marketplace as commercial products even though the basics for the generation of phase contrast have been known for more than fifty years. Phase contrast electron microscopes are described in U.S. Pat. Nos. 6,744,048 and 6,797,956 which are incorporated herein by reference. In U.S. Pat. No. 6,744,048, the suggestion is made to image the back focal plane of the objective by a lens system and to arrange the phase-shifting element in the image plane of the diffraction plane of the objective with the image plane being generated by the lens system. It is an object of the invention to provide a phase contrast electron microscope wherein the dimensional requirements on the phase-shifting element are reduced. Another object of the invention is to avoid information loss in a phase contrast electron microscope. The first object is achieved with a phase contrast electron microscope which includes: an objective mounted on the optical axis and defining a back focal plane; a first diffraction lens for imaging the back focal plane into a diffraction intermediate image plane; a second diffraction lens having a principal plane disposed in the vicinity of the diffraction intermediate image plane; and, a phase-shifting element mounted in or near the diffraction intermediate image plane. The second object is realized with a phase contrast electron microscope including: an objective mounted on the optical axis and defining a back focal plane; a diffraction lens for imaging the back focal plane into a diffraction intermediate image plane; a first phase-shifting element mounted in the back focal plane; and, a second phase-shifting element mounted in or near the diffraction intermediate image plane. According to a first aspect of the invention, a phase contrast electron microscope has an objective having a back focal plane and a first diffraction lens which images the back focal plane of the objective magnified into a diffraction intermediate image plane. In the propagation direction of the electrons, a second diffraction lens follows the first diffraction lens and has a principal plane which is arranged in the vicinity of the diffraction intermediate image plane. A phase-shifting element is mounted in or near the diffraction intermediate image plane. According to a second aspect of the invention, a phase contrast electron microscope has an objective having a back focal plane and a first diffraction lens which images the back focal plane of the objective into a diffraction intermediate image plane. A first phase-shifting element is mounted in or near the back focal plane of the objective and a second phase-shifting element is mounted in or near the diffraction intermediate image plane. The imaging of the back focal plane is effected with the first diffraction lens and is imaged, preferably magnified, into the diffraction intermediate image plane. The diffraction plane of the objective is imaged magnified into a diffraction intermediate image plane. For this reason, the phase-shifting element can be configured to be correspondingly geometrically larger so that the dimensional requirements imposed on the phase-shifting element are correspondingly reduced. At the same time, the blanking out of electrons with the hardware of the phase plate, especially electrodes and support elements, is reduced because the dimensions of these parts can remain the same. Furthermore, it can simultaneously be ensured that no deterioration of resolution occurs because of the aberrations of the additional diffraction lenses and the image of the diffraction plane behind the system of the diffraction lenses can lie clearly rearward of the image planes which are conjugated to the object plane of the objective. Because, according to the first aspect of the invention, the principal plane of the second diffraction lens lies in or near the diffraction intermediate image plane, the second diffraction lens has an almost exclusive influence on the position of the image plane, which is conjugated to the object plane of the objective, and virtually no influence on the additional imaging of the diffraction intermediate image plane. The required structural length increase can thereby be kept low. The transmission electron microscope of FIG. 1 includes an electron source 1, for example, a thermal field emission source. An extraction electrode 2 follows the electron source 1 and this extraction electrode has a potential which draws electrons from the electron source 1. The extraction electrode 2 is followed by one or several focusing electrodes 3 in order to optically fix the location of the source position and one or several anodes 4. Because of the potential of the anode 4, the electrons, which emanate from the electron source 1, are accelerated to the desired electron energy of 100 keV or more. A multi-stage condenser follows the anode in the direction of movement of the electrons. In the embodiment shown, the condenser has three individual magnet lenses (5, 6, 7) and the entrance end part of the condenser-objective single-field lens 8. With a condenser of this kind, the illumination aperture as well as the field of the object plane 9 can be independently adjusted by a corresponding adjustment of the lens currents of the magnetic lenses (5, 6, 7, 8). The field of the object plane 9 is the field illuminated by the electron beam and the object plane 9 lies in the principal plane of the condenser-objective single-field lens 8. U.S. Pat. No. 5,013,913 is incorporated herein by reference with respect to the beam guidance in the condenser for different illumination fields and apertures. In lieu of a four-stage condenser, a simple condenser can, however, be provided as shown, for example, in U.S. Pat. No. 6,531,698 incorporated herein by reference. In the condenser-objective single-field lens 8, the prefield functions as a last condenser lens and the back field functions as an objective lens. The object plane lies approximately at the elevation of the pole piece gap of the condenser-objective single-field lens 8. A specimen manipulator (not shown) in the form of a goniometer is mounted in the object plane 9. The specimen manipulator is guided through the pole piece of the condenser-objective single-field lens 8. In lieu of the condenser-objective single-field lens, also other objective lenses can be provided. A first diffraction lens 11 follows the condenser-objective single-field lens 8 and is likewise configured as a magnet lens. This first diffraction lens 11 images the back focal plane or diffraction plane 10 of the condenser-objective single-field lens 8 magnified into a diffraction intermediate image plane 21. At the same time, the first diffraction lens 11 generates a real intermediate image 14 of the object plane 9. A first deflection system 12 is mounted in the plane of the intermediate image 14 generated by the first diffraction lens 11. A second deflection system 13 follows this first deflection system 12 and thereafter, a second diffraction lens 15. The diffraction intermediate image plane 21 then lies simultaneously in or near the principal plane of the second diffraction lens 15. A phase plate 16 is mounted in the diffraction intermediate image plane 21. The image of the diffraction plane 10 of the objective lens or of the imaging part of the field of the condenser-objective single-field lens 8 is magnified because of the imaging scale of the image generated by the first diffraction lens 11. For this reason, the diameter of the phase plate 16 is magnified by this imaging scale relative to an arrangement of the phase plate 16 directly in the back focal plane 10 of the objective lens. The other dimensions of the phase plate 16, such as the width of the holding elements (22, 27) and the radial width (difference between inner and outer diameters) of the annularly-shaped electrode can, in contrast, stay dimensioned the same so that the blanking of electrons by these components is overall reduced. The magnification with which the diffraction plane 21 is imaged into the diffraction intermediate image plane should be greater by a factor of two, preferably by a factor of three. The magnification with which the diffraction plane 21 is imaged into the diffraction intermediate image plane should not exceed a tenfold magnification and should preferably lie between a threefold magnification and a sevenfold magnification. The second diffraction lens 15 is so mounted that its principal plane lies in or near the diffraction intermediate image plane. For this reason, the second diffraction lens 15 has no influence or no significant influence on the further imaging of the diffraction intermediate image plane 21. This second diffraction lens 15 primarily images the intermediate image 14 of the object plane 9 into the entrance image plane 17 of the projective system (18, 19). In this way, a suitably long distance of the next-following image planes of the object plane 9 and the diffraction plane 10 is ensured. The projective system (18, 19) then generates a greatly magnified image on a detector 20 of the specimen arranged in the object plane 9 and imaged into the entrance image plane 17 of the projective system (18, 19). In FIG. 1, the reference characters OA identify the optical axis of the entire electron optical system. The lateral position of the image of the diffraction plane 10 of the objective can be finely adjusted relative to the phase plate 16 with the aid of the deflection systems (12, 13) between the first diffraction lens 11 and the second diffraction lens 15. The first deflection system is mounted or operates in the plane in which the intermediate image 14 of the object plane arises. For this reason, a deflection, which is effected by this first deflection system, has no influence on the lateral position of the image of the object plane which arises in the entrance image plane 17 of the projective system; rather, the first deflection system only shifts the position of the diffraction image relative to the phase plate 16. The imaging characteristics described above between the objective lens 8 and the first projective lens 18 are shown in FIG. 2 with respect to the beam paths. The solid lines show the beam paths for the electrons which emanate parallel to the optical axis OA from the object plane 9 (the so-called illumination beam path or pupil beam path) and the broken lines show the beam paths for the electrons which leave from the object plane 9 on the optical axis OA at an angle to the optical axis OA (the so-called imaging beam path). An electrostatic phase plate 16 is shown in plan view in FIG. 3. The phase plate 16 has an outer carrier 21 having a round aperture 26 with the carrier 21 being opaque to electrons. An annular electrode 23 having a central annular opening 24 is mounted in the aperture 26 centered on the optical axis OA. The annularly-shaped electrode 23 is connected to the carrier 21 via two or three thin holding elements (22, 27). By applying an electrostatic potential to the annularly-shaped electrode, a phase shift is imparted to the unscattered ray or the rays which are not diffracted in the object plane 9 while the rays are uninfluenced which are scattered or diffracted in the object plane at higher orders. The unscattered ray passes through the annular opening. Reference can be made to U.S. Pat. No. 6,797,956 with respect to details as to the configuration of the phase plate 16. U.S. Pat. No. 6,797,956 is incorporated herein by reference. By changing the excitation of the second diffraction lens 15, the orientation of the diffraction image relative to the phase plate 16 can be changed by the image rotation effected by this diffraction lens 15. In this way, diffraction spots 25, which are essential for the image formation and which impinge on the holding elements (22, 27), can be moved on a circular path indicated by the circle drawn by a dotted line circle and can thereby be rotated into a region wherein they can pass through the phase plate undisturbed by the holding elements (22, 27). FIG. 4 shows the beam path in the case of the generation of a phase contrast with a conical illumination that is disclosed in U.S. Pat. No. 6,797,956. The unscattered ray is undiffracted in the object plane 49 and is focused in the diffraction plane 50 of the objective 48 and is imaged into the diffraction intermediate image plane 53 by the first diffraction lens 51. The first deflection system 52 in the vicinity of the intermediate image plane 56 guides the beam in an annular shape across the phase-shifting regions of an annularly-shaped phase plate 57 in the diffraction intermediate image plane 53. The second diffraction lens 58 has a principal plane which lies in or in the vicinity of the diffraction intermediate image plane 53 and generates an intermediate image of the object plane 49 in the entrance image plane of the first projective lens 54. A further deflection system 55 is arranged in the entrance image plane of the first projective lens 54 and is operated in synchronism with the first deflection system 52. With this deflection system 55, the ray can again be deflected back onto the optical axis in the entrance image plane of the first projective lens 54. An annularly-shaped phase plate is mounted in the diffraction intermediate image plane 53 and with this phase plate, a phase shift is imparted to the unscattered ray guided on a circular path while the electrons, which pass farther from the edge of the center opening of the phase plate, remain uninfluenced by the phase plate with these electrons having been diffracted in the object plane 49 into higher diffraction orders. A phase contrast again arises via interference of these higher diffraction orders with the phase-shifted unscattered ray. The arrangement of FIG. 5 is built up essentially in the manner of the arrangement shown in FIG. 4. However, the arrangement of FIG. 5 includes two phase plates. The first phase plate 58 is mounted in the diffraction plane 21 of the objective lens and a second phase plate 57 is mounted in the image plane of the first diffraction lens 51 conjugated to the diffraction plane 15. The first diffraction lens 51 lies in or near the principal plane of the second diffraction lens 53. As in the embodiment described above, the second diffraction lens images a real intermediate image of the object plane 49 into the entrance image plane of the projective system 54. The real intermediate image of the object plane 49 lies between the first and second diffraction lenses (51, 53). The two phase plates (57, 58) are so configured that they each influence different spatial regions of the diffraction images which spatial regions are complementary to each other. As indicated in FIG. 5, the phase plate 58, which is mounted in the diffraction plane 21 of the objective, generates a phase shift in the beam conical segment shown hatched in FIG. 5 while the second phase plate 57 generates a phase shift in the non-hatched beam conical segment. The second phase plate 57 is mounted in the diffraction intermediate image plane. The phase shift takes place sequentially in the two mutually conjugated diffraction images so that, from the two contributions together, the wanted phase shift results between the unscattered ray and the diffracted rays. This offers in principle the possibility of realizing matter-free and aberration-free electrostatic phase plates. Furthermore, the two phase plates (57, 58) can each be so configured that possibly present holding elements in total cannot negatively influence the image generation. If only the scattered electrons experience a phase shift of π with a half-planar phase plate, then a differential interference contrast can be generated. In FIG. 6, a somewhat different beam guidance is shown with the embodiment already described with respect to FIGS. 1 and 2. The essential difference is that the objective lens 60 is so highly excited that it generates an intermediate image of the object plane 61 already in or near the principal plane of the first diffraction lens 62. The first diffraction lens 62, in turn, images the diffraction plane 65 of the objective lens 60 into or in the vicinity of the principal plane of the second diffraction lens 63 so that there again a diffraction intermediate image plane arises. The phase-shifting element 67 is then again mounted in the diffraction intermediate image plane. As in the embodiment in FIG. 2, the second diffraction lens 63 images the intermediate image of the object plane 61 into the entrance image plane 66 of the downstream projective system 64 with this intermediate image arising in the principal plane or in the proximity of the principal plane of the first diffraction lens. In FIG. 7, the beam path is shown which is realized for operation of the system as a Lorentz microscope or with low magnification. During operation as a Lorentz microscope, the objective lens 70 is switched off and with low magnification (so-called low-mag-mode), the objective lens is only slightly excited. The excitation of the first diffraction lens 73 is so selected that its focal plane lies in the principal plane of the second diffraction lens 74. The phase plate 77 is, in turn, mounted in the principal plane of the second diffraction lens. The second diffraction lens 74 generates a real image of the virtual image 78 of the object plane 71 in the entrance image plane 76 of the downstream projective system 75, the virtual image being generated by the first diffraction lens 73. The projective system 75 then generates an image magnified by up to a factor of 10,000 of the object positioned in the object plane 71. This beam guidance thereby makes possible a maximum magnification approximately greater by a factor of 5 compared to a conventional low-mag magnification. With respect to the required dimensions in the beam guidance of FIG. 2, the following can be estimated: for a focal length of the objective lens 8 of approximately 3 mm and a desired telescope magnification by a factor of 10 (because the objective lens 8 and the first diffraction lens 11 are operated telescopically), there results for the first diffraction lens 11 a focal length of approximately 30 mm. It can be shown that for telescope systems of this kind, the increase of the spherical aberration is negligible compared to a system only with the objective lens. Should the diffraction intermediate image plane be magnified approximately by the factor of 5 compared to the diffraction plane 10 of the objective lens 8, then the first diffraction lens 11 has to have a distance of approximately 36 mm from the rearward diffraction plane 10 of the objective lens 8 and a distance of 180 mm is required between the first diffraction lens 11 and the second diffraction lens 15. The second diffraction lens 15 images the intermediate image of the object plane 9 into the entrance image plane 17 of the projective system 18 with low magnification with the intermediate image arising in the back focal plane 14 of the first diffraction lens 11. For this reason, a required distance of approximately 150 mm results between the second diffraction lens 15 and the entrance image plane 17 of the projective system 18 for a focal length of the second diffraction lens 15 of approximately 75 mm. In total, there results a structural length increase of approximately 260 mm compared to an arrangement wherein the objective lens 8 directly generates a real intermediate image in the entrance image plane of the projective system. It is understood that the foregoing description is that of the preferred embodiments of the invention and that various changes and modifications may be made thereto without departing from the spirit and scope of the invention as defined in the appended claims. |
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abstract | Devices position inspection and operation tools in a nuclear reactor without use of a bridge or other maintenance structure well above the reactor core. Devices can selectively join to in-reactor structures like steam dams through clamping that permits limited movement. Positioning assemblies may be supported by the devices. The positioning assembly includes a mast that moves relative to the positioning assembly. Rollers of the positioning assembly may so move the mast. The positioning assembly may be rotated with a sun gear having an exterior surface about which the mast and rollers can revolve. A local motor or external drive may provide power for this rotation and/or revolving as well as rotation of the rollers to move the mast and assembly. Devices are useable underwater or submerged in fluid and may include instruments with powering lines or mechanical extensions that permit powering or direct interfacing from operators outside the reactor. |
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description | The present invention relates to the collimation of a beam of radiation, and particularly to a multi-leaf collimator for use in radiotherapy systems. Radiotherapy involves the production of a beam of ionising radiation, usually x-rays or a beam of electrons or other sub-atomic particles. This is directed towards a target region of the patient, and adversely affects the target cells (typically tumour cells) causing an alleviation of the patient's symptoms. Generally, it is preferred to delimit the radiation beam so that the dose is maximised in the target cells and minimised in the healthy cells of the patient, as this improves the efficiency of treatment and reduces the side effects suffered by the patient. For example, the radiation beam may be shaped to conform to the cross-section of the target region. One principal component in delimiting the radiation dose is the so-called “multi-leaf collimator” (MLC). This is a collimator which sits inside the radiation head of the therapeutic system, and consists of a large number of elongate thin leaves arranged side by side laterally in an array. Each leaf is moveable longitudinally so that its tip can be extended into or withdrawn from a radiation field. The leaves can thus be positioned so as to define a variable edge to the beam of radiation, and this is used to impart a variable edge to the radiation beam passing through the radiation field. All the leaves can be withdrawn entirely to open the radiation field (even if in practice this should never occur during operation), or all the leaves can be extended to their fullest extent so as to close it down. Alternatively, some leaves can be withdrawn and some extended so as to define any desired shape, within operational limits. A multi-leaf collimator usually consists of two banks of such arrays, each bank projecting into the radiation field from opposite sides of the collimator. The depth of each leaf is one of the parameters which defines the leaf's ability to mitigate (i.e. block) the radiation beam passing through the window. The material of manufacture also plays a part, and for this reason each leaf is typically manufactured from an element with high atomic number, such as tungsten. However, even using such materials, each leaf must have a significant depth in the direction of the beam in order to adequately block the high-energy radiation used in radiotherapy (where photons usually have energies in the megavolt range). Most leaves have a depth of between 60 and 120 mm, but in practice the deeper a leaf is, the more effective it will be in blocking and shaping the radiation. In order to achieve a high resolution when collimating the radiation beam, each leaf should also be relatively thin in the lateral direction. That is, the tips of the leaves in the array collectively define an edge of the radiation beam. If each leaf is made as thin as possible, a greater number of leaves are used to define the edge and thus the shape of the radiation beam can be defined at a higher resolution. Of course, the leaves on the MLC leaf bank need to be driven in some way. Given the design parameters set out above (i.e. narrow leaves arranged closely together, heavy materials, significant depth etc) this is no trivial task. Typically, this is by a series of lead screws connected to geared electric motors. The leaves are fitted with a small captive nut in which the lead screws fit, and the electric motors are fixed on a mounting plate directly behind the leaves. Rotation of the leadscrew by the motor therefore creates a linear movement of the leaf. Our earlier application, WO 2009/129817, describes an improvement to this design in which each leaf has a lug which extends above or below the leaf, i.e. transverse to the lateral and longitudinal directions. The lug engages with a leadscrew which is itself driven by a motor. The set of motors for each leaf bank can thus sit above or below the banks of leaves rather than behind or to the side of the leaves. However, in both prior designs the motors are arranged to the side of the leaf array. Thus a large amount of space in the radiation head is taken up by the motors rather than the leaves. If the motors could be made more compact, the depth of the leaves could be increased to take up the available space in the radiation head, in turn leading to an increase in the radiation-blocking effect of the collimator. According to a first aspect of the present invention, there is provided a multi-leaf collimator, comprising: a plurality of leaves arranged next to each other in a lateral direction, each leaf having a width in the lateral direction, and being extendible across a window in a longitudinal direction to delimit a radiation beam directed through said window; and a plurality of motors, each motor for driving a respective leaf of the plurality of leaves in said longitudinal direction, wherein each leaf comprises a first portion for delimiting said radiation beam, and a second portion for engagement with a respective motor of the plurality of motors, wherein the second portion has a cut-out section defining an edge for coupling to the motor, wherein each motor has a width in the lateral direction equal to or less than the width of its respective leaf, and wherein the motor is arranged within the lateral extent of the leaf. FIG. 1 shows a radiotherapy system 10 according to embodiments of the present invention. The system comprises a rotatable gantry 12 and a patient support 14 located on or near the rotation axis of the gantry 12. In the illustrated embodiment the gantry 12 is depicted as a circular ring for simplicity, but those skilled in the art will appreciate that the gantry 12 may take any convenient form. A source of therapeutic radiation 16 is mounted on the gantry 12 and directed inwards towards the axis of rotation. According to embodiments of the present invention, the source 16 comprises a linear accelerator, or linac, arranged to accelerate charged particles (such as electrons) to relativistic speeds and energies in the megavoltage (MV) range. In one embodiment, the charged particles are used to treat the patient directly, typically for targets on or near the surface of the patient as the particles do not penetrate human tissue deeply. In another embodiment, the particles are fired towards a high-density target (e.g. tungsten) to generate secondary radiation via mechanisms such as Bremsstrahlung radiation. The secondary radiation so generated includes x-rays up to and including the energy of the charged particle. The therapeutic radiation generated by the source 16 is collimated into a beam having a primary shape (cone-shaped and fan-shaped beams are well known but other shapes are possible) by primary collimators. Further collimation is performed by secondary collimators 18, to adapt the beam to take a desired cross section. Typically the primary collimators will be fixed in place such that the overall shape of the treatment beam (i.e. before secondary collimation) is not changed during treatment. The secondary collimators tend to be more complex, however, and these may be updated during treatment to ensure the treatment beam conforms to a desired cross section. One particularly common secondary collimator is known as a multi-leaf collimator (MLC). The combined effect of the source 16 and the collimator 18 is to produce a beam of radiation 20 having a collimated shape and an energy (typically in the MV range) which has a therapeutic effect in the patient. In use, the therapeutic beam 20 is directed generally towards the rotation axis of the gantry 12. A patient 15 is positioned on the support 14 such that the target for treatment lies on or near the rotation axis of the gantry 12. Rotation of the gantry 12 during treatment causes the beam 20 to be directed towards the target from multiple directions. The target remains in the treatment beam for most (or all) of the time and thus radiation dose accumulates to a relatively high level there. The surrounding healthy tissue also lies within the radiation beam 20 but only for a limited period of time before the gantry rotates and the beam passes through a different part of the patient 15. Radiation dose in the healthy tissue is therefore kept at a relatively low level. FIG. 2 shows a beam's eye view of the multi-leaf collimator 18, with the axis of the beam directed into the page. A housing defines a radiation window 52 through which the radiation beam passes after its primary collimation. A bank of elongate leaves 54 is arranged to the side of the radiation window 52, with the leaves arranged side-by-side in a lateral direction perpendicular to the beam axis. Each leaf is relatively narrow in that lateral direction, and relatively long in a longitudinal direction (perpendicular to both the beam axis and the lateral direction). Each leaf may be manufactured from a high-density material (such as tungsten), and has a significant depth in the direction of the beam axis in order to block the radiation from passing through. In use, the leaves are individually controllable to move in the longitudinal direction, in the direction indicated in FIG. 2, across the radiation window 52 to a greater or lesser extent as required. In one embodiment, each leaf can be extended across the entire radiation window or withdrawn from the entire radiation window, and can be arranged to take any position in between those two extremes. The leaves can therefore be positioned so as to define an aperture through the window 52 of an arbitrary shape, thus collimating the radiation beam to conform to that shape. Only one array of leaves is illustrated for clarity. However, those skilled in the art will appreciate that more than one bank of leaves may be provided, with a common arrangement being to have two banks of leaves arranged on opposing sides of the window 52. Moreover, FIG. 2 is schematic in that the leaves will in practice be much thinner relative to the radiation window 52. A typical leaf may have a width of 2.5 mm, although leaves may have any width as dictated by the design of the apparatus. In one embodiment the MLC leaves may have a width in a range from 1 to 5 mm. For example, each bank of leaves may have 40 or more leaves rather than the 20 leaves illustrated. The multi-leaf collimator 18 further comprises drive means 56 for driving the leaves 54 in the longitudinal direction illustrated. The drive means comprises a plurality of motors 58, at least one for each leaf 54. As can be seen from FIG. 2, the width of each motor is equal to or less than the width of its respective leaf, and further is arranged to lie within the lateral extent of the respective leaf. As will be described in greater detail below, each leaf may have a cut-out section in which the motor is arranged, and in this way the drive means 56 can be made extremely compact. Each motor 58 is coupled to a shaft 60 running along the lateral direction; the position of each motor is thus fixed relative to the shaft, and the action of the motor is to move the corresponding leaf in the longitudinal direction. The shaft cross section may be circular or take any other shape. FIG. 3 shows an isometric view of three leaves 54a, 54b, 54c according to embodiments of the present invention, removed from other elements of the multi-leaf collimator 18 for clarity. FIG. 4 shows a side elevation of a single leaf 54. The orientation of this illustration is such that the radiation beam passes vertically down the page in the direction indicated. The convention used in the following description is that locations nearer to the source of radiation are referred to as being “upper” or “top”, while those locations further from the source of radiation are referred to as being “lower” or “bottom”. As will be appreciated by the skilled reader, such terms are not necessarily indicative of the location's height relative to ground, as the radiation source 16 and collimator 18 may be attached to a gantry 12 able to rotate to any orientation (see FIG. 1). Each leaf comprises a first portion 62, to the right of the dashed line in FIG. 4, which is moved across the radiation window 52 during use and acts to block radiation passing through the window. The first portion 62 comprises a continuous block of material from a top edge 64 to a bottom edge 66, and in the illustrated embodiment these edges 64, 66 define the upper- and lower-most extremes of the leaf 54 as a whole respectively. The continuous nature of the first portion 62 in the direction of the radiation beam maximizes its effectiveness in blocking the radiation. A leaf tip 68 connects the top edge 64 to the bottom edge 66, and in one embodiment this is curved in the longitudinal direction. A curved leaf tip 68 can result in a more accurately collimated radiation beam (i.e. with a narrower penumbra) owing to the divergent nature of the radiation emanating from the source 16. Each leaf 54 further comprises a second portion 70, to the left of the dashed line in FIG. 4, with which the motor 58 can drive the leaf in the longitudinal direction. The second portion is not used to block radiation and therefore the depth of the leaf 54 in the direction of the radiation beam is not of importance. Moreover, as the second portion 70 plays no part in blocking radiation, it may be made from a different material than the first portion 62. For example, while the first portion 62 is generally manufactured from a dense material with relatively high atomic number (such as tungsten), the second portion 70 may be manufactured from a lighter material such as aluminium in order to reduce weight. The second portion 70 has a cut-out section defining an edge 72 to which the motor is coupled. In the illustrated embodiment the edge 72 comprises a plurality of teeth, but in other embodiments the edge may be modified to present some other high-friction surface or may not be modified at all. The cut-out section is such that the edge 72 lies between the upper- and lowermost extremes of the leaf 54, as defined by the top and bottom edges 64, 66 of the first portion 62. In this way, the motor 58 is arranged at least partially within the upper- and lowermost extremes of the leaf 54. In the illustrated embodiment the motor 58 lies entirely within the upper- and lowermost extremes of the leaf 58. Those skilled in the art will appreciate that the cut-out section can take many different shapes, and is not limited to the shape illustrated in FIGS. 3 and 4. For example, in the illustrated embodiment the cut-out section cuts short the top edge 64, as well as the rearmost edge; however, in other embodiments the cut-out section may lie entirely within the body of the leaf 54, without spoiling the edges at the extremes of the leaf's dimensions. The motor 58 illustrated in FIG. 4 and the three motors 58 illustrated in FIG. 3 show one embodiment of the present invention. In this embodiment, the motor comprises a hub 74, coupled to the shaft 60, and a rotor 76 which rotates relative to the hub 74 about an axis coincident with the shaft 60. A power supply 78 (e.g. electric wires coupled to a power source) is connected to the hub 74 and provides power for the motor 58 to operate. Wires coupled to the motor may further provide control signals for controlling the operation of the motor 58, and/or feedback signals from the motor to the controlling entity (so as to provide information on the leaf position, for example, or to report a malfunction in the motor or the leaf). In one embodiment, the rotor 76 is axially concentric with the hub 74, and has a greater radius than the hub 74. In this way, the rotor projects outwards of the hub 74 and can engage with the edge 72 of the leaf 54. In embodiments where the edge comprises a plurality of teeth the rotor 76 may also comprise a corresponding plurality of teeth for engagement between the two in a rack-and-pinion style system. The combination of the hub 74 and the rotor 76 together have a lateral width which is equal to or less than the lateral width of the leaf 54, and thus the entire motor 58 lies within the lateral extent of the leaf. It is necessary for the leaves 54 to lie close together such that radiation does not pass unblocked between them. When multiple leaves 54 and multiple motors 58 are combined together, as illustrated in FIG. 3, the motors are sufficiently narrow that they do not impede this close-packed arrangement. Multiple motors can be coupled to the same shaft 60, each independently operable to drive a corresponding leaf in the longitudinal direction. As the motors 58 effectively lie within the space which would ordinarily be occupied by the leaves 54 themselves, more room is created in the radiation head of the radiotherapy apparatus. The leaves 54 can thus be made deeper (i.e. in the direction of the radiation beam) than would otherwise be the case, and radiation can be blocked more effectively. In other embodiments, the motors 58 may take a different form. However, in each case the motors have a lateral width which is equal to or less than the width of the leaf, such that they can be arranged within the lateral extent of the leaf. One example of such an alternative motor is illustrated in FIG. 5, where piezoelectric motors are used to drive leaves in the multi-leaf collimator. Such motors employ a piezoelectric element 80 and an actuator 82 coupled between the element 80 and the leaf edge 72. By changing the voltage applied to the piezoelectric element 80 rapidly, the actuator 82 can be moved with sufficient force to overcome the frictional force between the actuator and the leaf edge, such that the actuator moves relative to the leaf (while the leaf remains stationary). The voltage to the piezoelectric element can then be changed in the opposite direction more slowly, such that the frictional forces between the actuator and the leaf edge are not overcome. If the piezoelectric element is fixed with respect to the multi-leaf collimator (for example through a fixing to the shaft 60), the leaf will be moved a short distance through its frictional coupling to the actuator. By repeating the process, the leaf can be moved through greater distances in a stepping motion. A more detailed explanation of this method of driving multi-leaf collimator leaves can be found in U.S. Pat. No. 7,792,252. Other motors which may be manufactured in a form which is narrower or equal in width to the leaves include: a piezoelectric “squiggle”® motor manufactured by New Scale Technologies; and motors mounted on a printed circuit board (such as those manufactured by PCBMotor). The invention is not limited to any particular type of motor, except that they have a lateral width which is equal to or less than the width of the leaves. Embodiments of the present invention thus provide a multi-leaf collimator with a plurality of leaves and at least one motor for each leaf. The motor for each leaf has a lateral width which is equal to or narrower than the corresponding leaf, and in this way the motors can be arranged within the lateral extent of the leaf. A cut-out section in the leaf allows the motor to lie at least partially within the depth of the leaf, and in this way the drive mechanism and the multi-leaf collimator as a whole are made extremely compact. This in turn allows the leaves to be deeper than would otherwise be the case, increasing their efficacy in blocking radiation. Those skilled in the art will appreciate that various amendments and alterations can be made to the embodiments described above without departing from the scope of the invention as defined in the claims appended hereto. |
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043485910 | claims | 1. A straight bore flat field masking collimator for multiple detector imaging gamma cameras comprising: (A) an open generally cylindrical radiation shield, having a camera end and an aperture end, the greatest inner dimension thereof being substantially between two and five inches, and (B) a radiation shielding portion at the camera end of said cylindrical radiation shield for mounting to a gamma camera and for masking off the greater portion of the gamma camera's multiple detectors. (A) a generally cylindrical radiation shield having a camera end and an aperture end, the greatest inner dimension thereof being substantially between two and five inches, said shield being a solid cylinder of radiation shielding material having a plurality of channels, and (B) a radiation shielding portion at the camera end of said cylindrical radiation shield for mounting to a gamma camera and for masking off the greater portion of the gamma camera's multiple detectors. 2. The collimator of claim 1 wherein said radiation shielding portion comprises an upstanding open sleeve-like radiation shield having a camera end and an aperture end, the maximum inside dimension of said aperture end being equal to or greater than three inches, means at the camera end of said sleeve-like radiation shield for attachment to a gamma camera, and means at the aperture end for mounting said cylindrical radiation shield. 3. The collimator defined in claim 2 wherein said sleeve-like radiation shield is frustoconical. 4. The collimator defined in claim 1 wherein the walls of said generally cylindrical radiation shield are thicker at the camera end thereof than at the aperture end thereof. 5. The collimator of claim 1 wherein said radiation shield is substantially a right circular cylinder. 6. The collimator of claim 1 wherein the length of said cylindrical radiation shield is substantially greater than its inner diameter. 7. The collimator of claims 1, 5, or 6 wherein said radiation shielding portion is a substantially flat flange at the camera end of said cylindrical radiation shield. 8. A masking collimator for a multiple detector imaging gamma camera comprising: 9. The collimator of claim 8 wherein said channels are parallel. 10. The collimator of claim 8 wherein said channels are conically arranged, said channels radiating from a common point external thereto. |
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060758384 | description | DETAILED DESCRIPTION An example of a plasma x-ray source in accordance with the present invention is shown in FIG. 1. An enclosed chamber 10 defines a pinch region 12 having a central axis 14. The chamber 10 may include an x-ray transmitting window 16 located on axis 14. A gas inlet 20 and a gas outlet 22 permit a gas at a prescribed pressure to be introduced into the pinch region 12. The example of FIG. 1 has a generally cylindrical pinch region 12. A cylindrical dielectric liner 24, which can be a ceramic material, surrounds pinch region 12. An RF electrode 26 is disposed on the outside surface of dielectric liner 24. A pinch anode 30 is disposed at one end of the pinch region 12, and a pinch cathode 32 is disposed at the opposite end of pinch region 12. The portion of pinch anode 30 adjacent to pinch region 12 has an annular configuration disposed on the inside surface of the dielectric liner 24. Similarly, the portion of cathode 32 adjacent to pinch region 12 has an annular configuration inside dielectric liner 24 and spaced from dielectric liner 24. In a preferred embodiment, the pinch cathode 32 includes an annular groove 50 which controls the location at which the plasma shell attaches to cathode 32. Preferably, the anode 30 has an axial hole 31, and the cathode 32 has an axial hole 33 to prevent vaporization by the collapsed plasma, as described below. The anode 30 and the cathode 32 are connected to an electrical drive circuit 36 and are separated by an insulator 40. The anode 30 is connected through a cylindrical conductor 42 to the drive circuit 36. The cylindrical conductor 42 surrounds pinch region 12. As described below, a high current pulse through cylindrical conductor 42 contributes to an azimuthal magnetic field in pinch region 12. An elastomer ring 44 is positioned between anode 30 and one end of dielectric liner 24, and an elastomer ring 46 is positioned between cathode 32 and the other end of dielectric liner 24 to ensure that the chamber 10 is sealed vacuum tight. In the example of FIG. 1, the chamber 10 is defined by cylindrical conductor 42, an end wall 47 and an end wall 48. The cylindrical conductor 42 and end wall 47 are electrically connected to anode 30, and end wall 48 is electrically connected to cathode 32. It will be understood that different chamber configurations can be used within the scope of the invention. The RF electrode 26 is connected through an RF power feed 52 to an RF generator 200 which supplies RF power for preionizing the gas in a cylindrical shell of pinch region 12. The RF power preferably has a power level greater than one kilowatt. In a preferred embodiment, the RF power is 5 kilowatts at 1 GHz. It will be understood that different RF frequencies and power levels can be used within the scope of the present invention. In a preferred embodiment, the RF electrode 26 comprises a center-fed spiral antenna wrapped around the dielectric liner 24, with a total angular span of +/-200.degree.. It will be understood that different spiral configurations and different RF electrode configurations can be utilized for preionizing the gas in the pinch region 12. The spiral configuration described above has been found to provide satisfactory results. The drive circuit 36 supplies a high energy, short duration of electrical pulse to anode 30 and cathode 32. In a preferred embodiment, the pulse is 25 kilovolts at a current of 300 kiloamps and a duration of 200-250 nanoseconds. The inside wall of dielectric liner 24, the anode 30 and the cathode 32 define a cylinder of low density gas. RF power is applied to the RF electrode 26 to cause ionization within the gas cylinder. It is a property of the application of intense RF power to a gas surface that the ionization is concentrated in a surface layer. This is exactly what is needed to create a precise cylindrical plasma shell 56 for the subsequent passage of current. Once the gas has been preionized by RF energy, the drive circuit 36 is activated to apply a high energy electrical pulse between anode 30 and cathode 32. Typically, the RF power is applied 1-100 microseconds before the drive circuit 36 is activated. The high energy pulse causes electrons to flow from the pinch cathode 32 to the pinch anode 30. Initially, the current flows in the preionized outer layer of the gas cylinder and forms plasma shell 56. The return current flows back to the drive circuit 36 through the outer cylindrical conductor 42. An intense azimuthal magnetic field is generated between the outer current sheet through cylindrical conductor 42 and the current sheet in the plasma shell 56. The magnetic field applies a pressure which pushes the plasma shell 56 inward toward the axis 14. After approximately 200-250 nanoseconds, the drive circuit 36 is discharged and the current drops to a lower value. At approximately the same time, the plasma shell reaches the axis 14 with high velocity, where its motion is arrested by collisions with the incoming plasma shell from the opposite radial direction. The result of this stagnation process is the conversion of kinetic energy into heat, which further ionizes the gas into high ionization states that radiate x-rays strongly in all directions. In the case of population inversion on an x-ray transition and in cases when the plasma is optically dense in the axial direction but optically thin in radial directions, the radiation is concentrated in the two axial directions via amplified spontaneous emission. Thus with reference to FIG. 1, the plasma shell 56 collapses to form a collapsed plasma 60 on axis 14 in approximately 200-250 nanoseconds. RF generator 200 supplies RF energy to RF electrode 26 through RF power feed 52. The RF generator 200 may be any suitable source of the required frequency and power level. A regulated gas supply 202 is connected to gas inlet 20, and a vacuum pump 204 is connected to gas outlet 22. The gas supply 202 and the vacuum pump 204 introduce gas into pinch region 12 and control the pressure at the desired pressure level. In drive circuit 36, multiple circuits are connected in parallel to the pinch anode 30 and the pinch cathode 32 to achieve the required current level. A preferred embodiment utilizes six to eight drive circuits connected in parallel, each generating about 20 to 40 kiloamps. As shown in FIG. 1, each drive circuit includes a voltage source 210 connected to an energy storage capacitor 212. A switch 214 is connected in parallel with storage capacitor 212. The switch 214 may comprise a multiple channel pseudospark switch as described in U.S. Pat. No. 5,502,356 issued Mar. 26, 1996 to McGeoch, which is hereby incorporated by reference. The switch 214 may also comprise a hydrogen thyratron. The switches 214 in the parallel circuits are closed simultaneously to generate a high energy pulse for application to the anode 30 and cathode 32. Additional information regarding the Z-pinch plasma X-ray source is disclosed in U.S. Pat. No. 5,504,795, which is hereby incorporated by reference. According to the present invention, the gas introduced into the pinch region 12 is a gas mixture including a diluent gas and a primary X-ray emitting gas. The gas mixture renders radiating transitions of the primary gas optically thin in directions other than axial, thereby enhancing the axial radiation intensity that is achievable during recombination. Typically, the diluent gas is a substantial fraction of the gas mixture introduced into the pinch region prior to electrical excitation of the source. Because a smaller volume of the relatively expensive primary X-radiating gas is used, the cost of operating the X-ray source is reduced. The diluent gas should have low atomic number (preferably less than Z=8) in order to completely ionize without requiring too great an energy input, which would otherwise detract from the energy available for ionization of the primary radiating gas. The diluent gas typically can be, but is not limited to, helium, hydrogen, deuterium, nitrogen and combinations thereof. An example of the invention is the enhanced Z-pinch axial emission of xenon in the 134 angstrom band useful for lithography using helium as the diluent gas. Data from a 4 centimeter long Z-pinch region indicates an approximate 40% increase in the xenon band axial intensity at 134 angstroms as the helium diluent fraction is increased from 0% to 75% of a helium-xenon mixture. The typical evolution of the xenon band spectrum with helium dilution is shown in FIG. 2, with a spectral range from 100 angstroms to 150 angstroms as shown. Curves 300, 302 and 304 represent xenon percentages of 17%, 25% and 35%, respectively, in the gas mixture, with the balance being helium. In FIG. 2, the total gas density in the pinch region has been adjusted in each case to yield optimum spectral intensity at 134 angstroms. A corresponding set of data from an 8 centimeter Z-pinch region is shown as curve 320 in FIG. 3. Although the enhancement with dilution appears to be less for the longer pinch, it amounts to a 20% increase, with the optimum again being observed for the 25% Xe/75% He mixture. It has also been shown that both hydrogen and nitrogen can be substituted for helium with very little change in axial radiation efficiency. It is presumed that deuterium would perform in a similar manner. The use of helium as a diluent is preferred over more chemically active elements, such as hydrogen or nitrogen, in order to give the source maximum compatibility with user systems that might be exposed to low concentrations of the pinch gas mixture at remote locations down an evacuated X-ray beamline. Very low xenon concentrations can be employed in helium diluent with little loss of efficiency. FIG. 3 shows that as little as 0.7% Xe in helium will yield 80% of the intensity that occurs with 25% Xe in helium. This circumstance allows very efficient photon production per flowing xenon atom, although it is to be noted that approximately two times the total gas pressure is required for the lowest xenon cases, in order to optimize the spectral intensity in the band at 134 angstroms. The primary X-radiating gas contained within pinch region 12 can be any gas having suitable transitions for X-ray generation. Examples include, but are not limited to xenon, argon, krypton, neon and oxygen. The total gas pressure is selected to give high enough gas density to ensure a high collision rate as the gas stagnates on the axis, but not so high a density that the motion is slow and the incoming kinetic energy is too low to create the high temperature for needed for X-ray emission. Typically, the total gas pressure of the X-radiating gas and the diluent gas is in a range of about 0.1 torr to 1.0 torr. Gas may be caused to flow through pinch region 12 continuously or may be pulsed with a relatively long time constant. The pressure in the pinch region 12 should be substantially uniform when the high current electrical pulse is applied to the source. As described above, a higher total gas pressure is required when the primary X-radiating gas is a small fraction of the gas mixture. While there have been shown and described what are at present considered the preferred embodiments of the present invention, it will be obvious to those skilled in the art that various changes and modifications may be made therein without departing from the scope of the invention as defined by the appended claims. |
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claims | 1. A method for processing radioactive waste comprising:separating radioactive waste material into liquid waste with radioactive isotopes and solid waste;passing the liquid waste with radioactive isotopes through an inlet line of an ion exchange column through a dip tube oriented in a generally vertical direction within the ion exchange column, wherein the liquid waste is dispersed throughout the ion exchange column by a distribution ring attached to the dip tube and suspended within the ion exchange column in a generally horizontal orientation, wherein the distribution ring comprises a plurality of distribution holes that direct the liquid waste in a downward direction towards a bottom end of the ion exchange column, and wherein the dispersed liquid waste is pushed through media contained within the ion exchange column for capturing one or more of the radioactive isotopes from the liquid waste;pushing the liquid waste with a reduced amount of the radioactive isotopes out of the ion exchange column through an outlet line; andvitrifying the media and captured radioactive isotopes contained in the ion exchange column into a glass matrix wherein the dip tube and the distribution ring are both encased within the glass matrix. 2. The method of claim 1 wherein the ion exchange column comprises a first column containing a first media for capturing a first group of radioactive isotopes and further comprising:passing the liquid waste from the outlet line of the first column into a second inlet line of a second column, through a dip tube into the second column, and pushing the liquid waste through a second media contained within the second column for capturing a second different group of radioactive isotopes from the liquid waste; andpushing the liquid waste with a reduced amount of the second different group of radioactive isotopes out through a second outlet line of the second column. 3. The method of claim 2 further comprising:passing the liquid waste from the second outlet line into a third inlet line of a third column, through a dip tube into the third column, and pushing the liquid waste through a third media contained within the third column for capturing a third different group of radioactive isotopes from the liquid waste; andpushing the liquid waste with a reduced amount of the third different group of radioactive isotopes out through a third outlet line of the third column. 4. The method of claim 3, wherein: the first media is configured to capture cesium radioactive isotopes;the second media is configured to capture strontium radioactive isotopes; andthe third media is configured to capture technetium radioactive-isotopes. 5. The method of claim 1 further comprising, after passing the liquid waste through the ion exchange column, recycling the liquid waste to a nuclear reactor. 6. The method of claim 1 wherein the one or more radioactive isotopes include at least one isotope of cesium. 7. The method of claim 1 wherein the one or more radioactive isotopes include at least one isotope of strontium. 8. The method of claim 7 wherein the media comprises hydroxyapatite. 9. The method of claim 7 wherein the media comprises microspheres that include hydroxyapatite. 10. The method of claim 1 wherein the one or more radioactive isotopes include at least one isotope of technetium. 11. The method of claim 1 wherein the one or more radioactive isotopes include at least one isotope of iodine. 12. The method of claim 1 wherein the one or more radioactive isotopes include at least one isotope selected from the group consisting of nickel, cobalt, and lead. 13. The method of claim 1 wherein the media comprises glass beads fabricated from a mixture of sodium, calcium and boron, the glass beads are mixed with potassium phosphate solution, and the sodium, calcium, and boron react with the potassium phosphate solution to form a hydroxyapatite layer on the glass beads. 14. The method of claim 1 wherein the dip tube and the distribution ring are both formed from a ceramic material. 15. The method of claim 1 wherein the dip tube and the distribution ring are both formed from a porous graphite material. 16. The method of claim 1 wherein the ion exchange column comprises a Herschelite material. 17. The method of claim 1 wherein the ion exchange column comprises an outer steel layer and an inner graphite layer, and wherein vitrifying the media comprises inductively heating the media using the inner graphite layer of the ion exchange column as a susceptor. 18. The method of claim 17 wherein the inner graphite layer forms a crucible for melting the media. 19. The method of claim 17 wherein the ion exchange column further comprises a middle layer of insulation located between the outer steel layer and the inner graphite layer. 20. The method of claim 1, wherein the distribution ring is suspended above the bottom end of the ion exchange column. 21. The method of claim 20, wherein both the inlet line and the outlet line are located at a top end of the ion exchange column, and wherein the dispersed liquid waste is pushed through the media from the bottom end of the ion exchange column to the top end of the ion exchange column. |
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description | The present disclosure generally relates to nuclear fuel and a method for fabricating a nuclear fuel, and, more particularly, to a swelling resistant nuclear fuel and a method for fabricating a swelling resistant nuclear fuel. In one aspect, a nuclear fuel includes, but is not limited to, a volume of a nuclear fuel material defined by a surface, the nuclear fuel material including a plurality of grains, some of the plurality of grains having a characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a grain interior to at least one grain boundary in some of the grains, the nuclear fuel material including a boundary network configured to transport the fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material. In another aspect, a nuclear fuel includes, but is not limited to, a volume of a nuclear fuel material defined by a surface, the nuclear fuel material including a plurality of nuclear fuel elements, the nuclear fuel elements including a metal, some of the plurality of nuclear fuel elements having a characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, the plurality of nuclear fuel elements consolidated to a selected density. In another aspect, a nuclear fuel includes, but is not limited to, a volume of a nuclear fuel material defined by a surface, the nuclear fuel material including a plurality of nuclear fuel elements, the nuclear fuel elements including a ceramic material, some of the plurality of nuclear fuel elements having an characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, the plurality of nuclear fuel elements consolidated to a selected density, the nuclear fuel material including a boundary network configured to transport the fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. In another aspect, a nuclear fuel includes, but is not limited to, a volume of a nuclear fuel material defined by a surface, the nuclear fuel material including a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having an characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, and a plurality of dispersant particles dispersed within the volume of a nuclear fuel material, wherein some of the dispersant particles are configured to create preferential fission product occupation sites within the nuclear fuel material. In one aspect, a method for fabricating a nuclear fuel may include, but is not limited to, providing a nuclear fuel material, the nuclear fuel material consolidated into a solid volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains, and performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the selected distance is suitable for maintaining adequate diffusion of a fission product from a grain interior to at least one grain boundary in some of the grains. In another aspect, a method for fabricating a nuclear fuel may include, but is not limited to, providing a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having an characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, some of the nuclear fuel elements including a metal nuclear fuel material, and consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. In another aspect, a method for fabricating a nuclear fuel may include, but is not limited to, providing a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having an characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, some of the nuclear fuel elements including a ceramic nuclear fuel material, and consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, the volume of nuclear fuel material including a boundary network configured to transport the fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. In another aspect, a method for fabricating a nuclear fuel may include, but is not limited to, providing a nuclear fuel material, dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles are configured to create preferential fission product occupation sites within the nuclear fuel material, consolidating the nuclear fuel material into a volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains, and performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a fission product from a grain interior to at least one grain boundary in some of the grains. In another aspect, a method for fabricating a nuclear fuel may include, but is not limited to, providing a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having a characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles are configured to create preferential fission product occupation sites within the nuclear fuel material, and consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. In addition to the foregoing, various other nuclear fuel and/or method aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is NOT intended to be in any way limiting. Other aspects, features, and advantages of the devices and/or processes and/or other subject matter described herein will become apparent in the teachings set forth herein. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. Referring generally to FIGS. 1A through 4, a nuclear fuel and a method for fabricating a nuclear fuel are described. The nuclear fuel of the present disclosure may be structured in order to provide a more efficient release of a fission product 108 (e.g., fission gas 118, fission liquid 119, or fission solid 120) created within a volume of the nuclear fuel during a fission reaction process. The efficient release of fission gas 118, for instance, may minimize the growth and development of void regions within the nuclear fuel volume. As pressure builds within the void regions the resultant force may lead the nuclear fuel to “swell.” As the nuclear fuel swells, the outer surface of the nuclear fuel volume may exert a force on the surrounding cladding. Moreover, in addition to swelling avoidance, efficient release of fission products 108 may also reduce parasitic neutron capture by the various fission products 108 and the residual decay heat from the nuclear fuel. Referring now to FIGS. 1A and 1B, a nuclear fuel 100 and methods for making a nuclear fuel are described in accordance with the present disclosure. A given volume 102 of a nuclear fuel 100 may include a plurality of grains 104 of one or more nuclear fuel materials. The one or more nuclear fuel materials of the nuclear fuel 100 may be microstructurally engineered such that the grains 104 of the nuclear fuel material 100 have a characteristic length 106 along at least one dimension that is smaller than or equal to a selected distance. The distance may be selected based on a critical distance necessary to maintain an adequate level of diffusion of a fission product 108, such as a gaseous fission product 118 (e.g., krypton or xenon), a liquid fission product 119 (e.g., liquid sodium), or a solid fission product 120 (e.g., tellurium or cesium), from the interior 110 of the grains 104 to the grain-boundaries 112 of the grains 104. It is recognized that by increasing the ratio between the grain-boundary area and the volume of the grain interior 110 in a given grain 104 the number of fission gas 118 bubbles formed at the grain-boundary 112 as a result of diffusion of fission gas 118 from the grain interior 110 may increase. Therefore, by decreasing the size in one or more dimensions 106 (e.g., average size) of the grains 104 of the nuclear fuel material, thereby increasing the ratio between grain-boundary area and grain interior volume, it is possible to enhance the diffusion of fission gas 118, or other fission products 108, from the grain interiors 110 to the grain-boundaries 112 of the grains 104 of the nuclear fuel 100. In doing so, the likelihood of fission gas nucleation at the grain-boundary 112 may be increased, while the likelihood of fission gas nucleation within the grain interior 110 is simultaneously decreased. In this sense, as the size of one or more grains 104 is decreased in one or more dimensions the fission product 108 (e.g., fission gas 118), which is produced in proportion to the power of the reactor (i.e., flux in reactor core), concentration gradient is increased. The increased fission product concentration aids in regulating the maximum fission product concentration level within the one or more grains 104 of the nuclear fuel 100. Further, the nuclear fuel 100 may include a boundary network 114 configured to transport a fission product 108, such as a fission gas 118, from the grain-boundaries 112 of the grains 104 of the nuclear fuel 100 to the external geometric surface 101 of a given volume 102 of the nuclear fuel 100. If a given fission gas 118 bubble has an open transportation pathway 116 to the geometric surface 101 of the nuclear fuel 100 then the fission gas bubble 118 may be released from the nuclear fuel material volume 102. The aggregated effects of fission gas 118 release across the entire volume 102 of the nuclear fuel 100 may result in a reduction or elimination of swelling in the nuclear fuel 118 upon implementation in an operational setting of a nuclear reactor. In addition to improving fission gas diffusion from a grain-interior 110 to a grain-boundary 112, the engineering of reduced sized grains 104 also increases the spatial density of transportation pathways 116 of the boundary network 114 of the nuclear fuel 100, thereby increasing the likelihood of transportation of a given fission gas bubble from a grain-boundary 112 to the geometric surface 101 of the nuclear fuel 100. In one aspect of the present invention, one or more processes may be utilized in order to achieve the characteristic length 106 along one or more directions (i.e., grain size) of the one or grains 104 required for adequate diffusion of fission products 108 and the boundary network 114 in the nuclear fuel 100. In the context of the present disclosure the term “size” is used interchangeably with “characteristic length along one or more dimensions” and “size along one or more dimensions” for purposes of brevity and clarity. Referring now to FIG. 1C, the fission product 108 produced within the interior 110 of one or more grains 104 of the nuclear fuel 100 may include a fission gas 118, a liquid fission product 119, or a solid fission product. For example, the grains 104 of the nuclear fuel 100 may have a grain size 106 below a critical level required for a fission gas 108 to adequately diffuse from the interiors 110 of the grains 104 of the nuclear fuel to the grain-boundaries 112 of the nuclear fuel 100. For instance the grains 104 of the nuclear fuel 100 may be engineered to have a size smaller than the critical size necessary for adequate diffusion of krypton produced during a fission process within the nuclear fuel 100. In another instance, the grains 104 of the nuclear fuel 100 may be engineered to have a size smaller the critical size necessary for adequate diffusion of xenon produced during a fission process within the nuclear fuel 100. In the case of xenon, which is a fission product of uranium, it is of particular interest to provide a means for efficient transport of the gas from the nuclear fuel 100 interior to the nuclear fuel geometric surface 101. Xenon is a significant neutron absorber and its build up within the nuclear fuel 100 may have a significant negative neutronic impact on a nuclear fuel 100. In another example, the grains 104 of the nuclear fuel 100 may have a grain size 106 below a critical level required for a solid fission product 120 to adequately diffuse from the interiors 110 of the grains 104 of the nuclear fuel to the grain-boundaries 112 of the nuclear fuel 100. For instance, the grains 104 of the nuclear fuel 100 may have a grain size 106 below a critical level required for cesium to adequately diffuse from the interiors 110 of the grains 104 of the nuclear fuel to the grain-boundaries 112 of the nuclear fuel 100. In another example, the grains 104 of the nuclear fuel 100 may have a grain size 106 below a critical level required for a liquid fission product 119 to adequately diffuse from the interiors 110 of the grains 104 of the nuclear fuel to the grain-boundaries 112 of the nuclear fuel 100. For instance, the grains 104 of the nuclear fuel 100 may have a grain size 106 below a critical level required for a liquid metal to adequately diffuse from the interiors 110 of the grains 104 of the nuclear fuel to the grain-boundaries 112 of the nuclear fuel 100. The diffusion of fission products in uranium dioxide is generally described in S. G. Prussin et al., “Release of fission products (Xe, I, Te, Cs, Mo, and Tc) from polycrystalline UO2,” Journal of Nuclear Materials, Vol. 154, Issue 1 pp. 25-37 (1988), which is incorporated herein by reference. The diffusion of fission products in thorium metal is generally described in C. H. Fox Jr. et al., “The diffusion of fission products in thorium metal,” Journal of Nuclear Materials, Vol. 62, Issue 1 pp. 17-25 (1976), which is incorporated herein by reference. The migration of gaseous and solid fission products in a uranium-plutonium mixed oxide fuel is generally described in L. C. Michels et al., “In-pile migration of fission product inclusions in mixed-oxide fuels,” Journal of Applied Physics, Vol. 44, Issue 3 pp. 1003-1008 (1973), which is incorporated herein by reference. Referring now to FIG. 1D, the nuclear fuel 100 may incorporate any known nuclear fissile material. For example, the nuclear fuel 100 may include, but is not limited to, a uranium based material 121, a plutonium based material 122, or a thorium based materials 123. For instance, the nuclear fuel 100 of the present invention may contain 235U. In another instance, the nuclear fuel 100 of the present invention may contain 239PU. Further, it should be recognized that the nuclear fuel 100 need not be fissile directly upon fabrication. For instance, the nuclear fuel 100 of the present invention may implement a 232Th based material, which is not fissile. Thorium-232 may, however, be implemented in a breeder reactor context, wherein 232Th may be bred into 233U, which is suitable for fission. Therefore, in a general sense, the nuclear fuel 100 of the present invention may incorporate a non-fissile material, which may then be bred into a fissile material. It should be recognized that the fissile and non-fissile materials described above should not be interpreted as limitations, but merely illustrations as it is anticipated that additional materials may be suitable for implementation in the nuclear fuel 100 of the present invention. Referring now to FIG. 1E, the nuclear fuel material of the nuclear fuel 100 may include one or more metallic nuclear fuel material 124, such as, but not limited to, a substantially pure metal nuclear fuel material 125, a metal alloy nuclear fuel material 126, or an intermetallic nuclear fuel material 127. For example, a pure metal nuclear fuel material 124 may include, but is not limited to, uranium-235, plutonium-239, or thorium-233. In another example, a metal alloy nuclear fuel material 126 may include, but is not limited to, uranium-zirconium, uranium-plutonium-zirconium, uranium-zirconium-hydride, or uranium aluminum. By way of a further example, an intermetallic nuclear fuel material 127 may include, but is not limited to, UFe2 or UNi2. It should be recognized that the above list of suitable metallic nuclear fuel materials for inclusion in the nuclear fuel material of the nuclear fuel 100 of the present invention should not be interpreted as a limitation but rather merely an illustration. In another embodiment, the nuclear fuel material of the nuclear fuel 100 may include one or more ceramic nuclear fuel materials 128, such as, but not limited to, an oxide nuclear fuel material 129, a nitride nuclear fuel material 131, or a carbide nuclear fuel material 132. For example, an oxide based nuclear material 129 may include, but is not limited to, uranium dioxide (UO2), plutonium dioxide (PuO2), or thorium dioxide (ThO2). Moreover, an oxide based nuclear fuel material 129 may include a mixed oxide nuclear fuel material, such as, but not limited to, a mixture of PuO2 and depleted or natural UO2. In another example, a nitride based nuclear fuel material 131 may include, but is not limited to, uranium-nitride or plutonium nitride. By way of a further example, a carbide base nuclear fuel material may include, but is not limited to, uranium carbide 132. It should be recognized that the above list of suitable ceramic nuclear fuel materials for inclusion in the nuclear fuel 100 of the present invention should not be interpreted as a limitation but rather merely as an illustration. While the nuclear materials described above are done so in the context of material “grains” and FIG. 1A, it should be appreciated that the implementation of these materials may be extended to other contexts, such as those described in FIGS. 2A through 4 of the present disclosure. It should be recognized that, in addition to the fissionable nuclear material described above, the nuclear fuel 100 of the present invention may also include portions of non-fissionable material, such as, but not limited to, neutron moderating material or neutron reflective material. In a general sense, the term “nuclear fuel” in the context of the present disclosure is not limited to fissionable material but may encompass an entire volume of an object or material used as a fuel source in a nuclear reactor setting. Therefore, while the term “nuclear fuel” may be used to refer to the material in a given volume, it may also be extended to embodiments of the nuclear fuel material implemented in a nuclear reactor setting, such as fuel pellets, fuel pebbles, or fuel rods. Referring now to FIGS. 1F through 1H, the characteristic length 106 along at least one dimension of one or more grains 104 may include a characteristic length 106 along all dimensions of one or more grains 104 of the nuclear fuel 100. For example, the grains 104 of the nuclear fuel 100 may be engineered such that the “height”, represented by “a,” and “width,” represented by “b” are similar in size. Therefore, notwithstanding of factors (e.g., stress or thermal gradients), a fission product 108 may efficiently diffuse from the grain interior 110 to the grain boundary 112 along all directions within the grain. In this context, a grain structure may be characterized by the “grain size” of the grains 106 of the nuclear fuel 100. The “grain size” may be selected such that the grains are small enough to allow for adequate diffusion from the interiors 110 of the one or more grains 104 to the boundaries 112 of the one or more grains 104. As shown in FIG. 1G, the characteristic length 106 along at least one dimension of one or more grains 104 may include a characteristic length 106 along a selected dimension of one or more grains 104. For example, as shown in FIG. 1G, the grains 106 within the nuclear fuel 100 may be engineered to have a selected characteristic length 106 along a given dimension of the grains 106. For instance, in a grain 104 having an elongated grain structure, the grain may have a selected characteristic length along the “thin” dimension, shown as dimension “a” in FIG. 1G, of the grain 106. In another instance, in a grain 106 having an elongated grain structure, the grain 104 may have a selected characteristic length along the “thick” dimension, shown as dimension “b” in FIG. 1G, of the grain 106. It should be recognized that the grain 104 need only have at least one characteristic length 106 smaller than the distance required for adequate diffusion from the interiors 110 of the one or more grains 104 to the boundaries 112 of the one or more grains 104. It is further recognized, however, that all dimensions of a grain 104 may have a characteristic length 106 smaller than or equal to a distance required for adequate diffusion of fission product 108 from the interiors 110 of the one or more grains 104 to the boundaries 112 of the one or more grains 104. As shown in FIG. 1H, the characteristic length 106 along at least one dimension of one or more grains 104 may include a characteristic length 106 along a selected direction 134. For example, the grains 106 within the nuclear fuel 100 may be engineered to have a selected characteristic length 106 along a given direction in the nuclear fuel 100. For instance, a grain 104 having an elongated grain structure may have a selected characteristic length 106 along a selected direction 134 within the nuclear fuel 100. It should be recognized that engineering the grain structures to have a characteristic length 106 along a selected direction 134 smaller than the length required for adequate diffusion of a fission product 108 from a grain interior 110 to a grain-boundary 112 may supply a more efficient means for transferring fission product 108 (e.g., fission gas 118) from the grain interior 110. In another embodiment, one or more grains 104 may have a characteristic length 104 along a dimension of the one or more grains selected to maximize heat transfer from a grain-interior 110 to a grain-boundary 112. For example, the one or more grains 106 may be oriented such that their narrow dimensions, shown as “a” in FIG. 1H, are aligned substantially perpendicular to a thermal gradient 136 in the nuclear fuel 100. Such an arrangement aids in the heat transfer from the grain-interior 110 to the grain-boundary, aiding in the diffusion of a fission product 108 from the grain interior 110 to its grain boundary 112. By way of another example (not shown), in a cylindrical fuel pellet fabricated utilizing the nuclear fuel 100 of the present invention the grains 104 of the nuclear fuel 100 may be arranged (i.e., on average the grains of the material may be arranged) to have their the narrow dimension substantially perpendicular to the radial thermal gradient of the cylindrical pellet. It should be noted that the illustrations in FIGS. 1H, 1G, and 1F represent simplified conceptual illustrations of a plurality of grains 106 consistent with the present invention and should not be interpreted as schematical in nature. Further, it should be recognized by those skilled in the art that a variety of materials processing techniques (e.g., cold-working and/or annealing, compression, or extrusion) may be implemented in order to develop the symmetrical grain structure in FIG. 1F, and the deformed elongated grain structure illustrated in FIGS. 1G and 1H. A variety of materials processing techniques are discussed further herein. In another embodiment, the grains 106 of the nuclear fuel 100 may have an average characteristic length 106 along at least one dimension smaller than or equal to a selected distance necessary for adequate diffusion of a fission product. For example, the grains 106 of the nuclear fuel 100 may have an average characteristic length along a selected dimension of the grains 104 of the nuclear fuel. The average length may be selected to maintain adequate diffusion from the interiors of the grains 104 of the nuclear fuel 100 to the grain boundaries 112 of the grains of the nuclear fuel 100. It is recognized that there may exist a maximum average grain size 106 which will provide adequate diffusion of fission products 108 from the interiors 110 of the grains 104 to the grain boundaries 112 of the grains 104. In another embodiment, the grains 106 of the nuclear fuel 100 may have an average characteristic length 106 along a selected direction smaller than or equal to a selected distance necessary for adequate diffusion of a fission product. For example, the grains 106 of the nuclear fuel 100 may have an average characteristic length along a selected dimension of the grains 104 of the nuclear fuel. The average length along a selected direction may be selected to maintain adequate diffusion from the interiors of the grains 104 of the nuclear fuel 100 to the grain boundaries 112 of the grains of the nuclear fuel 100. It is recognized that there may exist a maximum average grain size along a selected direction 106 which will provide adequate diffusion of fission products 108 from the interiors 110 of the grains 104 to the grain boundaries 112 of the grains 104. In another embodiment, the grains 104 of the nuclear fuel may have a selected statistical distribution of characteristic lengths. For example, the grains 104 of the nuclear fuel 100 may have a grain size distribution having a selected percentage of the grains having a grain size 106 below a selected distance. For instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 75% of the grains have a grain size 106 equal to or less than 5 μm, with an average grain size of 3 μm. In another embodiment, the grains 104 of the nuclear fuel 100 may have multiple statistical distributions of characteristic lengths. For instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 25% of the grains have a grain size 106 equal to or less than 10 μm, 25% of the grains have a grain size 106 equal to or less than 5 μm, and 10% of the grains are below 1 μm. In another instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 25% of the grains have a grain size 106 equal to or less than 10 μm and 25% of the grains have a grain size 106 equal to or greater than 50 μm. In another instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 25% of the grains have a grain size 106 between 1 μm and 5 μm, 50% of the grains have a grain size between 5 μm and 10 μm, and 25% of the grains have a grain size 106 greater than 10 μm. It is further contemplated that the grain sizes 106 may be spatially distributed throughout the volume 102 of the nuclear fuel. For example, the average grain size 106 of grains within a first region may be selected to be greater or less than the average grain sizes 106 with a second, third, or up to and including an Nth region. Moreover, it is also contemplated herein that the spatial grain size 106 distribution may be continuous or discrete in nature. For example, in a cylindrical fuel pellet, the grains 104 may be engineered such that the grains are on average smallest at the center of the pellet and monotonically increase in size along the radial direction towards the pellet's surface. In another example, the grains 104 within a cylindrical fuel pellet may be distributed such that multiple discrete grain size zones exist within the pellet, with each zone containing grains with a selected average grain size 106. For instance, the central grain zone may have a first average grain size (e.g., 10 nm), a first concentric ring zone around the central zone may have a second average grain size (e.g., 100 nm), and a second concentric ring zone around the first concentric ring zone may have a third average grain size (e.g., 1 μm). It may be advantageous to have grain sizes in a central region of a cylindrical fuel pellet to have smaller grain sizes the outer pellet regions as the central region may experience larger fission process activity, and may require a larger degree of fission product 108 diffusion in order to avoid swelling. In another embodiment, the maximum characteristic length 106 along one or more dimensions of one or more grains 104 may be selected based on an operation condition of the nuclear fuel 100. For example, the operational condition of the nuclear fuel 100 may include the temperature a nuclear fuel 100 is utilized in a nuclear reactor system. For example, the higher the operational temperature of the nuclear reactor fuel 100 the smaller the average grain size 106 must be in order to provide adequate fission product 108 diffusion from the grain interiors 110 to the grain boundaries 112. In another example, the operational condition of the nuclear fuel 100 may include a thermally induced pressure within the nuclear fuel. For instance, as the nuclear fuel 100 thermally expands into a cladding structure housing the nuclear fuel 100, the interaction between the fuel surface 101 and the cladding may induce a stress within the nuclear fuel 100. In another embodiment, the maximum characteristic length 106 along one or more dimensions of one or more grains 104 may be selected based on the chemical composition of the nuclear fuel 100. For example, in the case of uranium-zirconium (UZr) and uranium-plutonium-zirconium (U—Pu—Zr) alloys the average grain size 106 required to provide adequate fission product 108 diffusion from the interiors 110 of the grains 104 to the grain-boundaries 112 may be dictated by the relative zirconium content in the UZr or U—Pu—Zr alloy. Zirconium is used as an alloying agent in metallic nuclear fuels in order to stabilize the phases (e.g., stabilize the migration of constituent materials) of metallic nuclear fuels. Moreover, in the case of U—Pu—Zr, for example, past studies by D. L. Porter et al. have indicated that migration of constituent materials does not occur for Pu concentrations of less than 8 wt. percent during irradiation. In the context of cylindrical fuel pellets, in U—Pu—Zr alloys with which migration of constituent materials does occur, it is recognized that the constituent materials tend to migrate to multiple radial zones within the cylindrical pellet, with Zr tending to migrate radially outward toward the cylindrical fuel pellet surface. Due to this outward migration, the central zone of a cylindrical U—Pu—Zr pellet may develop depleted Zr concentrations. This shifting in relative concentration may have large effects on fission product 108 production as well as diffusivity within a given region of the pellet. Therefore, the average grain size 106 required to ensure adequate diffusion from the grain interiors 110 to the grain boundaries 112 within a nuclear fuel 100 will depend upon the chemical composition and the geometric arrangement of the constituent materials of the given nuclear fuel 100. Moreover, the fission product generation rate of a given fuel may dictate the maximum allowable average grain size 106 required to ensure adequate diffusion from a grain-interior 110 to a grain-boundary 112 in one or more grains 104 of a nuclear fuel 100. The fission product generation rate is proportional to the fission rate within a given nuclear fuel 100. The fission rate within the given fuel is dependent upon, among other things, the fissionable materials implemented to form the nuclear fuel 100 and their relative concentration. In another embodiment, the maximum allowable characteristic length 106 along one or more dimensions of one or more grains 104 may be selected based on a desirable fission product concentration level. For example, the characteristic length 106 may be selected such that it is smaller than a critical distance with which fissiongas 118 nucleation occurs. In this manner, the characteristic length 106 may be selected such that the average grain size 106 of the nuclear fuel 100 is small enough to limit the fission product 108 concentration and as a result limit the fission gas 118 nucleation within the nuclear reactor fuel 100. It should be appreciated by those skilled in the art that the fission product generation rate, the chemical composition, and the temperature of implementation are intimately related quantities within a given nuclear fuel 100. For this reason, the exact evolution of fission product production is highly dynamic and may depend precisely on quantities, such as, but not limited to, the relative proportions of material constituents of the nuclear fuel 100, the geometry of the nuclear fuel 100, the operating temperature of the nuclear fuel 100, the density of the nuclear fuel 100 and the nuclear reactor type. It is, therefore, contemplated herein that any implementation of the nuclear fuel 100 of the present invention may rely on a trial and error method (e.g., using trial and error utilizing nuclear reactor or utilizing simulated nuclear reactor conditions) or any computational modeling process known in the art suitable to determine a maximum grain size 106 for a selected fuel composition parameters (e.g., type of fissionable material, relative concentration of constituent fissionable materials, geometrical distribution of fissionable material, density, or size of fuel piece) and nuclear reactor system parameters (e.g., type of reactor, temperature of operation, type of fuel material piece (e.g., fuel rod, fuel pellet, fuel pebble, or the like). For a detailed description of nuclear fuel swelling, fission product generation, and constituent material distribution and migration in U—Pu—Zr systems, see D. L. Porter et al., “Fuel Constituent Redistribution during the Early Stages of U—Pu—Zr Irradiation,” Metallurgical Transactions A, Vol. 21A, July 1990 p. 1871; and G. L. Hofman et al., “Swelling Behavior of U—Pu—Zr Fuel,” Metallurgical Transactions A, Vol. 21A, July p. 517 (1990), the disclosures of which are incorporated herein by reference. Referring again to FIGS. 1A and 1B, the one or more transportation pathways 116 of the boundary network 114 of the nuclear fuel 100 may be defined by a region between two or more adjacent grains 104. For example, as shown in FIG. 1B, the grain-boundary 112 between adjacent edges of neighboring grains 104 may define a transportation pathway 116 of the boundary network 114 of the nuclear fuel material 100. Referring now to FIG. 1I, the formation of a transportation pathway 116 of the boundary network 114 of the nuclear fuel 100 is illustrated. In one embodiment, the transportation pathway 116 between adjacent grains 104 of a nuclear fuel 100 may be formed via growth of an open bubble 150 of a fission gas 118 along the grain boundary 112 between adjacent grains 104. For example, in a first step 138, fission gas bubbles 144 begin to nucleate along a grain boundary 112 between two adjacent grains 104. The bubbles 144 are referred to herein as “closed” bubbles as they represent closed spherical voids within the nuclear fuel material. As discussed previously in the present disclosure, as grains 104 within a nuclear fuel 100 are reduced in size the grain-boundary area/grain interior volume ratio increases. The increase in the boundary area/interior volume ratio may lead to a relative increase in the number of fission gas bubbles nucleated at the grain boundary 144 and a relative decrease in the fission gas nucleation bubbles within the interior 146 of a given grain 104, during a fission process. Further, in step 140, as more and more fission gas bubbles continue to nucleate at a given grain-boundary the closed bubbles begin to coalesce and connect with one another to form a coalesced closed bubble structure 148. Then, in step 142, due to surface diffusion the coalesced closed bubbles 148 fully transform into an “open” bubble structure 150. As a result of the diffusion of fission gas atoms to the grain-boundary 112, a denuded region 152 within the interior 110 of the grain 104 is formed near the grain-boundary 112. The formation of an open bubble 150 forms the transportation pathway 116, defined on its edges by the grain-boundaries 112 of adjacent grains 104. If the transportation pathway 116 formed by the open bubbles (i.e., cracks) extends to the geometrical surface 101 of the nuclear fuel 100 then the fission gas may escape the volume 102 of the nuclear fuel 100. The migration of fission gas bubbles in irradiated uranium dioxide is generally described in Mary Ellen Gulden, “Migration of gas bubbles in irradiated uranium dioxide,” Journal of Nuclear Materials, Vol. 23, Issue 1 July pp. 30-36 (1967), which is incorporated herein by reference. In another embodiment, a plurality of transport pathways 116 may form a system of interconnected pathways 114. For example, as previously described, as the grain size 104 decreases within the nuclear fuel 100 the spatial density of grain-boundaries, and therefore transportation pathways 116, within the nuclear fuel 100 increases. An increase in transportation pathway density serves two purposes. First, number of transportation pathways that intersect the geometric surface 101 of the volume 102 of the nuclear fuel 100 will increase as the number of transportation pathways 116 increases within the nuclear fuel 100. As a result of the increase in transportation pathways 116 intersecting with the geometric surface 101 of the nuclear fuel 100, the amount of fission gas that may be transported via the boundary network 114 from the grain-boundaries 104 of the grains 104 increases. Second, the likelihood that a given transportation pathway 116 will intersect with another transportation pathway 116 will increase as the transportation pathway density increases with the nuclear fuel 100. Thus, a reduced grain size 106 in the grains 104 of the nuclear fuel 100 may lead to an increase in the number of transportation pathways 116 open to the geometric surface 101 and an increase in the frequency of interconnection between the multiple transportation pathways 116, both facilitating efficient fission gas transport from the grain-boundaries 112 to the geometric surface 101. It is further contemplated that the transportation pathways 116 of the interconnected boundary network 114 may be formed or their growth may be facilitated utilizing a volatile precipitating agent. For example, a volatile precipitating agent may be added to a metallic 124 or ceramic nuclear fuel material 128 prior to a casting process. During casting, a heat treatment (e.g., annealing process) may be applied to the nuclear fuel material. The heat treatment may cause the precipitating agent to precipitate out to the grain-boundaries 112 of the nuclear fuel 100. If large enough concentrations of the precipitating agent are present within the pre-cast nuclear fuel the precipitation of the precipitating agent may act to form one or more void regions within the nuclear fuel 100. Moreover, the precipitating agent may form a plurality of interconnected void regions within the nuclear fuel 100 which act to form the boundary network 114 of the nuclear fuel 100. It should also be recognized that the utilization of a precipitating agent may facilitate the growth of the boundary network 114 along the grain-boundaries 112 within the nuclear fuel 100. The precipitating agent may include, but is not limited to, nitrogen or carbon. In addition, it is contemplated herein that the grain-boundary 112 formation of the nuclear fuel 100 may be manipulated utilizing a precipitating metal agent to the nuclear fuel material prior to casting. For example, a metallic precipitating agent (e.g., niobium) may be added to a metallic fuel material 124 (e.g., uranium-zirconium) prior to a casting process. It is recognized that at a threshold metal precipitating agent concentration, upon cooling, the metal precipitating agent may precipitate out of the metal nuclear fuel material 124. It should further be recognized that the amount of metal precipitating agent which precipitates out of the nuclear fuel material upon cooling may depend on the cooling rate. As a result of precipitation, upon solidification, the metallic precipitating agent may form an additional phase within the nuclear fuel 100. For example, the metallic precipitating agent may form a distribution of solid regions of the metallic precipitating agent within the nuclear fuel 100. These solid metallic precipitating agent regions may facilitate the growth of the one or more grain-boundaries at the location of the metallic precipitating agents. It is further recognized that fission gas pressure that develops within the boundary network 114 as a result of fission gas diffusion from the grain interior may facilitate fission gas release due to grain-boundary 112 fracture (i.e., cracking). Grain-boundary 112 fracture may increase the boundary network 114 area, allowing for the boundary network to more readily transport fission gas to the nuclear fuel surface 101. It is also further recognized that the addition of a precipitating agent may facilitate the grain-boundary 112 fracture as the precipitating agent pressure at the grain-boundary may act to hasten the grain-boundary 112 fracture. It is further contemplated that the boundary network 114 may be formed by a plurality of void regions. While the above description generally relates to the formation of a boundary network 114 defined by the region between grain-boundaries 112, developed via fission gas nucleation at one or more grain-boundaries 112, it is recognized that any plurality of void regions outside of the one or more grain interiors 110 may lead to formation of a boundary network 114. For example, as will be further described herein, dispersant particles (e.g., zirconium oxide particles) may be dispersed throughout the nuclear fuel 100 along the grain-boundaries 112. The dispersant particles may act to create preferential fission gas occupation sites. If the gas occupation sites are distributed within the nuclear fuel 100 in a manner that provides for an interconnection of the bubbles formed at these gas occupation sites a boundary network 114 may be formed. Moreover, in a general sense, any method known in the art suitable for controlling porosity within the nuclear fuel 100 (e.g., metal nuclear fuel or ceramic nuclear fuel) may be utilized in order to create or facilitate the creation of a boundary network 114. It should be recognized that the boundary network 114 of the nuclear fuel 100 of the present invention may be formed prior to or during a nuclear fission process within the nuclear fuel 100. For example, as described above, the nuclear fuel 100 of the grain structure of the nuclear fuel 100 may be configured to develop a boundary network 114 upon production of fission products 108 (e.g., fission gas) during utilization of the nuclear fuel 100 within a nuclear reactor setting. In this manner, the nuclear fuel 100 may have an average grain size 106 below a critical value necessary for providing adequate diffusion of fission products 108 to the grain boundaries 112 of the nuclear fuel 100. Then, when the nuclear fuel 100 undergoes fission in the nuclear reactor 100, the fission products 108 nucleate more readily at the grain-boundaries 112, ultimately forming an interconnected boundary network 114. In another example, as described above, the boundary network 114 may be substantially formed prior to utilization in a nuclear reactor system. For instance, utilization of precipitating agents during a casting and annealing process may produce a boundary network 114 in the nuclear fuel 100. In another instance, any known void forming or porosity control process may be implemented during the fabrication of the nuclear fuel 100 in order to develop a boundary network 114 adequate to transport fission products, such as fission gas, from the grain-boundaries 112 of the nuclear fuel 100 to the geometric surface 101 of then nuclear fuel 100. Referring now to FIG. 1J, one or more grains 104 of the nuclear fuel 100 may include an interfacial layer 154. For example, one or more processes (e.g., chemical process or annealing process) may be implemented in order to grow an interfacial layer 154 a grain-boundary 112 of one or more grains 104. For instance, the formation of an interfacial region 154 may inhibit grain growth within the nuclear fuel 100 upon crystallization during a casting process. In this manner, an interfacial region may aid in maintaining the grain sizes 106 of the grains at or below the critical size necessary to maintain adequate diffusion of a fission product 108 from the grain interiors 110 to the grain-boundaries 112 of the nuclear fuel 100. For example, the interfacial region 154 may include an oxide layer, a nitride layer, or a carbide layer. For instance, nitrogen or carbon within the nuclear fuel material may precipitate out of the fuel matrix during a heat treatment process prior to casting. The nitrogen or carbon precipitates may then, upon cooling, form a metal nitride or metal carbide layer, respectively, at the surface of the crystallized grain structures. In another example, an oxygen atmosphere may be applied during, the casting phase of the fabrication process in order to form a metal oxide interfacial layer 154 at the grain-boundaries 112 of the nuclear fuel 100. It is contemplated herein that the thickness of the interfacial layer 154 may be a fraction of the size of the grain 104 with which it is grown. For instance, the interfacial layer 154 may have a thickness between 0.1 and 10 nm, whereas the interior grain structure may have a thickness between 0.1 and 10 μm. By way of another example, in the case of metallic nuclear materials, the interfacial region may include an intermetallic, which may be different in chemical composition from the interior 110 of the grains 104. In a general sense, any treatment process known in the art suitable for growing a grain-boundary interfacial layer 154 may be implemented in accordance with the present intention. The above description pertaining to interfacial layer 154 growth is illustrative only and should not be interpreted as a limitation. In another embodiment, the one or more processes implemented in order to achieve a characteristic length 106 along at least one dimension in some of the grains 104 required for adequate fission gas diffusion and the corresponding boundary network of the nuclear fuel 100 may include one or more material processing techniques. A variety of material processing techniques may be implemented in order to control the grain sizes 106 and the development of the boundary network 114 of the nuclear fuel 100. For example, the nuclear fuel 100 may be processed utilizing a cold-working process, an annealing process, a normalization process, or a tempering process. It should be recognized that the above list of material processing techniques is not exhaustive and should not be interpreted as a limitation as a variety of other material processing techniques may be suitable for fabricating the nuclear fuel 100 of the present invention. In one aspect, the grains 104 of the nuclear fuel 100 may be engineered to have a characteristic length 106 smaller than or equal to a selected distance along one or more dimensions utilizing one or more material processing techniques. In one embodiment, a cold-working process may be utilized to produce grains 104 within the nuclear fuel 100 having a characteristic length 106 smaller than or equal to a selected distance along one or more dimensions. It is recognized that the grain sizes 106 of a nuclear fuel 100 material may be reduced through a plastic deformation process that may occur when a given volume of a nuclear fuel 100 is cold-worked. For example, a solid consolidated metal 125 (e.g., uranium, plutonium or thorium) or metal alloy 126 (e.g., uranium zirconium, uranium zirconium hydride, uranium aluminum, or the like) nuclear fuel piece may undergo a cold-working process in order to reduce the grain sizes 106 of the nuclear fuel material grains 104, thereby shifting the average grain size of the material to smaller values. For instance, the solid metal 125 or metal alloy 126 nuclear fuel piece may include a cast metal or metal alloy nuclear fuel piece, such as a fuel rod. The cast metallic nuclear fuel may then be processed utilizing a cold-working process. For example, the cast metallic nuclear fuel piece may be cold-worked at a temperature below its recrystallization temperature (e.g., room temperature). The metallic piece may be cold-worked until the average grain size of the nuclear fuel material is at or below the size necessary to provide adequate diffusion of a fission product 108 to the grain boundaries 112 of the grains 104 of the material. For instance, a uranium-plutonium-zirconium fuel rod may be cold-worked until the average grain size 106 within the fuel rod is approximately 1 μm. It is further contemplated that a metallic fuel rod may be fabricated utilizing an extrusion process performed at ambient temperatures. Extruding the metallic fuel material at room temperatures provides the necessary plastic deformation required for reduction of grain sizes 106 within the material. As a result, extrusion of the metallic nuclear fuel material at room temperate may create a cold-worked grain structure, wherein the grain sizes 106 of the material are below the critical size required for adequate fission product 108 diffusion. Further, the room temperature extruded metallic fuel rod may then be annealed at a low recrystallization temperature in order to achieve the desired grain size within the material. It should be noted that if room temperature fuel rod extrusion is not possible an extrusion process may also be performed at a temperature low enough to inhibit recrystallization and grain growth, but high enough to allow for fuel road extrusion. It should be recognized that any cold-working process known in the art may be implemented to reduce the average grain size within a metal 125, metal alloy 126, or intermetallic 127 nuclear fuel. For example, a compression process, a bending process, a drawing process, an extrusion process, a forging process or a shearing process may be applied to a metal 125, a metal alloy 126, or an intermetallic 127 nuclear fuel material at a selected temperature below the material's recrystallization temperature. It should be recognized that the above cold-working processes do not represent limitations and should be interpreted as illustrations as it is contemplated that a variety of cold-working methods and conditions may be applicable in other contexts. Moreover, it should be recognized that a cold-working process may be applied to metal 125, a metal alloy 126, or an intermetallic 127 nuclear fuel irrespective of prior casting. The description of casting and extrusion above is merely for illustrative purposes and should not be interpreted as a required limitation prior to the cold-working of a metal or metal alloy nuclear fuel material in order to reduce the average grain of the material below a size required for adequate diffusion of fission products 108. It is contemplated that a variety of other metal 125, a metal alloy 126, or an intermetallic 127 nuclear fuel piece fabrication methods may be implemented within the context of the present invention. By way of another example, a thorium or thorium alloy may be cold-rolled in order to form a fuel piece suitable for implementation in a nuclear reactor setting. Thorium or a thorium alloy is particularly useful in the context of cold-rolling processing due to its high level of ductility. Utilizing a cold-rolling process allows for control of the average grain size of a rolled thorium or thorium alloy fuel piece without a prior process step, such as casting. Thus, a cold-rolling process may be implemented in a manner which controls the grain size distribution of the grains of the thorium or thorium alloy piece as the piece is formed into a fuel rod. For instance, a solid piece of thorium may be cold-rolled into a thin planar sheet, wherein the grain sizes 106 within the sheet are below the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112. This grain-engineered sheet may then be further manipulated by rolling the sheet into a cylindrical or pellet shape. In another example, a solid consolidated ceramic nuclear fuel piece may undergo a cold-working process in order to reduce the average grain size of the grains within the ceramic nuclear fuel material. The solid ceramic nuclear fuel piece may be fabricated utilizing any ceramic nuclear fuel fabrication process known in the art. For instance, the ceramic nuclear fuel piece may be fabricated by compacting and pressing a ceramic nuclear fuel powder (e.g., uranium dioxide powder), or a precursor of a nuclear fuel powder (e.g., U3O8), into a fuel pellet or fuel pebble. For example, an organic binder agent may be added to the nuclear fuel powder prior to pressing. After pressing the powder and binder mixture into a desired shape, the binder may be evaporated off using a high temperature treatment, wherein the ceramic piece is heated above the organic agent boiling point but below the ceramic nuclear fuel melting point. The compacted nuclear fuel powder may then be sintered to a selected density, up to 98% of the theoretical density. The compacted ceramic nuclear fuel material may then be processed utilizing a cold-working process, such as a compression process. The ceramic piece may be cold-worked until the average grain size of the nuclear fuel material is at or below the size necessary to provide adequate diffusion of a fission product 108 to the grain boundaries 112 of the grains 104 of the ceramic material 128. The cold-working processes described above are generally suitable for implementation in the context of cold-working a ceramic nuclear fuel piece. While cold-working is often difficult to implement in the context of ceramic materials, due to their brittle nature, it is contemplated herein that cold-working processes, such as those described above may be implemented to control the average grain size of ceramic nuclear fuel material. The cold-working of ceramic materials is generally described in David W. Richerson, Modern ceramic engineering: properties, processing, and use in design, 3rd ed, CRC Press-Taylor & Francis Group, 2003, pp. 235-240, which is incorporated herein by reference. The description of ceramic nuclear fuel material sintering above should not be interpreted as a required limitation of the present invention, rather sintering is but one method used to create a ceramic nuclear fuel piece suitable for implementation in the present invention. It is contemplated that a variety of other ceramic nuclear fuel piece fabrication methods (e.g., casting, in-situ reaction, injection molding or the like) may be implemented within the context of the present invention. Moreover, the above description of uranium dioxide as a material suitable for cold-working should not be interpreted as a limitation as any ceramic nuclear fuel material including, but not limited to, oxides, carbides, and nitrides may be implemented in this context. In a further embodiment, an annealing process may be implemented in order to achieve the desired grain size 106 within a nuclear fuel material. For example, after cold-working a metallic nuclear fuel material (e.g., cold-working a cast piece or extrusion of material at room temperature or low temperature) an annealing process may be utilized in order to achieve the desired average grain size 106 within the metallic nuclear fuel material. It should be recognized by those skilled in the art that after introducing cold-work into a given material a subsequent anneal at temperatures below the recrystallization temperature may result in a refinement of the grains of the material. For example, after extruding or applying another cold-working process to a metallic nuclear fuel piece, the metallic nuclear fuel 125 may be annealed to a low temperature below the recrystallization temperature in order to further refine the grains 106 of the nuclear fuel material. In order to facilitate the production of smaller grain structures in the nuclear fuel 100, the temperature at which the subsequent annealing processes takes place should be above the temperature at which the recovery phase of the cold-worked metallic nuclear fuel material initiates. Moreover, it should also be recognized that the recrystallization temperature is a function of the amount of cold-work introduced into the nuclear fuel 100. In another embodiment, an annealing process may be implemented in order to increase the grain size 106 of the grains of the nuclear fuel 100. For instance, the room temperature extrusion process may result in an average grain size within a material that is smaller than the target average grain size. An annealing process may then be implemented in order to grow the average grain size to the target level. It should be noted that the target grain size described herein is below the critical size necessary to achieve adequate diffusion within the metallic nuclear fuel 100, but smaller than the required size for other purposes (e.g., achieving a target material density, a target porosity, and the like). Generally speaking, an annealing process at temperatures above the nuclear fuel material's recrystallization temperature may be implemented in order to achieve the desired grain size after implementation of any cold-working process known in the art or described herein. The annealing temperature, the annealing rate, and the soak time may be selected based on the requirements of the specific material in use and the amount of cold-work previously introduced into the system. In another embodiment, an annealing process may be implemented in order to achieve the desired grain size 106 within a ceramic nuclear fuel material. For example, upon cold-working of a ceramic nuclear fuel material an annealing process may be utilized in order to achieve the desired average grain size 106 within the ceramic nuclear fuel material 128. For instance, the cold-working process may result in an average grain size within a material that is smaller than the target average grain size. An annealing process may then be implemented in order to grow the average grain size to the target level. Generally speaking, an annealing process may be implemented in order to achieve the desired grain size after implementation of any cold-working process suitable for ceramic material processing known in the art or described herein. The principles of annealing, recovery, and recrystallization are generally described in F. J. Humphreys and M. Hatherly, Recrystallization and Related Annealing Phenomena, 2nd ed, Elsevier, 2004, which is incorporated herein by reference. It should be recognized that, in the context of a metallic nuclear fuel material, an annealing temperature should be selected well below the melting temperature of the metallic nuclear fuel. For example, in metal alloy nuclear fuel materials such as U—Pu—Zr and U—Pu, a spatial redistribution of materials may occur upon annealing. Implementing an annealing temperature that is too near the melting temperature of the metallic fuel may exacerbate this redistribution of materials. For instance, upon heating above the melting temperature an existing thermal gradient within the material may lead to a redistribution of Pu in either the U—Pu—Zr or U—Pu alloys. A redistribution of Pu may lead to an altered temperature profile within the fuel during implementation in a nuclear reactor with higher temperature readings at the redistributed Pu sites. Therefore, metallic nuclear fuels should undergo heat treatment (e.g., annealing, normalization, tempering and the like) at a temperature low enough to minimize material redistribution within the nuclear fuel material. In another embodiment, a normalization process may be utilized to engineer grains 104 within the nuclear fuel 100 to have a characteristic length 106 smaller than or equal to a selected distance along one or more dimensions. For example, after a cold-worked nuclear fuel material has undergone a heat treatment process (e.g., annealing), the material may then be cooled in air. This process may relieve stress in the material and may result in reduced grain sizes 106 with the nuclear fuel 100. For instance, a metal 125 or metal alloy 126 nuclear fuel piece may be formed via a casting process. After the casting process, the metallic nuclear fuel material piece may be heated to a temperature above its upper critical point. The metallic nuclear fuel material piece material may then be held at the elevated temperature for sufficient time to allow the production of smaller grains within the material. Then, the material may be cooled in air to a temperature well below the critical point. A normalization process may lead to a reduction in the average grain size in the nuclear fuel 100 at or below the average grain size required to maintain adequate fission product 108 diffusion within the material. In another embodiment, a tempering process may be utilized to engineer grains 104 within the nuclear fuel 100 to have a characteristic length 106 smaller than or equal to a selected distance along one or more dimensions. It is recognize that any known tempering process is suitable for implementation in the context of the present invention. In another embodiment, the one or more processes implemented in order to achieve a grain size 106 required for adequate fission product 108 diffusion and the corresponding boundary network 114 of the nuclear fuel 100 may include one or more chemical treatment process. In one embodiment, a chemical process utilized to reduce grain size 106 and develop the boundary network 114 in the nuclear fuel 100 material may include, but is not limited to, an oxygen reduction process. For example, in the case of an oxide based nuclear fuel material, such as UO2 or PO2, an oxygen reduction process may be applied to the metal oxide fuel utilizing a reduction gas. By chemically reducing a given metal oxide nuclear fuel into a sub-stoichiometric state, the average grain size 106 of the metal oxide nuclear fuel may be reduced in size relative to the stoichiometric phase. For instance, exposing a UO2 based nuclear fuel 100 to a reducing gas consisting of an argon/hydrogen mixture may reduce the uranium oxide to a sub-stoichiometric phase, such as, but not limited to, UO1.8. It should be recognized by those skilled in the art that an oxygen reduction to a sub-stoichiometric state may “shrink” the exposed grains. It is recognized that an oxygen reduction process may be implemented in order to further develop the boundary network 114 as a result of the increased grain-boundary area which results when adjacent grains 104 shrink. It is contemplated that 8 to 16% mixture of argon to hydrogen should be suitable for reduction. Moreover, a reducing gas consisting of nitrogen and hydrogen may also be suitable for implementation in the present invention. It should be further recognized by those skilled in the art that non-sintered UO2 may often solidify into a hyperstoichiometric state. As such, a subsequent oxygen-reducing treatment as described above may be implemented to reduce the hyperstoichiometric UO2 to a stoichiometric or sub-stoichiometric state. In another embodiment, the porosity of the nuclear fuel 100 may be controlled via a porosity control process. For example, a porosity control process may be implemented to establish or further develop the boundary network of the nuclear fuel 100. For instance, porosity of the nuclear fuel 100 may be controlled during a compacting and sintering process, wherein porosity may be controlled via the compaction parameters (e.g., pressure, binder agent concentration, temperature, and the like). In another embodiment, the textures of two or more of the grains 104 within the nuclear reactor fuel 100 may be controlled via a grain texture control process. Any grain texture control process in the art is suitable for implementation in the context of the present invention. For example, an annealing process may be used to at least partially impart grain texture into the grain structure of the grains 104 of the nuclear fuel 100. In another example, a shear deformation process (e.g., shear rolling) may be used to impart grain texture into the grain structure of the grains 104 of the nuclear fuel 100 It is further contemplated that the grain sizes 106 of the grains 104 and the boundary network 114 of the nuclear fuel 100 may need not be achieved upon fabrication in a fabrication facility setting. Rather, it is contemplated herein that the required grain structure and boundary network 114 of the nuclear fuel 100 of the present invention may be established upon initiation of a fission process during implementation in a nuclear reactor setting. For example, the high temperature of the nuclear reactor environment may result in an annealing effect in the nuclear fuel 100. In another example, when the grain sizes 106 of the grain structure are properly configured, which is an object of the present invention, the irradiation leading to fission product 108 production within the nuclear fuel 100 may lead to a further development of the boundary network 114 While the above description relates to the material processing of a macroscopic piece of nuclear fuel material, it is further contemplated that the grain sizes of microscopic particles and the corresponding boundary network may be controlled utilizing a variety of material processing techniques. It should be recognized that the creation of the boundary network 114 in the nuclear fuel 100 of the present invention is intimately related to the control of the average grain size 106 of the nuclear fuel 100. For example, as the average grain size 106 is reduced in a given nuclear fuel material, the spatial density of grain-boundaries 112 increases, thereby increasing the relative proportion of the boundary network 114 area to the volume 102 of the nuclear fuel 100. As a result, as the average grain sizes of the nuclear fuel 100 decrease, the number of boundary network pathways 116 intersecting the geometric surface 101 of the nuclear fuel increases. Therefore, any of the material processes described in the present disclosure to control the grain sizes 106 of the nuclear fuel 100 may also be implemented in order to control the extent of the boundary network 114 of the nuclear fuel 100 of the present invention. For example, just as a cold-working process may be used to control the average grain size 106 within a metal nuclear fuel 125 or metal alloy nuclear fuel 126, a cold-working process may be utilized to control the growth of the boundary network 114. It is recognized, however, that in some instances a user may achieve adequate average grain size 106 within a given nuclear fuel 100 (i.e., size required to ensure adequate diffusion of a fission product 108 from the interior 110 of a grain 104 to its grain-boundary 112) without necessarily achieving adequate boundary network 114 development within the nuclear fuel 100 (i.e., network density and interconnectedness required to ensure transport of the fission product 108 to the fuel's geometric surface 101). In this instance, the average grain size 106 of the nuclear fuel 100 may be further reduced in order to achieve the adequate grain-boundary density and likelihood of interconnectedness within the fuel to achieve adequate transport of a fission product 108 from the grain boundaries 112 of the grains to the geometric surface 101 of the nuclear fuel 101. In another instance, an average grain size 106 within a given nuclear fuel 100 required for adequate diffusion of a fission product 108 from the interior 110 of a grain 104 may be achieved utilizing a first process, such as cold-working. Then, the boundary network 114 may be further developed utilizing a second process, such as an oxygen reduction step, utilizing a forming gas ambient, such as a hydrogen/argon mixture. In a general sense, a first material process step may be utilized to achieve a first level of reduction in the grain sizes 106 of the nuclear fuel 100, while a second material process step may be utilized to further reduce the grain sizes 106 in order to further develop the boundary network 114 of the nuclear fuel 100. It is further contemplated herein that the nuclear fuel of the present disclosure may be configured to operate in a variety of nuclear reactor system contexts. For example, the nuclear fuel 100, 200, 300, and 400 of the present invention may be utilized in a thermal spectrum nuclear reactor, a fast spectrum nuclear reactor, a multi-spectrum nuclear reactor, a breeder nuclear reactor, or a traveling wave reactor. It is contemplated herein that the previously provided disclosure of the nuclear fuel 100 and the various methods and processes utilized to make the nuclear fuel 100 should be considered to extend to the remainder of the disclosure. Referring now to FIGS. 2A and 2B, alternative embodiments of the present invention are illustrated. A nuclear fuel 200 and methods for making a nuclear fuel are described in accordance with the present disclosure. A given volume 202 of a nuclear fuel 200 may include a plurality of nuclear fuel elements 204 of one or more nuclear fuel materials. In one embodiment, the nuclear fuel elements 204 may be fabricated using one or more metallic nuclear fuel materials 124. In another embodiment, the nuclear fuel elements 204 may be fabricated using one or more ceramic nuclear fuel materials 128. The nuclear fuel elements 204 may be engineered to have a characteristic length 206 along at least one dimension that is smaller than or equal to a selected distance. The distance may be selected based on the critical distance necessary to maintain an adequate level of diffusion of a fission product 108 (e.g., fission gas 118, fission liquid 119, or a fission solid 120) from the interior 210 of the nuclear fuel elements 204 to one or more free surfaces 212 of the nuclear fuel elements 204. As is the case in the grain structure context illustrated in FIGS. 1A through 1X, it is recognized that by increasing the ratio between the nuclear fuel element free surface area and the nuclear fuel element interior volume the number of fission gas 118 bubbles formed at the free surface 212 of a nuclear fuel element 204 as a result of fission gas diffusion from the nuclear fuel element interior 210 may increase. Therefore, by decreasing the size of nuclear fuel elements 204 of the nuclear fuel material, thereby increasing the ratio between free surface area and element interior volume, it is possible to enhance the diffusion of fission gas 118, or other fission products 108, from the interiors 210 of the fuel elements 204 to the free surfaces 212 of the fuel elements 204. As in the case with grain-boundary fission product 108 nucleation, the decrease in nuclear fuel element size 206 increases the likelihood of fission gas 118 nucleation at the free surface 212 of the nuclear fuel element 204, while simultaneously decreasing the likelihood of fission gas 118 nucleation within the fuel element interior 210. The nuclear fuel elements 204 may further be consolidated to a selected density. The selected density may be chosen to balance the power density requirements of the nuclear fuel 200 and the boundary network requirements necessary for fission product 108 migration to the geometric surface 201 of the nuclear fuel 200. Further, the nuclear fuel 200 may include a boundary network 214 configured to transport a fission product 108, such as a fission gas 118, from the free surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 to the external geometric surface 201 of a given volume 202 of the nuclear fuel 200. If a given fission gas bubble 118, or other fission product 108, has an open transportation pathway 216 to the geometric surface 201 of the nuclear fuel 200 then the fission gas bubble 118 may be released from the nuclear fuel material volume 202. As previously described above, the aggregated effects of fission gas 118 release across the entire volume 202 of the nuclear fuel 200 may result in a reduction or elimination of swelling in the nuclear fuel 200 upon implementation of the nuclear fuel 200 in an operational setting of a nuclear reactor. In addition to improving fission product 108 diffusion to the free surfaces 212 of the nuclear fuel elements 204, the engineering of reduced sized nuclear fuel elements 204 may also increase the spatial density of transportation pathways 216 of the boundary network 214 of the nuclear fuel 200, thereby increasing the likelihood of fission product 108 transportation from a free surface 212 to the geometric surface 201 of the nuclear fuel 200. In one embodiment, the nuclear fuel elements 204 of the nuclear fuel 200 may include one or more metallic nuclear fuel material 124, such as, but not limited to, a metal nuclear fuel material 125, a metal alloy nuclear fuel material 126, or an intermetallic nuclear fuel material 127. For example, a metal nuclear fuel material may include, but is not limited to, uranium-235 metal, plutonium-239 metal, or thorium-233 metal. In another example, a metal alloy nuclear fuel material 126 may include, but is not limited to, uranium-zirconium, uranium-plutonium-zirconium, uranium-zirconium-hydride, or uranium aluminum. By way of a further example, an intermetallic nuclear fuel material 127 may include, but is not limited to, UFe2 or UNi2. It should be recognized that the above list of suitable metallic nuclear fuel materials for inclusion in the nuclear fuel elements 204 of the nuclear fuel 200 of the present invention should not be interpreted as a limitation but rather merely an illustration. In another embodiment, the nuclear fuel elements 204 of the nuclear fuel 200 may include one or more ceramic nuclear fuel material 128, such as, but not limited to, an oxide nuclear fuel material 129, a nitride nuclear fuel material 131, or a carbide nuclear fuel material 132. For example, an oxide based nuclear material 129 may include, but is not limited to, uranium dioxide (UO2), plutonium dioxide (PuO2), or thorium dioxide (ThO2). Moreover, an oxide based nuclear fuel material 129 may include a mixed oxide nuclear fuel material, such as, but not limited to, a mixture of PuO2 and depleted or natural UO2. In another example, a nitride based nuclear fuel material 131 may include, but is not limited to, uranium-nitride or plutonium nitride. By way of a further example, a carbide base nuclear fuel material may include, but is not limited to, uranium carbide 132. It should be recognized that the above list of suitable ceramic nuclear fuel materials for inclusion in the nuclear fuel elements 204 of the present invention should not be interpreted as a limitation but rather merely an illustration. The fabrication of the nuclear fuel elements 204 for implementation in the nuclear fuel 200 of the present invention may include a variety of material processing techniques. In one embodiment, the nuclear fuel elements may be fabricated utilizing a ball milling process. For example, a ceramic material 128 or metallic material 124 or a pre-cursor thereof may undergo a ball milling process in order to fabricate a plurality of nuclear fuel elements 204 having a characteristic length 206 along a selected dimension. For instance, a uranium dioxide powder may undergo further ball milling (e.g., wet milling, dry milling, high energy ball milling or reactive ball milling) processing in order to achieve an average particle size within the uranium dioxide at or below the critical size necessary to provide adequate fission product 108 diffusion in the nuclear fuel's consolidated form. Ball milling processing is well known in the art and is capable of producing particle sizes over a large range of sizes. In some instances, ball milling has been shown capable of producing particles sizes as small as 1-5 nm. For example, a milling process may be applied to a volume of uranium dioxide powder for a sufficient time to produce particles in the size range 0.001 to 100 μm. It should be recognized that the above examples do not represent limitations but should merely be interpreted as illustrations. Those skilled in the art will recognize that there exists a variety of ball milling procedures applicable to a variety of materials and material conditions suitable to produce particle shaped nuclear fuel elements 204 for implementation in the present invention. The principles of ball milling metal and ceramic powders to sub-10 nm levels are generally described in A. S. Edelstein and R. C. Cammarata, Nanomaterials: Synthesis, Properties, and Applications, 1st ed, Taylor & Francis Group, 1996, which is incorporated herein by reference. The principles of high energy ball milling of oxide ceramics are generally described in S. Indris et al., “Nanocrystalline Oxide Ceramics Prepared by High-Energy Ball Milling,” Journal of Materials Synthesis and Processing, Vol. 8, Nos. 3/4 (2000), which is incorporated herein by reference. In addition to ball milling, the nuclear fuel elements 204 of the nuclear fuel 200 may be fabricated utilizing additional mechanical processing techniques. For example, a mechanical process suitable for fabricating nuclear fuel elements 204 having a reduced thickness in at least one dimension may be utilized to fabricate the nuclear fuel elements 204 of the nuclear fuel 200. For instance, a cold-rolling process may be utilized to fabricate planar thin sheets of a metal nuclear fuel material, such as thorium. The metallic nuclear fuel sheets may be cold-rolled to a thickness smaller than the critical distance required for adequate diffusion of fission products 108 from the interior of the sheets to the surface of the sheets. By way of another example, a drawing process may be utilized to fabricate thin wire structures of a metal nuclear fuel material. The metallic nuclear fuel wires may be drawn to a cross-sectional radius smaller than the critical distance required for adequate diffusion of fission products 108 from the interior of the wires to the surface of the wires. Those skilled in the art will recognize that there exist a variety of mechanical process techniques suitable for fabricating nuclear fuel elements 204 of the nuclear fuel 200 of the present invention. It should further be recognized that the wires and planar sheets described above do not represent limitations on the shape of mechanically shaped nuclear fuel elements 204 of the present invention and are merely illustrative in nature. In another embodiment, the nuclear fuel elements 204 of the nuclear fuel 200 may be fabricated utilizing a nanostructuring technique. For example, nanostructuring techniques may be implemented to form nanowires, nanotubes, nanorods, nanosheets, nanorings, or the like. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may include nanorods formed from the nanostructuring of a metal oxide based nuclear fuel material. For instance, nanorods of metal oxide materials have been fabricated to have thicknesses as small as 40 nm with lengths of 10 μm. The principles of metal oxide nanorod formation are generally described in U.S. Pat. No. 5,897,945, issued on Apr. 27, 1999, and is incorporated herein by reference. It should be recognized that the nuclear fuel elements 204 of the present invention may be fabricated in variety of manners. It should be further recognized that, based on the context of the nuclear fuel 200 implementation, one fabrication method may be superior to another method of fabrication. The key feature of the various fabrication methods is that they may provide a means for producing nuclear fuel elements 204 having a size equal to or smaller than a critical distance necessary for providing adequate diffusion of a fission product 108 in the condensed nuclear fuel 200. In further embodiments, one or more processes may be utilized in order to refine the size, shape, or other characteristic of the fabricated nuclear fuel elements 204 of the nuclear fuel. For example, one or more material processing techniques may be utilized to reduce the size of the nuclear fuel elements 204 along one or more dimensions. Further, one or more material processing techniques may be utilized to reduce the grain sizes within the nuclear fuel elements 204 along one or more dimensions The material processing techniques may include, but are not limited to, cold-working, annealing, tempering, normalizing, chemical treatment, mechanical treatment, irradiation, exposure to high temperature environment, porosity control, or texture control. The various applicable processes have been described previously herein. It should be recognized that the previous description of the above material processing methods may be extended to the processing of non-consolidated nuclear fuel elements 204 currently presented. In one embodiment, a portion of the nuclear fuel elements 204 of the nuclear fuel 200 may include nuclear fuel elements 204 having a three dimensional geometric shape. For example, the three dimensional geometric shaped nuclear fuel elements 204 may include regular or irregular shaped nuclear fuel elements. For instance, the nuclear fuel elements 204 may include, but are not limited to, a spherical element, a cylindrical element, an ellipsoidal element, a toroidal element, or a rhomboidal element. In another embodiment, some of the nuclear fuel elements 204 of the nuclear fuel 200 may include, but are not limited to, a particle nuclear fuel element, a linear nuclear fuel element, or a planar nuclear fuel element. For instance a particle nuclear element may include, but is not limited to, a spherical particle, a cylindrical particle, an ellipsoidal particle, or an irregular shaped particle. In another instance, the linear nuclear fuel element may include, but is not limited to, a cylindrically shaped wire, or a cylindrical shaped rod or rodlet. In an additional instance, a planar nuclear fuel element, may include, but is not limited to, a rectangular “sheet” shaped nuclear fuel element. Referring now to FIGS. 2C through 2E, the characteristic length 206 along at least one dimension of one or more nuclear fuel elements 204 may include a characteristic length 206 along all dimensions of one or more nuclear fuel elements 204 of the nuclear fuel 200. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may be engineered such that the “height”, represented by “a,” and “width,” represented by “b” are similar in size. Therefore, a fission product 108 may efficiently diffuse from the nuclear fuel element interior 210 to the nuclear fuel element surface 212 along all directions within the grain. In this context, a nuclear fuel element 204 may be characterized by the nuclear fuel element “size.” The nuclear fuel element size 206 may be selected such that the nuclear fuel elements 204 are small enough to allow for adequate diffusion from the interiors 210 of the one or more nuclear fuel elements 204 to the boundaries 212 of the one or more nuclear fuel elements 204. As shown in FIG. 2D, the characteristic length 206 along at least one dimension of one or more nuclear fuel elements 204 may include a characteristic length 206 along a selected dimension of one or more nuclear fuel elements 204. For example, as shown in FIG. 2D, the nuclear fuel elements 204 within the nuclear fuel 200 may be engineered to have a selected characteristic length 206 along a given dimension of the nuclear fuel elements 106. For instance, in the case of a nuclear fuel element 204 having an elongated structure, the nuclear fuel element 204 may have a selected characteristic length along the “thin” dimension, shown as dimension “a” in FIG. 2D, of the nuclear fuel element 204. In another instance, in the case of a nuclear fuel element 204 having an elongated structure, the nuclear fuel element 204 may have a selected characteristic length along the “thick” dimension, shown as dimension “b” in FIG. 2D, of the nuclear fuel element 204. It should be recognized that the nuclear fuel element 104 need only have at least one characteristic length 206 smaller than the distance required for adequate diffusion from the interiors 210 of the one or more nuclear fuel elements 204 to the boundaries 212 of the one or more nuclear fuel elements 204. It is further recognized, however, that all dimensions of a nuclear fuel element 204 may have a characteristic length 206 smaller than or equal to a distance required for adequate diffusion of fission product 108 from the interiors 210 of the one or more nuclear fuel elements 204 to the boundaries 212 of the one or more nuclear fuel elements 204. As shown in FIG. 2E, the characteristic length 206 along at least one dimension of one or more nuclear fuel elements 204 may include a characteristic length 206 along a selected direction 234. For example, the nuclear fuel elements 204 within the nuclear fuel 200 may be engineered to have a selected characteristic length 206 along a given direction in the nuclear fuel 200. For instance, a nuclear fuel element 204 having an elongated structure may have a selected characteristic length 206 along a selected direction 234 within the nuclear fuel 200. In another embodiment, one or more nuclear fuel elements 204 may have a characteristic length 204 along a dimension of the one or more nuclear fuel elements 204 selected to maximize heat transfer from a nuclear fuel element-interior 210 to a nuclear fuel element-boundary 212. For example, the one or more nuclear fuel elements 204 may be oriented such that their narrow dimensions, shown as “a” in FIG. 2E, are aligned substantially perpendicular to a thermal gradient 236 in the nuclear fuel 200. Such an arrangement aids in the heat transfer from the nuclear fuel element-interior 210 to the nuclear fuel element-surface 212, aiding in the diffusion of a fission product 108 from the nuclear fuel element interior 210 to the nuclear fuel element surface 212. By way of another example (not shown), in a spherical fuel pebble fabricated utilizing the nuclear fuel 200 of the present invention the nuclear fuel elements 204 of the nuclear fuel 200 may be arranged to have their the narrow dimension substantially perpendicular to the radial thermal gradient of the cylindrical pellet. It should be noted that the illustrations in FIGS. 2C, 2D, and 2E represent simplified conceptual illustrations of a plurality of nuclear fuel elements 204 consistent with the present invention and should not be interpreted as schematical in nature. Further, it should be recognized by those skilled in the art that a variety of materials processing techniques (e.g., cold-working and/or annealing, compression, or extrusion) previously and further described herein may be implemented in order to develop the symmetrical nuclear fuel element structure in FIG. 2C, and the deformed elongated nuclear fuel element structure illustrated in FIGS. 2D and 2E. In other embodiments, it is contemplated herein that the plurality of nuclear fuel elements 204 of the nuclear fuel 200 may include controllable statistical attributes, such as average sizes and statistical distributions (e.g., counting statistics and spatial distribution statistics) similar to the plurality of grains 104 of the nuclear fuel 100 described previously herein, which extends to the instant context. In other embodiments, it is contemplated that the critical distance required to ensure adequate diffusion of fission products 108 from the interior of the nuclear fuel elements 204 to the surface of the nuclear fuel elements may depend on a variety of conditions The conditions include, but are not limited to, operational conditions of the nuclear fuel 200 (e.g., operational temperature or temperature induced pressure within the nuclear fuel 200), the chemical composition of the nuclear fuel 200, the fission product generation rate, or the size required to inhibit fission product nucleation within the nuclear fuel 200. The description of these conditions in the context of the nuclear fuel 100 should be interpreted to extend to the instant context. Referring again to FIGS. 2A and 2B, a plurality of transportation pathways 216 may form a system of interconnected pathways 214. For example, as previously described, as the nuclear element size 204 decreases within the nuclear fuel 200 the spatial density of nuclear element surfaces 212, and therefore transportation pathways 216, within the nuclear fuel 200 increases. An increase in transportation pathway density serves two purposes. First, the number of transportation pathways that intersect the geometric surface 201 of the volume 202 of the nuclear fuel 200 will increase as the number of transportation pathways 216 increases within the nuclear fuel 200. As a result of the increase in transportation pathways 216 intersecting with the geometric surface 201 of the nuclear fuel 200, the amount of fission gas 118 that may be transported via the boundary network 214 from the nuclear fuel element surfaces 212 of the nuclear fuel elements 204 increases. Second, the likelihood that a given transportation pathway 216 will intersect with another transportation pathway 216 will increase as the transportation pathway density increases within the nuclear fuel 200. Thus, a reduced nuclear fuel element size 206 of the nuclear fuel 100 may lead to an increase in the number of transportation pathways 216 open to the geometric surface 201 and an increase in the frequency of interconnection between the multiple transportation pathways 216, both of which facilitate the efficient fission gas transport from the nuclear fuel elements 212 to the geometric surface 201. In one embodiment, the boundary network 214 of the nuclear fuel 200 may generally be controlled by controlling the porosity within the nuclear fuel 200. In a further embodiment, the porosity of the nuclear fuel 200 may be controlled by variation of the pressing and sintering parameters upon consolidation of the plurality of nuclear fuel elements 204 into a solid consolidated volume 202 of nuclear fuel 200. For instance, the robustness of the boundary network 214 may controlled by varying at least one of the group including pressing pressure, sintering temperature, sintering time, presence of reducing atmosphere, binding agent parameters. Therefore, during the fabrication of the nuclear fuel 200, the qualities of the boundary network 214 of the nuclear fuel 200 may depend, among other things, upon: nuclear fuel element size 204, binding agent mixture concentration, type of binding agent, compaction pressure, sintering temperature, annealing temperature, annealing time and nuclear fuel element chemical composition. It should be noted that this merely represents an illustrative list of parameters which may dictate the formation of the boundary network 214 of the nuclear fuel 200 in the context of sintering. It is further contemplated that a sintering and/or compaction process may be applied to the consolidation of either metallic nuclear fuel elements or ceramic nuclear fuel elements. The principles of sintering of metals are generally described in U.S. Pat. No. 4,992,232, issued on Feb. 12, 1991; and U.S. Pat. No. 2,227,177, issued on Dec. 31, 1940, which are incorporated herein by reference. The principles of sintering ceramics are generally described in U.S. Pat. No. 6,808,656, issued on Oct. 26, 2004; and U.S. Pat. No. 3,995,000, issued on Nov. 30, 1976, which are incorporated herein by reference. The principles of sintering uranium dioxide and precursors thereof in the presence of various atmospheres are described in J. Williams et al., “Sintering uranium oxides of composition UO2 to U3O8 in various atmospheres,” Journal of Nuclear Materials, Vol. 1, Issue 1 April pp. 28-38 (1959), which is incorporated herein by reference. It is contemplated herein that previously described aspects of boundary network 114 formation, such as facilitation of boundary network growth via fission gas 118 diffusion, formation of a boundary network via control of void region growth, or development of a boundary network via precipitation, within the nuclear fuel 100 should be interpreted to extend to the instant context. In one embodiment, the selected density of the nuclear fuel 200 may include a density less than the theoretical density of the nuclear fuel material. For example, the nuclear fuel elements 204 may be consolidated into a solid consolidated volume having a density of 70% of the theoretical density of the material. In another instance, the density may be 98% of the theoretical material density. In a general sense, there is no specific requirement for the nuclear fuel 200 density. Rather, the density should be selected on a case by case basis, depending on the specifics of implementation. The minimum density required is a function of the required power density of the nuclear fuel 200. Based on currently implement fuels, most modern day nuclear reactor systems require a fuel density of approximately 68% or greater, however, this should not be interpreted as a limitation. It is contemplated herein that the density of the nuclear fuel 100 may be significantly below 68% of the theoretical density of the material. For instance, the fuel density of the nuclear fuel 100 may be below 50% of the theoretical density of the material. The selected density may balance the power density requirements of the nuclear fuel 200 and the fission product transport requirements provided by an open boundary network 114. It is contemplated herein that the precise density utilized in a given application may be determined on a trial and error basis given the specific implementation or via a computer modeling technique. In one embodiment, the consolidated volume 202 of nuclear fuel 200 may take on a variety of shapes. For example, the nuclear fuel elements 204 may be consolidated and compacted and sintered utilizing a mold. This process may result in a self-supporting fuel segment. The shape of the fuel segment may include, but is not limited to, a rod, a rodlet, a plate, a sheet, an annuli, a sphere, or any other three-dimensional shape. In another embodiment, the consolidated volume 202 of nuclear fuel 200 may be formed by consolidating the nuclear fuel elements 204 into a container, such as a tube. For instance, a powder of spherical particle shaped nuclear fuel elements 204 may be consolidated into a tubular container. Referring now to FIG. 2F, the nuclear fuel elements 204 of the nuclear fuel 200 may include two or more grains. For example, the individual nuclear fuel elements 204 (e.g., particles) of the present invention may include a plurality of grains. The nuclear fuel elements 204 may be fabricated in manner to ensure their constituent grains have sizes small enough to ensure fission product diffusion 108 from the grain-interiors to the grain-boundaries of the nuclear fuel elements 204. In another embodiment, the nuclear fuel elements 204 of the nuclear fuel 200 may include one or more transportation pathways configured to transport one or more fission products 108 from the grain-boundaries of the nuclear fuel element-interior 210 to the surface 212 of the nuclear fuel element 204. The grain structure and transportation pathway requirements of the nuclear fuel elements 204 are consistent with the description provided previously herein. Referring now to FIG. 2G, one or more nuclear fuel elements 204 of the nuclear fuel 200 may include an interfacial layer 218. It is contemplated herein that the previously described aspects of interfacial layer formation within the nuclear fuel 100 should be interpreted to extend to the instant context. Referring now to FIGS. 2H through 2J, the nuclear fuel elements 204 may be consolidated into a solid volume 202 utilizing a mechanical consolidation method. For example, as shown in FIG. 2H, a plurality of planar nuclear fuel elements 204 may be stacked to form a consolidated stack of nuclear fuel 200. In this example, it is further contemplated that an interfacial region 218 may be optionally grown or deposited on the surface of the planar nuclear fuel elements 204 in order to provide a spacer layer between subsequent nuclear fuel elements 204. Moreover, the spacer layer may act to define the boundary network 114 as the porosity of the spacer layer may be controlled, allowing for sufficient transport of fission gases 118 from the surfaces 212 of nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. It should be further noted that in this context the pathways of the boundary network 214 need not interconnect since any single pathway has an open pathway to the geometric surface 201 of the nuclear fuel 200. In another example, shown in FIG. 2I, a planar nuclear fuel element 204 may be “rolled” into a consolidated cylindrically shaped nuclear fuel 200. For example, a sufficiently ductile metallic nuclear fuel element 204, such as thorium, may be used to form the rolled fuel illustrated in FIG. 2I. Moreover, a space layer as described above may also be optionally utilized to define an open pathway to the nuclear fuel surface 200. By way of another example, shown in FIG. 2J, a plurality of wire shaped nuclear fuel elements 204 may be woven into consolidated nuclear fuel 200. For example, a wire structure formed from processing a metallic nuclear fuel material may be woven into the consolidated nuclear fuel 200 illustrated in FIG. 2J. It is contemplated herein that the diameter of the nuclear fuel wires 210 may have diameter of approximately 5 to 100 μm. This thickness, however, should not be considered a limitation but merely an illustration. In other embodiments, the consolidated volume of the nuclear fuel 200 may be processed utilizing a variety of processes (e.g., material processing techniques) previously described herein. For example, the nuclear fuel 200 may undergo one or more processing techniques, such as, but not limited to, cold-working, annealing, tempering, normalization, chemical treatment, or irradiation. The previous description of fuel piece processing provided previously herein should be apply to the instant context. Referring now to FIGS. 3 and 4, alternative embodiments of the present invention are illustrated. In one aspect, the nuclear fuel 100 and 200 of the present invention may further include a plurality of dispersant particles 318 dispersed within the volume of the nuclear fuel 100 and 200. The dispersant particles may serve as preferential fission product 108 (e.g., fission gas 108) occupation sites within the nuclear fuel 100. In one embodiment, the dispersant particles 318 may include one or more ceramic particles. For example, the dispersant particles may include one or more oxide particles, nitride particles, or carbide particles. For instance, some of the dispersant particles may include, but is not limited to, stable oxides. One type of stable oxide suitable for implementation in a nuclear fuel setting is zirconium dioxide. It is recognized, however, that that zirconium is neutronically problematic in a nuclear fuel setting due to neutron absorption. Therefore, the wt. percentage of zirconium in a metal alloy or ceramic based nuclear fuel should be approximately between 0 and 10%. This, however, should not be considered a limitation as it is anticipated in certain contexts zirconium concentration may exceed 10%. In addition to zirconium oxide based materials, it is further contemplated that a variety of other oxide based material may be suitable for implementation in the present invention, such as, but not limited to, yttrium oxide, scandium oxide, chromium oxide, and titanium oxide. In another embodiment, the dispersant particles 318 may include one or more metallic particles. For example, the dispersant particles may include one or more metal particles, metal alloy particles, or intermetallic particles. In another embodiment, the dispersant particles 318 may include particles shells. For example, the dispersant particles may consist of substantially hollow shells of oxide material. For instance, one or more metallic particles may undergo an oxidation process. This oxidation process may result in an oxide layer at the surfaces of the one or more particles. Then, the metallic interiors of the one or more particles may under an additional treatment process which acts to dissolve the metallic center of the particle, leaving the one or dispersant particles consisting of hollow oxide shells. For example, a uranium based metal may be utilized in order to fabricate uranium oxide hollow shell dispersant particles. It is further contemplated that the uranium oxide shell dispersant particles may be fabricated to have a size of approximately 1 μm. It should be recognized that the above description does not represent a limitation but should be interpreted merely as an illustration. It is contemplated herein that the concepts described above may be extend to other metals and metal alloys (e.g., plutonium, uranium-plutonium, uranium-zirconium, or thorium) and other shell materials (e.g., nitrides or carbides). In another embodiment, as illustrated in FIG. 3, the dispersant particles 318 may be distributed along the grain-boundaries 112 of nuclear fuel 100. For example, the dispersant particles 318 may be dispersed within a nuclear fuel material (e.g., prior to consolidation of the nuclear fuel material. Then, after dispersal of the dispersant particles into the nuclear fuel material the nuclear fuel material may then be consolidated into a solid volume 102 of nuclear fuel 100. The consolidated nuclear fuel 100 may include a plurality of grains having a characteristic length 106 along at least one dimension smaller than or equal to a critical distance necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel 100. Further, the consolidated nuclear fuel may include a boundary network 114 configured to transport a fission product 108 from the grain-boundaries 110 of the nuclear fuel 100 to the geometric surface of the nuclear fuel 100. In one instance, the dispersant particles 138 may be dispersed into a molten metal or metal alloy nuclear fuel prior to casting. In another instance, the dispersant particles may be interspersed with a metal oxide nuclear fuel powder prior to compaction and sintering. In either case, the dispersant particles 138 may localize along the grain-boundaries of the solidified and crystallized nuclear fuel material 100. In another embodiment, as illustrated in FIG. 4, the dispersant particles 318 may be distributed along the surfaces of the nuclear fuel elements 204 in nuclear fuel 200. For example, the dispersant particles 318 may be dispersed within a volume of unconsolidated nuclear fuel elements 204 (e.g., powder of spherically shaped uranium dioxide particles) having a characteristic length 206 suitable for maintaining adequate fission product 108 diffusion from the nuclear fuel element-interior 210 to the nuclear fuel element surface 212. Then, after dispersal of the dispersant particles 318 into the nuclear fuel material the nuclear fuel material may then be consolidated into a solid volume 102 of nuclear fuel 100. The consolidated nuclear fuel 200 may include a boundary network 114 configured to transport a fission product 108 from the nuclear fuel element surfaces 212 to the geometric surface of the nuclear fuel 200. For instance, the dispersant particles 318 may be dispersed into a metal oxide nuclear fuel powder prior to compaction and sintering. The plurality of nuclear fuel elements 204 (along with the dispersed dispersant particles 318) may then be consolidated utilizing a consolidation process, such as, but not limited to, compaction and sintering. It should be recognized that the dispersant particles located at the grain-boundaries 112 of nuclear fuel 100 and the nuclear fuel element surfaces 212 of nuclear fuel 200 may serve as preferential occupation sites for fission gas 118 that diffuses from within the grains 104 or nuclear fuel elements 204. In this manner, the dispersant particles 318 may act to facilitate the production of a boundary network 114 or 214 in nuclear fuel 100 and nuclear fuel 200 respectively. In the case of nuclear fuel 100, the preferential nucleation of fission gas 118 at a dispersant particle 318 location may act to facilitate the “open” bubble formation previously described herein. In the case of nuclear fuel 200, the preferential nucleation of fission gas 118 at a dispersant particle 318 may act to produce connected void regions in the regions between nuclear fuel elements 204 and may aid in porosity control. In a further embodiment, the dispersant particles 318 may be distributed uniformly throughout the volume of nuclear fuel 100 or 200 in order to produce a low density geometrical arrangement. For example, in the case of a cylindrical fuel pellet, the dispersant particles 318 may be distributed throughout the nuclear fuels 100 or 200 in manner which produces low density cylindrical concentric shells. In another example, in the context of a spherical fuel segment, the dispersant particles 318 may be distributed throughout the nuclear fuels 100 or 200 in a manner which creates low density spherical concentric shells. Moreover, it is also anticipated that the density of dispersant particles within a given fuel segment may vary spatially within the fuel segment. For instance, in the case of a cylindrical fuel pellet, the maximum density may exist at the center of the fuel pellet, with the dispersant particle density decreasing as a function of the distance from the center of the fuel pellet. [see 4:00 of #53] In another embodiment, it is further contemplated that a dispersant may be introduced into the nuclear fuel 100 in order to inhibit the recrystallization of the grain structure of a cold-worked nuclear fuel material. As a result, the dispersion of particles into the volume of the nuclear fuel material may aid in achieving an average grain size 106 in the nuclear fuel 100 below a critical size required for adequate fission product diffusion. For example, a selected particle type may be introduced into the nuclear material prior to consolidation into a solid metallic nuclear fuel piece. For example, particles may be introduced at volume fractions of between 0 and 40%. It has been observed that, in a general sense, an increase in volume fraction of dispersant particles may lead to a decrease in the grain size of the nuclear fuel 100 upon recrystallization. Further, the grain size 106 upon recrystallization may also be a function of the size of the dispersant particle introduced into the nuclear fuel material. Particle sizes introduced into the nuclear fuel material may range between 0.005 and 50 μm. In a general sense, as the particle size decreases the size of grains upon recrystallization also decrease. This concept is often referred to as “Zener pinning.” The ultimate choice of dispersant particle may depend, among other things on the desired grain size 106 or nuclear fuel element size 206, chemical compatibility of the dispersant particles with the primary materials of the nuclear fuel, the potential for migration within the nuclear fuel upon exposure to a high temperature environment, and neutron cross-section of the dispersant particles. It should be recognized that the precise sizes of the grains 104 of nuclear fuel 100 or the nuclear fuel elements 204 of nuclear fuel 200 may be determined on a case by case basis. The required sizes of grains 104 or nuclear fuel elements 204 may depend on a variety of factors, including, but not limited to, nuclear reactor type, density requirements (i.e., power density requirements may demand a minimum density), chemical composition of nuclear fuel, temperature of implementation, required lifetime of nuclear fuel and the like. Therefore, these factors should be considered when engineering the specific embodiment of the nuclear fuel of the present invention. Following are a series of flowcharts depicting methods of fabrication of the nuclear fuel. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an example implementation and thereafter the following flowcharts present alternate implementations and/or expansions of the initial flowchart(s) as either sub-component operations or additional component operations building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an example implementation and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular and/or object-oriented program design paradigms. FIG. 5 illustrates an operational flow 500 representing example operations related to a method for fabricating a nuclear fuel. In FIG. 5 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1A through 4, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 4. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 500 moves to a providing operation 510. Providing operation 510 depicts providing a nuclear fuel material, the nuclear fuel material consolidated into a solid volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, nuclear fuel material may be consolidated into a volume 102 of nuclear fuel material having a plurality of grains 104. For instance, a volume 102 of metallic nuclear fuel material 124 may be cast from a molten phase into a solid nuclear fuel piece. In another instance, a ceramic nuclear fuel material 128 may be formed during a compaction and sintering process. The consolidated volume 102 of nuclear fuel material may then be provided for further processing. Then, processing operation 520 depicts performing one or more processes on the consolidated volume 102 of nuclear fuel material in order to obtain a characteristic length 106 along at least one dimension of some of the grains 104 smaller than or equal to a selected distance and a boundary network 114 configured to transport a fission product 108 from at least one grain boundary 112 of some of the grains 104 to the surface 101 of the volume 102 of the nuclear fuel material, wherein the selected distance is suitable for maintaining adequate diffusion of a fission product 108 from a grain interior 110 to at least one grain boundary 112 in some of the grains. For example, as shown in FIGS. 1A through 4, a first process step may be performed on the volume 102 of nuclear fuel material (e.g., fuel rod, fuel pellet, or fuel pebble) in order to reduce the grain sizes 106 of the grains 104 within the volume 102 of nuclear fuel material to a size below a critical size required for adequate diffusion of a fission product 108 from the interior 110 of the grains 104 to the grain-boundaries 112. Additionally, in either the first process step or a second process step a boundary network 114 suitable for transporting a fission product 108 from the grain-boundaries 112 to the geometric surface 101 of the nuclear fuel 100. FIG. 6 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 6 illustrates example embodiments where the processing operation 520 may include at least one additional operation. Additional operations may include an operation 602, an operation 604, and/or an operation 606. The operation 602 illustrates performing one or more material process techniques on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be employed to reduce the grain sizes 106 of the grains 104 within the nuclear fuel 100 below a size required for adequate diffusion of a fission product 108. In another example, one or more material processing steps may be employed to form or facilitate the formation of the boundary network 114 within the nuclear reactor fuel 100. Moreover, as the grain sizes 106 decrease within the nuclear fuel 100 the number of potential transportation pathways 116 of the boundary network 114 increases, increasing the interconnection frequency within the boundary network 114 and increasing the number of pathways 116 that intersect with the geometric surface 101 of the nuclear fuel 100. Further, grain size 106 reduction and boundary network 114 formation may be carried out utilizing a single process step or multiple process steps. Further, the operation 604 illustrates cold-working the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be cold-worked in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. The cold-working process may include, but is not limited to, cold-rolling, extruding a cast nuclear fuel material at low temperature, bending, compression, or drawing. Further, the operation 606 illustrates annealing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be annealed in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, after being cold-worked, the nuclear reactor fuel 100 may be annealed to a temperature below the recrystallization temperature in order to achieve the desired grain size 106 within the nuclear fuel 100. In another instance, during a casting process, the nuclear reactor fuel 100 may be annealed in order to facilitate the migration of precipitating agents, such as carbon or nitrogen, out of the nuclear fuel material to the grain-boundaries 112 of the nuclear fuel 100. FIG. 7 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 7 illustrates example embodiments where the processing operation 520 may include at least one additional operation. Additional operations may include an operation 702, and/or an operation 704. Further, the operation 702 illustrates normalizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a normalizing process in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, after undergoing a cold-working process, the nuclear reactor fuel 100 may be annealed to a temperature above its upper critical temperature. The nuclear fuel 100 may be held at the elevated temperature for a selected amount of time and then cooled to ambient temperatures in air. Further, the operation 704 illustrates tempering the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a tempering process in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the composition of the nuclear fuel material of the consolidated volume 102 of the nuclear fuel 100 may be suitable for precipitation of a precipitant (e.g., carbon) upon annealing. For example, a tempering process may be utilized to precipitate out a precipitating agent, such as, but not limited to, carbon. The precipitation of this agent into the grain structure of the nuclear fuel 100 may then lead to a reduction in the grain sizes 106 of the grains 104 and/or development of the boundary network 114 of the nuclear fuel 100. FIG. 8 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 8 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 802, and/or an operation 804. The operation 802 illustrates chemically treating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a chemical treatment process in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, a volume of 102 uranium dioxide may undergo an annealing process in the presence of an oxygen reducing gas (e.g., hydrogen-argon mixture or hydrogen-nitrogen mixture) in order to convert a portion of the stoichiometric UO2 phase to a non-stoichiometric oxygen reduced phase, such as UO1.8. The sub-stoichiometric phase has a reduced grain size with respect to the stoichiometric phase. The operation 804 illustrates performing one or more porosity control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a porosity control process. For instance, porosity of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., an annealing process or melting process) or a chemical treatment process. FIG. 9 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 9 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 902, and/or an operation 904. The operation 902 illustrates performing one or more grain texture control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a grain texture control process. For instance, grain textures of the grains 104 of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., annealing) or a chemical treatment process (e.g., doping). The operation 904 illustrates performing one or more mechanical treatment processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a mechanical treatment process (e.g., compression, drawing, and the like) in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. FIG. 10 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 10 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1002, 1004 and/or an operation 1006. The operation 1002 illustrates irradiating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be irradiated in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the consolidated volume 102 of nuclear fuel material may be implemented in a nuclear reactor setting. Prior to implementation in the nuclear reactor setting, the grain sizes 106 of the nuclear fuel 100 may be engineered to have a size below the critical size necessary for adequate diffusion of a produced fission gases (e.g., xenon or krypton) from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel 100. As a result, when implemented in a nuclear reactor setting the fission gases produced during the nuclear fuel 100 fission processes may efficiently nucleate at the grain-boundaries 112 of the nuclear fuel 100. This may facilitate the production of a boundary network 114 suitable for transportation of the fission gases to the geometric surface 101 of the nuclear fuel 100. The operation 1004 illustrates performing a fission process utilizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be utilized in nuclear reactor in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the elevated radiation environment and/or the high temperatures within the nuclear fuel 100 may lead to the efficient nucleation at the grain-boundaries 112 of the nuclear fuel 100. This may facilitate the production of a boundary network 114 suitable for transportation of the fission gases to the geometric surface 101 of the nuclear fuel 100. The operation 1006 illustrates introducing the consolidated volume of nuclear fuel material into an elevated temperature environment. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be exposed to a high temperature environment in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the consolidated volume 102 of nuclear fuel material may be implemented in a nuclear reactor setting. The nuclear fuel grain structure may be configured (e.g., cold-worked) to take advantage of the high temperature environment which occurs when the nuclear reactor fuel 100 undergoes fission. The thermal energy produced by the fission of a portion of the nuclear fuel 100 may act to reduce or further reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the thermal energy produced during a fission process of the nuclear fuel 100 may act to facilitate migration of precipitant agents, such as carbon or nitrogen, within the nuclear fuel material. Upon thermal activation, the precipitants may migrate to the grain-boundaries 112 of the nuclear fuel 100, aiding in the developing the boundary network 114. FIG. 11 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 11 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1102, and/or an operation 1104. The operation 1102 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along a selected dimension of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1G, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along a selected dimension of some grains 104. For instance, in grains having an elongated structure, the grains 104 may have a “thin” dimension smaller than or equal to a selected distance. The operation 1104 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along a selected direction of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1H, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along a selected direction of some of the grains smaller than or equal to a selected distance. For instance, in grains having an elongated structure, the grains 104 may have a characteristic length 106 along a selected direction 134 with the nuclear fuel 100. For example, the grains may have a selected characteristic length 106 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). FIG. 12 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 12 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1202, and/or an operation 1204. The operation 1202 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain an average characteristic length along a selected dimension of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1G, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have an average characteristic length 106 along a selected dimension of some grains 104. The operation 1204 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain an average characteristic length along a selected direction of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1H, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have an average characteristic length 106 along a selected direction of some of the grains smaller than or equal to a selected distance. For instance, in grains having an elongated structure, the grains 104 may have an average characteristic length 106 along a selected direction 134 with the nuclear fuel 100. For example, the grains may have an average selected characteristic length 106 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). FIG. 13 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 13 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1302, and/or an operation 1304. The operation 1302 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a selected statistical distribution of characteristic lengths. For example, as shown in FIGS. 1A through 4, the grains 104 of the nuclear fuel 100 may have a selected statistical distribution of characteristic lengths. For example, the grains 104 of the nuclear fuel 100 may have a grain size distribution having a selected percentage of the grains 104 having a grain size 106 below a selected distance. For instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 65% of the grains have a grain size 106 equal to or less than 4 μm, with an average grain size of 2.5 μm. In another example, the grains 104 of the nuclear fuel 100 may have a selected spatial distribution of characteristic lengths. The operation 1304 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a selected set of statistical distributions of characteristic lengths. In another embodiment, the grains 104 of the nuclear fuel 100 may have multiple statistical distributions of characteristic lengths. For instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 25% of the grains have a grain size 106 equal to or less than 10 μm, 25% of the grains have a grain size 106 equal to or less than 5 μm, and 10% of the grains are below 1 μm. FIG. 14 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 14 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1402, and/or an operation 1404. The operation 1402 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of an operation condition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order to engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of an operation condition of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon an operational condition of the nuclear fuel 100. Further, the operation 1404 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of an operational temperature of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order to engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of an operation temperature of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the operation temperature of the nuclear fuel 100. FIG. 15 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 15 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1502. Further, the operation 1502 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of a temperature induced pressure of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order to engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of a temperature induced pressure of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the temperature induced pressure within the nuclear reactor fuel 100. FIG. 16 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 16 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1602, and/or an operation 1604. The operation 1602 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of a chemical composition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order to engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of the chemical composition of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the chemical composition (e.g., type of fissile material(s), types of alloying agents, relative concentration of fissile materials, or the like) of the nuclear reactor fuel 100. The operation 1604 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of a fission product generation rate of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order to engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of the fission product 108 generation rate within the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the fission product 108 generation rate of the nuclear reactor fuel 100. Further, the fission product 108 generation rate (e.g., fission gas 118 generation rate) is proportional to the fission rate within the nuclear fuel 100, which in turn may depend upon the power density of the nuclear fuel 100, which in turn may depend upon the chemical composition of the nuclear fuel 100. FIG. 17 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 17 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1702, and/or an operation 1704. The operation 1702 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the at least one dimension is selected to maximize heat transfer from a grain interior to a grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order to engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, wherein the dimension of the grains is selected to maximize heat transfer from the grain interiors 110 to the grain-boundaries 112 of the nuclear fuel 100. For instance, a dimension of the grains 104 to be minimize may be selected in order maximize (or at least improve) heat transfer from the grain-interiors 110 to the grain-boundaries 112. Further, the operation 1704 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the grains. For instance, in order to maximize diffusion of a fission gas 118 from the grain-interiors 110 to the grain-boundaries 112 a “thin” dimension of the grains 104 may be arranged so as to align substantially perpendicular to the direction of a thermal gradient within the nuclear reactor fuel 100. FIG. 18 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 18 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1802. The operation 1802 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the selected distance is suitable for maintaining a diffusion level necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a selected level. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration within the volume 102 of the nuclear fuel 100 at or below a selected level. For instance, in a general sense, the rate of diffusion from the grain-interiors 110 to the grain-boundaries 112 in the grains 104 may be inversely related to the average grain size 106 of the grains 104 of the nuclear fuel 100. In this sense, as the grain sizes 106 of the grains 104 decrease, the fission gas 118 diffusion rate from the grain-interiors 110 to the grain-boundaries 112 increases. Therefore, the concentration of a fission gas 118 within the grains 104 may be adjusted to fall within acceptable concentration levels by engineering the grain sizes 106 of the grains 104 of the nuclear fuel 100. FIG. 19 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 19 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 1902. The operation 1902 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the selected distance is suitable for maintaining a diffusion level necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a concentration required for nucleation of the fission product. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration below a concentration level required for nucleation of the fission product 108 within a grain-interior 110. For instance, the concentration of a fission gas 118 within the grains 104 may be adjusted to fall below the concentration level required for fission gas nucleation with the grain-interiors 110 by engineering the grain sizes 106 of the grains 104 of the nuclear fuel 100. FIG. 20 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 20 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2002. The operation 2002 illustrates performing a single fabrication process on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a single process step in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and develop the boundary network 114 of the nuclear fuel 100. It should be recognized that the reduction of grain sizes 106 within the nuclear fuel 100 and the development of the boundary network 114 are intimately related as the boundary network may be geometrically defined by the region(s) between two or more grains 104 of the nuclear fuel 100. For this reason, a process that alters the grain structure of the nuclear fuel 100 by reducing the grain sizes 106 of the nuclear fuel 100 will impact the state of the boundary network 114. For example, the reduction of grain sizes 106 leads to an increase in grain-boundaries 110, which in turn leads to an increase in the potential transportation pathways 116 of the boundary network 114. Moreover, a process, such as an oxygen reduction process, described previously, may act to reduce the volume of one or more grains 104 of the nuclear fuel 100. This reduction may lead to an increase in the grain-boundary area within the nuclear fuel 100, leading to a more robust boundary network 114. FIG. 21 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 21 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2102, and/or an operation 2104. The operation 2102 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network having at least one transportation pathway configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a one or more processes in order to develop a boundary network 114 within the nuclear fuel 100 having one or more transportation pathways 116. Further, the operation 2104 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network having at least one transportation pathway configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the transportation pathway is defined by a region between two or more adjacent grains. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a one or more processes in order to develop a boundary network 114 within the nuclear fuel 100 having one or more transportation pathways 116 defined by a region between two or more grain boundaries 112. For instance, as shown in FIG. 1I, during a fission process in the nuclear fuel 100, fission gas 118 may diffuse from a grain-interior 110 to a grain-boundary. At high enough diffusion levels fission gas 118 bubbles may begin to nucleate at the grain-boundary 112. As more and more fission gas bubbles form at grain-boundary 112, an “open” bubble formation may be formed, resulting in an open transportation pathway 116 suitable for transporting fission gas 118 from a grain-boundary 110 to the geometric surface of the fuel 100. FIG. 22 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 22 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2202. Further, the operation 2202 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network having at least one transportation pathway configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the transportation pathway intersects with the at least one grain boundary. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 within the nuclear fuel 100 having one or more transportation pathways 116, wherein one or more transportation pathways 116 intersect with one or more grain boundaries 112. FIG. 23 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 23 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2302, and/or an operation 2304. The operation 2302 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network having a plurality of interconnected pathways configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 within the nuclear fuel 100 having a plurality of interconnected pathways. For instance, as discussed above, as the density of transportation pathways 116 increases the likelihood of interconnection between transportation pathways 116 may increase. As such, any process (e.g., cold-working and annealing, oxygen reducing treatment, or the like) suitable for reducing grain size 106 within the nuclear fuel 100 may be utilized to form or further develop a plurality of interconnected pathways. Further, the operation 2304 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network having a plurality of interconnected pathways configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by a region between two or more adjacent grains. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 within the nuclear fuel 100 having a plurality of interconnected pathways defined by two or more regions between two or more grains 104. For instance, as discussed above, as the density of transportation pathways 116 increases the likelihood of interconnection between transportation pathways 116 may increase. As such, any process suitable for reducing grain size 106 within the nuclear fuel 100 may be utilized to form or further develop a plurality of interconnected pathways defined by the regions between two or more grains 104. FIG. 24 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 24 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2402. Further, the operation 2402 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network having a plurality of interconnected pathways configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by one or more void regions. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 within the nuclear fuel 100 defined by one or more void regions. For instance, as discussed above, void regions may be formed by utilizing a nuclear fuel 100 doped with a dispersant (e.g., zirconium oxide particles) in a nuclear reactor setting as the dispersant particles form preferential fission gas 118 occupation sites, which create voids within the grain structure of the nuclear fuel 100. As the density of these voids grows with increasing dispersant doping levels and fission gas generation rate, the void regions may become interconnected, forming a boundary network 114. FIG. 25 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 25 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2502. The operation 2502 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a gaseous fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the selected distance suitable for maintaining adequate diffusion of a gaseous fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 suitable for transporting a fission gas 118 from the grain-boundaries 112 of the grains 104 to the geometric surface 101 of the nuclear fuel. For instance, the consolidated volume 102 of nuclear fuel 100 material may undergo one or more processes in order to develop a boundary network 114 suitable for transporting xenon or krypton gas from the grain-boundaries 112 of the grains 104 to the geometric surface 101 of the nuclear fuel 100. FIG. 26 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 26 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2602. The operation 2602 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a liquid fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the selected distance is suitable for maintaining adequate diffusion of a liquid fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 suitable for transporting a liquid fission product 119 from the grain-boundaries 112 of the grains 104 to the geometric surface 101 of the nuclear fuel. For instance, the consolidated volume 102 of nuclear fuel 100 material may undergo one or more processes in order to develop a boundary network 114 suitable for transporting liquid sodium or liquid cesium from the grain-boundaries 112 of the grains 104 to the geometric surface 101 of the nuclear fuel 100. FIG. 27 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 27 illustrates example embodiments where the operation 520 may include at least one additional operation. Additional operations may include an operation 2702. The operation 2702 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance and a boundary network configured to transport a solid fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the selected distance is suitable for maintaining adequate diffusion of a solid fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo one or more processes in order to develop a boundary network 114 suitable for transporting a solid fission product 120 from the grain-boundaries 112 of the grains 104 to the geometric surface 101 of the nuclear fuel. For instance, the consolidated volume 102 of nuclear fuel 100 material may undergo one or more processes in order to develop a boundary network 114 suitable for transporting a solid fission product 120, such as tellurium or cesium, from the grain-boundaries 112 of the grains 104 to the geometric surface 101 of a metal oxide, such as uranium dioxide, based nuclear fuel 100. FIG. 28 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 28 illustrates example embodiments where the operation 510 may include at least one additional operation. Additional operations may include an operation 2802, and/or an operation 2804. The operation 2802 illustrates providing a nuclear fuel material, the nuclear fuel material consolidated into a solid volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains, wherein some of the plurality of grains have an interfacial layer including a material different from the material of a grain interior. For example, as shown in FIG. 1J, the grains 104 of the solid volume of provided nuclear fuel material may include an interfacial layer of a material different than the grain-interiors 110. For instance, the grains 104 may include an oxide-based or carbide-based interfacial layer. The operation 2804 illustrates providing a ceramic nuclear fuel material, the ceramic nuclear fuel material consolidated into a solid volume of ceramic nuclear fuel material having a surface, the consolidated ceramic nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel material may include a ceramic based material nuclear fuel material. For instance, nuclear fuel material may include, but is not limited to a metal oxide (e.g., uranium dioxide, plutonium dioxide, or thorium dioxide) nuclear fuel material, a mixed oxide nuclear fuel material (e.g., blend of plutonium dioxide and depleted uranium dioxide), a metal nitride (e.g., uranium nitride) based nuclear fuel material, or a metal carbide (e.g., uranium carbide) based nuclear fuel material. FIG. 29 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 29 illustrates example embodiments where the operation 510 may include at least one additional operation. Additional operations may include an operation 2902, and/or an operation 2904. The operation 2902 illustrates providing a metal nuclear fuel material, a metal alloy nuclear fuel material, or an intermetallic nuclear fuel material consolidated into a solid volume of metal nuclear fuel material, a metal alloy nuclear fuel material, or an intermetallic nuclear fuel material having a surface, the consolidated metal nuclear fuel material, a metal alloy nuclear fuel material, or an intermetallic nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel material may include a metal based nuclear fuel material. For instance, nuclear fuel material may include, but is not limited to a metal (e.g., uranium, plutonium, or thorium) nuclear fuel material, a metal alloy fuel material (e.g., uranium zirconium, uranium-plutonium-zirconium, or uranium zirconium hydride), or an intermetallic (e.g., UFe2 or UNi2) based nuclear fuel material. The operation 2904 illustrates providing a nuclear fuel material including at least one of a uranium isotope, a plutonium isotope, or a thorium isotope, the nuclear fuel material consolidated into a solid volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel may include a fissile nuclear material including, but not limited to, uranium-235 or plutonium-239. By way of another example, the provided nuclear fuel may include a non-fissile nuclear material including, but not limited to, thorium-232. While thorium-232 is not by itself fissile, it may be utilized to breed uranium-233, which is fissile in nature. FIG. 30 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 30 illustrates example embodiments where the operation 510 may include at least one additional operation. Additional operations may include an operation 3002, and/or an operation 3004. The operation 3002 illustrates providing a nuclear fuel material, the nuclear fuel material consolidated into a solid volume having a density equal to or below a theoretical density, the nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, the consolidation process (e.g., casting, compacting, sintering, or the like) used to create the volume 102 of consolidated nuclear fuel material may fabricate a nuclear fuel piece having a selected density, wherein the selected density is less than the theoretical density. For instance, the nuclear fuel material may be consolidated to a density of 70% of the theoretical density. The operation 3004 illustrates providing a nuclear fuel material, the nuclear fuel material consolidated into a solid volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains, the volume of nuclear fuel contained in a geometry maintaining container. For example, as shown in FIGS. 1A through 4, a casting process may consolidate a metallic nuclear fuel material inside a fuel rod, where the molten metallic nuclear fuel material may then solidify. FIG. 31 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 31 illustrates example embodiments where the operation 510 may include at least one additional operation. Additional operations may include an operation 3102, and/or an operation 3104. The operation 3102 illustrates providing a nuclear fuel material, the nuclear fuel material consolidated into a solid self-supporting volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be formed into a self-supporting geometry. The operation 3104 illustrates compacting a nuclear fuel material into a consolidated solid self-supporting volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be placed in a mold and compacted to form a self-supporting fuel pellet. FIG. 32 illustrates alternative embodiments of the example operational flow 500 of FIG. 5. FIG. 32 illustrates example embodiments where the operation 510 may include at least one additional operation. Additional operations may include an operation 3202, 3204, and/or operation 3206. The operation 3202 illustrates sintering a nuclear fuel material into a consolidated solid self-supporting volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be placed in a mold and compacted and sintered to form a self-supporting fuel pellet. The operation 3204 illustrates casting a nuclear fuel material into a consolidated solid self-supporting volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, metallic nuclear fuel material, such as a metal alloy (e.g., Uranium-Plutonium), may be cast from a molten phase into a mold. Upon casting into a mold, the molten nuclear fuel material may undergo a cooling process until solidification. The operation 3206 illustrates extruding a nuclear fuel material into a consolidated solid self-supporting volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, metallic nuclear fuel material, such as a metal alloy (e.g., Uranium-Plutonium), may undergo an extruding process at room temperature or nearly room temperature to form a solid nuclear fuel piece. As has been discussed above, low-temperature extrusion has the added benefit of producing a grain structure having a reduced average grain size FIG. 33 illustrates an operational flow 3300 representing example operations related to a method for fabricating a nuclear fuel. In FIG. 33 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1A through 4, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 4. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 3300 moves to a providing operation 3310. Providing operation 3310 depicts providing a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having an characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, some of the nuclear fuel elements including a metal nuclear fuel material. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204 may be fabricated via a ball milling process such that their average size is smaller than a critical distance suitable for maintaining adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. For instance, a plurality of spherical nuclear fuel particles may be fabricated to have an average radius of 100 nm. Then, consolidating operation 3320 depicts consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, the provided plurality of nuclear fuel elements 204 (e.g., particles) may be consolidated into a solid volume 202 utilizing a compaction process. FIG. 34 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 34 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 3402, an operation 3404, and/or an operation 3406. The operation 3402 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 3404 illustrates performing one or more material processing techniques on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 3406 illustrates performing one or more cold-working processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a cold-working process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The cold-working process may include, but is not limited to, cold-rolling, drawing, bending, or compression. FIG. 35 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 35 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 3502. Further, the operation 3502 illustrates performing one or more annealing processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, an annealing process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the nuclear fuel elements 204 may be annealed in the presence of a processing gas, such as an oxygen reducing gas. FIG. 36 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 36 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 3602. Further, the operation 3602 illustrates performing one or more normalizing processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a normalizing process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204, as described previously herein. FIG. 37 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 37 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 3702. Further, the operation 3702 illustrates performing one or more tempering processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a tempering process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204, as described previously herein. FIG. 38 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 38 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 3802. Further, the operation 3802 illustrates performing one or more chemical treatment processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a chemical treatment process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. For instance, an oxygen reducing treatment may be performed on the provided nuclear fuel elements 204, as described previously herein. FIG. 39 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 39 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 3902. Further, the operation 3902 illustrates performing one or more mechanical treatment processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a mechanical process (e.g., ball milling) may be performed on the provided nuclear fuel elements 204 in order to reduce one or more dimensions of the nuclear fuel elements 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 40 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 40 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4002. Further, the operation 4002 illustrates performing one or more porosity control processes on a plurality of nuclear fuel elements in order to achieve a selected porosity within some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a porosity control process may be performed on the provided nuclear fuel elements 204 in order to achieve a selected porosity in the nuclear fuel elements 204 the nuclear fuel elements 206. For instance, porosity of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., an annealing process or melting process) or a chemical treatment process. FIG. 41 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 41 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4102. Further, the operation 4102 illustrates performing one or more grain texture control processes on a plurality of nuclear fuel particles in order to achieve a selected grain texture within some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a grain texture control process may be performed on the provided nuclear fuel elements 204 in order to achieve a selected grain texture in two or more grains of the nuclear fuel elements 204. For instance, grain textures of the grains of the nuclear fuel elements 204 may be controlled via a heat treatment process (e.g., annealing) or a chemical treatment process (e.g., doping). FIG. 42 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 42 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4202, an operation 4204, and/or an operation 4206. The operation 4202 illustrates irradiating a plurality of nuclear fuel particles. For example, as shown in FIGS. 1A through 4, an irradiating process (e.g., exposure to neutron flux) may be performed on the provided nuclear fuel elements 204 in order to reduce one or more dimensions of the nuclear fuel elements 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 4204 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along a selected dimension of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIG. 2D, one or more processes may be utilized in order engineer the nuclear fuel elements 204 to have a characteristic length 206 along a selected dimension of some of the nuclear fuel elements 204. For instance, in nuclear fuel elements 204 having an elongated structure, the nuclear fuel elements 204 may have a “thin” dimension that is smaller than or equal to a selected distance. The operation 4206 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along a selected direction of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIG. 2E, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along a selected direction smaller than or equal to a selected distance. For instance, in nuclear fuel elements having an elongated structure, the nuclear fuel elements 204 may have a characteristic length 206 along a selected direction 134 within the nuclear fuel 200. For example, the nuclear fuel elements may have a selected characteristic length 206 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). FIG. 43 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 43 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4302, an operation 4304, and/or an operation 4306. The operation 4302 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain an average characteristic length along a selected dimension of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have an average characteristic length 206 along a selected dimension of some nuclear fuel elements 204. The operation 4304 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain an average characteristic length along a selected direction of some of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have an average characteristic length 206 along a selected direction of some of the nuclear fuel elements 204 smaller than or equal to a selected distance. For instance, in nuclear fuel elements 204 having an elongated structure, the nuclear fuel elements 204 may have an average characteristic length 206 along a selected direction 134 with the nuclear fuel 200. For example, the nuclear fuel elements may have an average selected characteristic length 206 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). The operation 4306 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a selected statistical distribution of characteristic lengths in the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 of the nuclear fuel 200 may have a selected statistical distribution of characteristic lengths 206. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may have a element size distribution with a selected percentage of the nuclear fuel elements 204 having a size 206 below a selected distance. For instance, the nuclear fuel 200 of the present invention may have a nuclear fuel element (e.g., particle) size 206 distribution such that 65% of the nuclear fuel elements 204 have a size 206 equal to or less than 4 μm, with an average size of 2.5 μm. In another example, the nuclear fuel elements 204 of the nuclear fuel 200 may have a selected spatial distribution of characteristic lengths, within the consolidated volume of nuclear fuel 200. FIG. 44 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 44 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4402, an operation 4404, and/or an operation 4406. The operation 4402 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a selected set of statistical distributions of characteristic lengths. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may have multiple statistical distributions of characteristic lengths 206. For instance, the nuclear fuel 200 of the present invention may have a nuclear fuel element size 206 distribution such that 25% of the nuclear fuel elements 204 have a size equal to or less than 10 μm, 25% of the nuclear fuel elements have a nuclear fuel element size 106 equal to or less than 5 μm, and 10% of the nuclear fuel elements are below 1 μm. The operation 4404 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a chemical composition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of the chemical composition of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements of the nuclear fuel 200 may depend upon the chemical composition (e.g., type of fissile material(s), types of alloying agents, relative concentration of fissile materials, or the like) of the nuclear reactor fuel 200. The operation 4406 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a fission product generation rate of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of the fission product 108 generation rate within the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend upon the fission product 108 generation rate of the nuclear reactor fuel 200. Further, the fission product 108 generation rate (e.g., fission gas 118 generation rate) is proportional to the fission rate with the nuclear fuel 200, which in turn is proportional to the power density of the nuclear fuel 200, which in turn is dependent upon the chemical composition of the nuclear fuel 200. FIG. 45 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 45 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4502, an operation 4504, and/or an operation 4506. The operation 4502 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of an operation condition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel smaller than a selected distance, which is a function of an operation condition of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend upon an operational condition of the nuclear fuel 200. Further, the operation 4504 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of an operational temperature of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of an operation temperature of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend may depend upon the operation temperature of the nuclear fuel 200. Further, the operation 4506 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a temperature induced pressure of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel smaller than a selected distance, which is a function of a temperature induced pressure of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend may depend upon the temperature induced pressure within the nuclear reactor fuel 100. FIG. 46 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 46 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4602, and/or an operation 4604. The operation 4602 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the at least one dimension is selected to maximize heat transfer from a nuclear fuel element interior to a free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, wherein the dimension of the nuclear fuel elements is selected in order to maximize heat transfer from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200. For instance, a dimension of the nuclear fuel elements 204 to be minimized may be selected in order maximize (or at least improve) heat transfer from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 4604 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a nuclear fuel element interior in some of in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the nuclear fuel elements. For instance, in order to maximize diffusion of a fission gas 118 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 a “thin” dimension of the nuclear fuel elements 204 may be arranged so as to align substantially perpendicular to the direction of a thermal gradient within the nuclear reactor fuel 100. FIG. 47 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 47 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4702, and/or an operation 4704. The operation 4702 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level in the plurality of nuclear fuel elements necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a selected level. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration within the volume 102 of the nuclear fuel 100 at or below a selected level. For instance, the rate of diffusion from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 in the nuclear fuel elements 204 may be inversely related to the average nuclear fuel element size 206 within the nuclear fuel 200. In this sense, as the nuclear fuel element sizes 206 of the nuclear fuel elements 204 decrease, the fission gas 118 diffusion rate from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 may increase. Therefore, the concentration of a fission gas 118 within the nuclear fuel elements 204 may be adjusted to fall within acceptable concentration levels by engineering the nuclear fuel element sizes 206 of the nuclear fuel elements 204 of the nuclear fuel 200. Further, the operation 4704 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level in the plurality of nuclear fuel elements necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a concentration required for nucleation of the fission product. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 concentration below a concentration level required for nucleation of the fission product 108 within an interior 210 of a nuclear fuel element 204. For instance, the concentration of a fission gas 118 within the nuclear fuel elements 204 may be adjusted to fall below the concentration level required for fission gas nucleation within the interiors 210 of the nuclear fuel elements 204 by engineering the nuclear fuel element sizes 206 of the nuclear fuel elements 204 of the nuclear fuel 200. FIG. 48 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 48 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4802, and/or an operation 4804. The operation 4802 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a gaseous fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a gaseous fission product (e.g., krypton or xenon) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 4804 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a liquid fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a liquid fission product (e.g., a liquid metal) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 49 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 49 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 4902, and/or an operation 4904. The operation 4902 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a solid fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a solid fission product (e.g., tellurium or cesium) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 4904 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements have an interfacial layer including a material different from an interior of a nuclear fuel element. For example, as shown in FIG. 2G, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include an interfacial layer of a material different from the material within the interiors 210 of the nuclear fuel elements 204. For instance, the nuclear fuel elements 204 may include an oxide-based or carbide-based interfacial layer. FIG. 50 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 50 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 5002, and/or an operation 5004. The operation 5002 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include two or more grains. For example, as shown in FIG. 2F, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include two or more grains. The operation 5004 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include a plurality of pathways configured to transport a fission product from at least one portion of a nuclear fuel element interior to at least one free surface of the nuclear fuel element. For example, as shown in FIG. 2F, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include one or more internal pathways suitable for transporting fission gas 118 from the nuclear fuel element interior 210 to the nuclear fuel element surface 212. Moreover, as previously described herein, the internal pathways 110 may be defined by a grain-boundary 112 between adjacent grains within a common nuclear fuel element 204. FIG. 51 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 51 illustrates example embodiments where the operation 3310 may include at least one additional operation. Additional operations may include an operation 5102, and/or an operation 5104. The operation 5102 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include at least one of a pure metal, a metal alloy, or an intermetallic nuclear fuel material. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel elements 204 may include a metal based nuclear fuel material. For instance, the plurality of nuclear fuel elements 204 of the nuclear fuel 200 may include, but is not limited to a metal (e.g., uranium, plutonium, or thorium) nuclear fuel material, a metal alloy fuel material (e.g., uranium zirconium, uranium-plutonium-zirconium, or uranium zirconium hydride), or an intermetallic (e.g., UFe2 or UNi2) based nuclear fuel material. The operation 5104 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include at least one of a uranium isotope, a plutonium isotope, or a thorium isotope. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel elements 204 may include a fissile nuclear material including, but not limited to, uranium-235 or plutonium-239. By way of another example, the provided nuclear fuel elements 204 may include a non-fissile nuclear material including, but not limited to, thorium-232. While thorium-232 is not by itself fissile, it may be utilized to breed uranium-233, which is fissile in nature. FIG. 52 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 52 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5202, an operation 5204, and/or an operation 5206. The operation 5202 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 5204 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes at least one transportation pathway defined by a region between two or more adjacent nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216. Further, the operation 5206 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes at least one transportation pathway intersecting the at least one free surface. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216 intersecting a surface 212 of one or more nuclear fuel elements 204. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216 intersecting a surface 212 of one or more nuclear fuel elements 204. FIG. 53 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 53 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5302, and/or an operation 5304. Further, the operation 5302 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 plurality of interconnected transportation pathways 216 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 5304 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways are defined by a region between two or more adjacent nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having a plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements. 204 and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 54 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 54 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5402. Further, the operation 5402 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by one or more void regions. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having a plurality of interconnected transportation pathways 216 defined by one or more void regions and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 55 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 55 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5502, and/or an operation 5504. Further, the operation 5502 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of non-interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of non-interconnected transportation pathways 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 5504 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of non-interconnected pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality non-interconnected transportation pathways is defined by a region between surfaces of adjacent and substantially parallel or concentric nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of non-interconnected transportation pathways 214 defined by a region between surfaces of adjacent and substantially parallel or concentric nuclear fuel elements 204. FIG. 56 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 56 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5602, an operation 5604, and/or an operation 5606. The operation 5602 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, wherein the nuclear fuel material is consolidated to a density equal to or below a theoretical density. For example, as shown in FIGS. 1A through 4, the consolidation process (e.g., compacting, sintering, or the like) used to create the volume 202 of consolidated nuclear fuel 200 may produce a nuclear fuel piece having a selected density, wherein the selected density is less than the theoretical density. For instance, the nuclear fuel 200 may be consolidated to a density of 70% of the theoretical density. The operation 5604 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, wherein the nuclear fuel material is contained in a geometry maintaining container. For example, as shown in FIGS. 1A through 4, the plurality of nuclear fuel elements 204 may be compacted into a fuel containing vessel or container suitable for maintaining the shape of the nuclear fuel piece. The operation 5606 illustrates consolidating the plurality of nuclear fuel elements into a self-supporting volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be formed into a self-supporting geometry via a compacting and sintering. FIG. 57 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 57 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5702, and/or an operation 5704. The operation 5702 illustrates compacting the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a metal powder, may be placed in a mold and compacted to form a self-supporting fuel pellet. The operation 5704 illustrates sintering the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a metal powder, may be placed in a mold and compacted and sintered to form a self-supporting fuel pellet. FIG. 58 illustrates alternative embodiments of the example operational flow 3300 of FIG. 33. FIG. 58 illustrates example embodiments where the operation 3320 may include at least one additional operation. Additional operations may include an operation 5802, an operation 5804, an operation 5806, and/or an operation 5808. The operation 5802 illustrates mechanically arranging the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a plurality of metal (e.g., thorium) or metal alloy (e.g., uranium alloy) nuclear fuel elements, may be mechanically arranged into a volume 202 of nuclear fuel 200. Further, the operation 5804 illustrates weaving a plurality of linear nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2J, a plurality of nuclear fuel elements 204, such as a plurality of metal (e.g., thorium) or metal alloy (e.g., uranium alloy) nuclear fuel elements, may be woven into a woven structure 224 of nuclear fuel 200. Further, the operation 5806 illustrates rolling a plurality of planar nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2I, a nuclear fuel element 204, such as a metal or metal alloy planar sheet, may be rolled into a cylindrical volume 222. It is further recognized that two or more of the cylindrical rolled volumes 222 may be combined to form a nuclear fuel 200. Further, the operation 5808 illustrates stacking a plurality of planar nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2H, two or more nuclear fuel elements 204, such as a metal or metal alloy planar sheet, may be stacked together in order to form a volume of nuclear fuel 200. FIG. 59 illustrates an operational flow 5900 representing example operations related to a method for fabricating nuclear fuel. FIG. 59 illustrates an example embodiment where the example operational flow 3300 of FIG. 33 may include at least one additional operation. Additional operations may include an operation 5910, an operation 5912, an operation 5914, and/or an operation 5916. After a start operation, a providing operation 3310, and a consolidating operation 3320, the operational flow 5900 moves to a processing operation 5910. Operation 5910 illustrates performing one or more processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more process steps may be performed on the volume 202 of nuclear fuel 200 (e.g., fuel rod, fuel pellet, or fuel pebble). The operation 5912 illustrates performing one or more material processing techniques on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the volume 202 of nuclear fuel 200 in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 5914 illustrates cold-working the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be cold-worked in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. The cold-working process may include, but is not limited to, cold-rolling, extruding at low temperature, bending, compression, or drawing. Further, the operation 5916 illustrates annealing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel material may be annealed. For instance, after being cold-worked, the nuclear reactor fuel 200 may be annealed in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 60 illustrates alternative embodiments of the example operational flow 5900 of FIG. 59. FIG. 60 illustrates example embodiments where the operation 5910 may include at least one additional operation. Additional operations may include an operation 6002, and/or an operation 6004. Further, the operation 6002 illustrates melting a portion of the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, a portion of the consolidated volume 202 of nuclear fuel 200 may be melted. Further, the operation 6004 illustrates normalizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be normalized in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 61 illustrates alternative embodiments of the example operational flow 5900 of FIG. 59. FIG. 61 illustrates example embodiments where the operation 5910 may include at least one additional operation. Additional operations may include an operation 6102, and/or an operation 6104. Further, the operation 6102 illustrates tempering the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be tempered in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 6104 illustrates chemically treating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be chemically treated in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 62 illustrates alternative embodiments of the example operational flow 5900 of FIG. 59. FIG. 62 illustrates example embodiments where the operation 5910 may include at least one additional operation. Additional operations may include an operation 6202. Further, the operation 6202 illustrates performing one or more porosity control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may undergo a porosity control process (e.g., annealing or chemical treatment). FIG. 63 illustrates alternative embodiments of the example operational flow 5900 of FIG. 59. FIG. 63 illustrates example embodiments where the operation 5910 may include at least one additional operation. Additional operations may include an operation 6302, an operation 6304, and/or an operation 6306. The operation 6302 illustrates introducing the consolidated volume of nuclear fuel material into an elevated temperature environment. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be introduced into a high temperature environment, such as a operation within a nuclear reactor. The operation 6304 illustrates irradiating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be irradiated (e.g., irradiated in nuclear reactor implementation or irradiated via neutron source) in order to refine the sizes of the nuclear fuel elements 204 or the boundary network 214. The operation 6306 illustrates performing a fission process utilizing the consolidated volume of nuclear fuel material For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be utilized in a fission process (e.g., utilized in a nuclear reactor). It is recognized that the sizes of the nuclear fission elements 204 may become more refined and/or the boundary network 214 of the nuclear fuel 200 may become more developed upon implementing the nuclear fuel 200 in a nuclear reactor 200. FIG. 64 illustrates an operational flow 6400 representing example operations related to a method for fabricating a nuclear fuel. In FIG. 64 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1A through 4, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 4. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 6400 moves to a providing operation 6410. Providing operation 6410 depicts providing a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having an characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements, some of the nuclear fuel elements including a ceramic nuclear fuel material. For example, as shown in FIGS. 1A through 4, a plurality of ceramic nuclear fuel elements 204 may be fabricated via a ball milling process such that their average size is smaller than a critical distance suitable for maintaining adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. For instance, a plurality of spherical ceramic nuclear fuel particles may be fabricated to have an average radius of 100 nm. Then, consolidating operation 6420 depicts consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, the volume of nuclear fuel material including a boundary network configured to transport the fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 (e.g., uranium dioxide particles) may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. By way of further example, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 65 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 65 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 6502, an operation 6504, and/or an operation 6506. The operation 6502 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes (e.g., ball milling, nanostructuring, or chemical treatment) may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 6504 illustrates performing one or more material processing techniques on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 6506 illustrates performing one or more cold-working processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a cold-working process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The cold-working process may include, but is not limited to, cold-rolling, drawing, bending, or compression. FIG. 66 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 66 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 6602. Further, the operation 6602 illustrates performing one or more annealing processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, an annealing process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the nuclear fuel elements 204 may be annealed in the presence of a processing gas, such as an oxygen reducing gas. FIG. 67 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 67 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 6702. Further, the operation 6702 illustrates performing one or more normalizing processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a normalizing process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204, as described previously herein. FIG. 68 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 68 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 6802. Further, the operation 6802 illustrates performing one or more tempering processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a tempering process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204, as described previously herein. FIG. 69 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 69 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 6902. Further, the operation 6902 illustrates performing one or more chemical treatment processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a chemical treatment process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. For instance, an oxygen reducing treatment may be performed on the provided nuclear fuel elements 204, as described previously herein. FIG. 70 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 70 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7002. Further, the operation 7002 illustrates performing one or more mechanical treatment processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a mechanical process (e.g., ball milling) may be performed on the provided nuclear fuel elements 204 in order to reduce one or more dimensions of the nuclear fuel elements 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 71 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 71 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7102. Further, the operation 7102 illustrates performing one or more porosity control processes on a plurality of nuclear fuel elements in order to achieve a selected porosity within some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a porosity control process may be performed on the provided nuclear fuel elements 204 in order to achieve a selected porosity in the nuclear fuel elements 204 the nuclear fuel elements 206. For instance, porosity of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., an annealing process or melting process) or a chemical treatment process. FIG. 72 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 72 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7202. Further, the operation 7202 illustrates performing one or more grain texture control processes on a plurality of nuclear fuel elements in order to achieve a selected grain texture within some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a grain texture control process may be performed on the provided nuclear fuel elements 204 in order to achieve a selected grain texture in two or more grains of the nuclear fuel elements 204. For instance, grain textures of the grains of the nuclear fuel elements 204 may be controlled via a heat treatment process (e.g., annealing) or a chemical treatment process (e.g., doping). FIG. 73 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 73 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7302, an operation 7304, and/or an operation 7306. The operation 7302 illustrates irradiating a plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, an irradiating process (e.g., exposure to neutron flux) may be performed on the provided nuclear fuel elements 204 in order to reduce one or more dimensions of the nuclear fuel elements 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 7304 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along a selected dimension of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIG. 2D, one or more processes may be utilized in order engineer the nuclear fuel elements 204 to have a characteristic length 206 along a selected dimension of some of the nuclear fuel elements 204. For instance, in nuclear fuel elements 204 having an elongated structure, the nuclear fuel elements 204 may have a “thin” dimension that is smaller than or equal to a selected distance. The operation 7306 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along a selected direction of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIG. 2E, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along a selected direction smaller than or equal to a selected distance. For instance, in nuclear fuel elements having an elongated structure, the nuclear fuel elements 204 may have a characteristic length 206 along a selected direction 134 within the nuclear fuel 200. For example, the nuclear fuel elements may have a selected characteristic length 206 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). FIG. 74 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 74 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7402, an operation 7404, and/or an operation 7406. The operation 7402 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain an average characteristic length along a selected dimension of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have an average characteristic length 206 along a selected dimension of some nuclear fuel elements 204. The operation 7404 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain an average characteristic length along a selected direction of some of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have an average characteristic length 206 along a selected direction of some of the nuclear fuel elements 204 smaller than or equal to a selected distance. For instance, in nuclear fuel elements 204 having an elongated structure, the nuclear fuel elements 204 may have an average characteristic length 206 along a selected direction 134 with the nuclear fuel 200. For example, the nuclear fuel elements may have an average selected characteristic length 206 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). The operation 7406 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a selected statistical distribution of characteristic lengths in the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 of the nuclear fuel 200 may have a selected statistical distribution of characteristic lengths 206. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may have an element size distribution with a selected percentage of the nuclear fuel elements 204 having a size 206 below a selected distance. For instance, the nuclear fuel 200 of the present invention may have a nuclear fuel element (e.g., particle) size 206 distribution such that 65% of the nuclear fuel elements 204 have a size 206 equal to or less than 1 μm, with an average size of 0.750 μm. In another example, the nuclear fuel elements 204 of the nuclear fuel 200 may have a selected spatial distribution of characteristic lengths, within the consolidated volume of nuclear fuel 200. FIG. 75 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 75 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7502, an operation 7504, and/or an operation 7506. The operation 7502 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a selected set of statistical distributions of characteristic lengths. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may have multiple statistical distributions of characteristic lengths 206. For instance, the nuclear fuel 200 of the present invention may have a nuclear fuel element size 206 distribution such that 25% of the nuclear fuel elements 204 have a size equal to or less than 1 μm, 25% of the nuclear fuel elements have a nuclear fuel element size 106 equal to or less than 0.5 μm, and 10% of the nuclear fuel elements are below 0.1 μm. The operation 7504 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a chemical composition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of the chemical composition of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements of the nuclear fuel 200 may depend upon the chemical composition (e.g., type of fissile material(s), types of alloying agents, relative concentration of fissile materials, or the like) of the nuclear reactor fuel 200. The operation 7506 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a fission product generation rate of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of the fission product 108 generation rate within the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend upon the fission product 108 generation rate of the nuclear reactor fuel 200. Further, the fission product 108 generation rate (e.g., fission gas 118 generation rate) is proportional to the fission rate with the nuclear fuel 200, which in turn is proportional to the power density of the nuclear fuel 200, which in turn is dependent upon the chemical composition of the nuclear fuel 200. FIG. 76 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 76 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7602, an operation 7604, an operation 7606, an operation 7608, and/or an operation 7610. The operation 7602 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of an operation condition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel smaller than a selected distance, which is a function of an operation condition of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend upon an operational condition of the nuclear fuel 200. Further, the operation 7604 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of an operational temperature of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of an operation temperature of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend may depend upon the operation temperature of the nuclear fuel 200. Further, the operation 7606 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a temperature induced pressure of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel smaller than a selected distance, which is a function of a temperature induced pressure of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend may depend upon the temperature induced pressure within the nuclear reactor fuel 100. The operation 7608 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the at least one dimension is selected to maximize heat transfer from a nuclear fuel element interior to a free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, wherein the dimension of the nuclear fuel elements is selected in order to maximize heat transfer from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200. For instance, a dimension of the nuclear fuel elements 204 to be minimized may be selected in order maximize (or at least improve) heat transfer from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 7610 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a nuclear fuel element interior in some of in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the nuclear fuel elements. For instance, in order to maximize diffusion of a fission gas 118 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 a “thin” dimension of the nuclear fuel elements 204 may be arranged so as to align substantially perpendicular to the direction of a thermal gradient within the nuclear reactor fuel 100. FIG. 77 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 77 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7702, and/or an operation 7704. The operation 7702 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level in the plurality of nuclear fuel elements necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a selected level. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration within the volume 102 of the nuclear fuel 100 at or below a selected level. For instance, the rate of diffusion from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 in the nuclear fuel elements 204 may be inversely related to the average nuclear fuel element size 206 within the nuclear fuel 200. In this sense, as the nuclear fuel element sizes 206 of the nuclear fuel elements 204 decrease, the fission gas 118 diffusion rate from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 may increase. Therefore, the concentration of a fission gas 118 within the nuclear fuel elements 204 may be adjusted to fall within acceptable concentration levels by engineering the nuclear fuel element sizes 206 of the nuclear fuel elements 204 of the nuclear fuel 200. Further, the operation 7704 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level in the plurality of nuclear fuel elements necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a concentration required for nucleation of the fission product. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 concentration below a concentration level required for nucleation of the fission product 108 within an interior 210 of a nuclear fuel element 204. For instance, the concentration of a fission gas 118 within the nuclear fuel elements 204 may be adjusted to fall below the concentration level required for fission gas nucleation within the interiors 210 of the nuclear fuel elements 204 by engineering the nuclear fuel element sizes 206 of the nuclear fuel elements 204 of the nuclear fuel 200. FIG. 78 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 78 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7802, and/or an operation 7804. The operation 7802 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a gaseous fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a gaseous fission product (e.g., krypton or xenon) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 7804 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a liquid fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a liquid fission product (e.g., a liquid metal) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 79 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 79 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 7902, and/or an operation 7904. The operation 7902 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a solid fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a solid fission product (e.g., tellurium or cesium) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 7904 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements have an interfacial layer including a material different from an interior of a nuclear fuel element. For example, as shown in FIG. 2G, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include an interfacial layer of a material different from the material within the interiors 210 of the nuclear fuel elements 204. For instance, the nuclear fuel elements 204 may include an oxide-based, nitride-based, or carbide-based interfacial layer. FIG. 80 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 80 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 8002, and/or an operation 8004. The operation 8002 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include two or more grains. For example, as shown in FIG. 2F, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include two or more grains (i.e., the nuclear fuel elements 204 are polycrystalline). The operation 8004 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include a plurality of pathways configured to transport a fission product from at least one portion of a nuclear fuel element interior to at least one free surface of the nuclear fuel element. For example, as shown in FIG. 2F, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include one or more internal pathways suitable for transporting fission gas 118 from the nuclear fuel element interior 210 to the nuclear fuel element surface 212. Moreover, as previously described herein, the internal pathways 110 may be defined by a grain-boundary 112 between adjacent grains within a common nuclear fuel element 204. FIG. 81 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 81 illustrates example embodiments where the operation 6410 may include at least one additional operation. Additional operations may include an operation 8102, and/or an operation 8104. The operation 8102 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include at least one of an oxide, a mixed oxide, a nitride, or a carbide nuclear fuel material. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel elements 204 may include a ceramic based nuclear fuel material. For instance, the plurality of nuclear fuel elements 204 of the nuclear fuel 200 may include, but is not limited to, metal oxide material (e.g., uranium dioxide, plutonium dioxide, or thorium dioxide), a metal nitride fuel material (e.g., uranium nitride), or metal carbide fuel material (e.g., uranium carbide). The operation 8104 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include at least one of a uranium isotope, a plutonium isotope, or a thorium isotope. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel elements 204 may include a fissile nuclear material including, but not limited to, uranium-235 or plutonium-239. By way of another example, the provided nuclear fuel elements 204 may include a non-fissile nuclear material including, but not limited to, thorium-232. While thorium-232 is not by itself fissile, it may be utilized to breed uranium-233, which is fissile in nature. FIG. 82 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 82 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8202, and/or an operation 8204. The operation 8202 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes at least one transportation pathway defined by a region between two or more adjacent nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216. The operation 8204 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes at least one transportation pathway intersecting the at least one free surface. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216 intersecting a surface 212 of one or more nuclear fuel elements 204. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216 intersecting a surface 212 of one or more nuclear fuel elements 204. FIG. 83 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 83 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8302, and/or an operation 8304. The operation 8302 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 plurality of interconnected transportation pathways 216 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. By way of further example, FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 8304 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by a region between two or more adjacent nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having a plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements. 204 and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 84 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 84 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8402. Further, the operation 8402 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by one or more void regions. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having a plurality of interconnected transportation pathways 216 defined by one or more void regions and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 85 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 85 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8502, and/or an operation 8504. The operation 8502 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of non-interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of non-interconnected transportation pathways 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 8504 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material including a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of non-interconnected pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality non-interconnected transportation pathways is defined by a region between surfaces of adjacent and substantially parallel or concentric nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of non-interconnected transportation pathways 214 defined by a region between surfaces of adjacent and substantially parallel or concentric nuclear fuel elements 204. FIG. 86 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 86 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8602, an operation 8604, and/or an operation 8606. The operation 8602 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, wherein the nuclear fuel material is consolidated to a density equal to or below a theoretical density. For example, as shown in FIGS. 1A through 4, the consolidation process (e.g., compacting, sintering, or the like) used to create the volume 202 of consolidated nuclear fuel 200 may produce a nuclear fuel piece having a selected density, wherein the selected density is less than the theoretical density. For instance, the nuclear fuel 200 may be consolidated to a density of between 65% to 99% of the theoretical density. The operation 8604 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, wherein the nuclear fuel material is contained in a geometry maintaining container. For example, as shown in FIGS. 1A through 4, the plurality of nuclear fuel elements 204 may be compacted into a fuel containing vessel or container suitable for maintaining the shape of the nuclear fuel piece. The operation 8606 illustrates consolidating the plurality of nuclear fuel elements into a self-supporting volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be formed into a self-supporting geometry via a compacting and sintering. FIG. 87 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 87 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8702, and/or an operation 8704. The operation 8702 illustrates compacting the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a metal oxide powder (e.g., uranium dioxide powder), may be placed in a mold and compacted to form a self-supporting fuel pellet. The operation 8704 illustrates sintering the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a metal oxide powder (e.g., uranium dioxide powder), may be placed in a mold and compacted and sintered to form a self-supporting fuel pellet. FIG. 88 illustrates alternative embodiments of the example operational flow 6400 of FIG. 64. FIG. 88 illustrates example embodiments where the operation 6420 may include at least one additional operation. Additional operations may include an operation 8802, an operation 8804, an operation 8806, and/or an operation 8808. The operation 8802 illustrates mechanically arranging the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a plurality of ceramic nuclear fuel elements, may be mechanically arranged into a volume 202 of nuclear fuel 200. Further, the operation 8804 illustrates weaving a plurality of linear nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2J, a plurality of nuclear fuel elements 204, such as a plurality of ceramic nuclear fuel elements, may be woven into a woven structure 224 of nuclear fuel 200. Further, the operation 8806 illustrates rolling a plurality of planar nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2I, a nuclear fuel element 204, such as a ceramic planar sheet, or a sheet containing a ceramic nuclear fuel material, may be rolled into a cylindrical volume 222. It is further recognized that two or more of the cylindrical rolled volumes 222 may be combined to form a nuclear fuel 200. Further, the operation 8808 illustrates stacking a plurality of planar nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2H, two or more nuclear fuel elements 204, such as a metal oxide or metal carbide planar sheet, may be stacked together in order to form a volume of nuclear fuel 200. FIG. 89 illustrates an operational flow 8900 representing example operations related to a method for fabricating nuclear fuel. FIG. 89 illustrates an example embodiment where the example operational flow 6400 of FIG. 64 may include at least one additional operation. Additional operations may include an operation 8910, an operation 8912, an operation 8914, and/or an operation 8916. After a start operation, an operation 6410, and an operation 6420, the operational flow 8900 moves to a processing operation 8910. Operation 8910 illustrates performing one or more processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more process steps may be performed on the volume 202 of nuclear fuel 200. The operation 8912 illustrates performing one or more material processing techniques on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the volume 202 of nuclear fuel 200 in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 8914 illustrates cold-working the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be cold-worked in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. The cold-working process may include, but is not limited to, cold-rolling, extruding at low temperature, bending, compression, or drawing. Further, the operation 8916 illustrates annealing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel material may be annealed. For instance, after being cold-worked, the nuclear reactor fuel 200 may be annealed in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 90 illustrates alternative embodiments of the example operational flow 8900 of FIG. 89. FIG. 90 illustrates example embodiments where the operation 8910 may include at least one additional operation. Additional operations may include an operation 9002, and/or an operation 9004. Further, the operation 9002 illustrates normalizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be normalized in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 9004 illustrates tempering the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be tempered in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 91 illustrates alternative embodiments of the example operational flow 8900 of FIG. 89. FIG. 91 illustrates example embodiments where the operation 8910 may include at least one additional operation. Additional operations may include an operation 9102, and/or an operation 9104. Further, the operation 9102 illustrates chemically treating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be chemically treated in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 9104 illustrates performing one or more porosity control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may undergo a porosity control process (e.g., annealing or chemical treatment). FIG. 92 illustrates alternative embodiments of the example operational flow 8900 of FIG. 89. FIG. 92 illustrates example embodiments where the operation 8910 may include at least one additional operation. Additional operations may include an operation 9202, and/or an operation 9204. The operation 9202 illustrates introducing the consolidated volume of nuclear fuel material into an elevated temperature environment. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be introduced into a high temperature environment, such as a operation within a nuclear reactor. The operation 9204 illustrates irradiating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be irradiated (e.g., irradiated in nuclear reactor implementation or irradiated via neutron source) in order to refine the sizes of the nuclear fuel elements 204 or the boundary network 214. FIG. 93 illustrates alternative embodiments of the example operational flow 8900 of FIG. 89. FIG. 93 illustrates example embodiments where the operation 8910 may include at least one additional operation. Additional operations may include an operation 9302. The operation 9302 illustrates performing a fission process utilizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be utilized in a fission process (e.g., utilized in a nuclear reactor). It is recognized that the sizes of the nuclear fission elements 204 may become more refined and/or the boundary network 214 of the nuclear fuel 200 may become more developed upon implementing the nuclear fuel 200 in a nuclear reactor 200. FIG. 94 illustrates an operational flow 9400 representing example operations related to a method for fabricating a nuclear fuel. In FIG. 94 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1A through 4, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 4. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 9400 moves to a providing operation 9410. Operation 9410 depicts providing a nuclear fuel material. For example, as shown in FIGS. 1A through 4, a variety of nuclear fuel types may be provided, including, but not limited to, metal oxide nuclear materials or metal alloy nuclear fuel materials. Moreover, the provided nuclear fuel material may have undergone processing in order to reduce the particle size of the nuclear fuel to a desirable level. For instance, a volume of nuclear fuel material may undergo ball-milling (e.g., reactive) in order to achieve a desired average particle size. Then, dispersing operation 9420 depicts dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles are configured to create preferential fission product occupation sites within the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the plurality of dispersant particles 318 may include, but is not limited to, a powder of particles of a selected material type. These particles may then be intermixed (e.g., dry mixing or wet mixing) with the provided nuclear fuel material. In another instance, the dispersant particles 318 may be dispersed into a molten metallic nuclear fuel material prior to casting of the nuclear fuel material. Then, consolidating operation 9430 depicts consolidating the nuclear fuel material into a volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains. For example, as shown in FIGS. 1A through 4, the nuclear fuel material and the intermixed dispersant particles 318 may be consolidated into a volume 102 of nuclear fuel material having a plurality of grains 104. For instance, a volume 102 of metallic nuclear fuel material 124 may be cast from a molten phase into a solid nuclear fuel piece. In another instance, a ceramic nuclear fuel material 128 may be formed during a compaction and sintering process. The consolidated volume 102 of nuclear fuel material may then be provided for further processing. Then, processing operation 9440 depicts performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, one or more process steps may be performed on the volume 102 of nuclear fuel material (e.g., fuel rod, fuel pellet, or fuel pebble) in order to reduce the grain sizes 106 of the grains 104 within the volume 102 of nuclear fuel material to a size below a critical size required for adequate diffusion of a fission product 108 from the interior 110 of the grains 104 to the grain-boundaries 112. FIG. 95 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 95 illustrates example embodiments where the operation 9420 may include at least one additional operation. Additional operations may include an operation 9502, and/or an operation 9504. The operation 9502 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles include a ceramic material. For example, the dispersant particles 318 may include one or more types of ceramic materials. Further, the operation 9504 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles include at least one of an oxide material, a nitride material, or a carbide material. For example, the dispersant particles 318 may include, but are not limited to, one or more oxide particles, nitride particles, or carbide particles. For instance, some of the dispersant particles may include a stable oxide, such as zirconium dioxide. FIG. 96 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 96 illustrates example embodiments where the operation 9420 may include at least one additional operation. Additional operations may include an operation 9602, and/or an operation 9604. The operation 9602 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles include a metallic material. For example, the dispersant particles 318 may include one or more types of metallic materials. Further, the operation 9604 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles include at least one of a metal material, a metal alloy material, or an intermetallic material. For example, the dispersant particles 318 may include, but are not limited to, one or more metal particles, metal alloy particles, or intermetallic particles. FIG. 97 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 97 illustrates example embodiments where the operation 9420 may include at least one additional operation. Additional operations may include an operation 9702, and/or an operation 9704. The operation 9702 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles are dispersed along one or more grain boundaries of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the some of the dispersant particles may be arranged such that they are localized on one or more grain-boundaries 112 of the nuclear fuel 100. The dispersant particles at the grain-boundaries 112 of the grains 104 of the nuclear fuel 100 may serve as preferential fission gas 118 occupation sites, which may facilitate “open” bubble formation along the grain-boundaries 112 during nuclear fuel operation. The operation 9704 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein some of the dispersant particles have a geometric shape. For example, as shown in FIGS. 1A through 4, the dispersant particles may have a substantially spherical shape. In a general sense, the dispersant particles may have any regular or irregular three dimensional shape. FIG. 98 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 98 illustrates example embodiments where the operation 9420 may include at least one additional operation. Additional operations may include an operation 9802. The operation 9802 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein the plurality of dispersant particles are arranged to form a low density geometric structure within a consolidated volume of the nuclear fuel material. For example, in the case of a cylindrical fuel pellet, the dispersant particles 318 may be distributed throughout the nuclear fuels 100 in a manner which produces low density cylindrical concentric shells. FIG. 99 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 99 illustrates example embodiments where the operation 9420 may include at least one additional operation. Additional operations may include an operation 9902, and/or an operation 9904. The operation 9902 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein the plurality of dispersant particles are dispersed within the nuclear fuel material prior to a solid volume forming process of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the dispersant particles 318 may be intermixed with a nuclear fuel material or a pre-cursor of a nuclear fuel material prior to being pressed. Further, the operation 9904 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein the plurality of dispersant particles are dispersed within the nuclear fuel material prior to a sintering process of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the dispersant particles 318 may be intermixed with a nuclear fuel material or a pre-cursor of a nuclear fuel material prior to being pressed sintered. FIG. 100 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 100 illustrates example embodiments where the operation 9420 may include at least one additional operation. Additional operations may include an operation 10002. Further, the operation 10002 illustrates dispersing a plurality of dispersant particles within the nuclear fuel material, wherein the plurality of dispersant particles are dispersed within the nuclear fuel material prior to a casting process of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the dispersant particles 318 may be dispersed within the volume of a molten nuclear fuel material prior to being cast. FIG. 101 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 101 illustrates example embodiments where the operation 9410 may include at least one additional operation. Additional operations may include an operation 10102, an operation 10104, and/or an operation 10106. The operation 10102 illustrates providing a ceramic nuclear fuel material. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel material may include a ceramic based nuclear fuel material. For instance, nuclear fuel may include, but is not limited to an oxide nuclear fuel material (e.g., uranium oxide), a mixed oxide fuel material (e.g., mixed plutonium oxide and depleted uranium oxide), a nitride (e.g., uranium nitride) or a carbide (e.g., uranium carbide). The operation 10104 illustrates providing a metal, metal alloy or intermetallic nuclear fuel material. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel material may include a metallic based nuclear fuel material. For instance, the plurality of nuclear fuel elements 204 of the nuclear fuel 200 may include, but is not limited to a metal (e.g., uranium, plutonium, or thorium) nuclear fuel material, a metal alloy fuel material (e.g., uranium zirconium, uranium-plutonium-zirconium, or uranium zirconium hydride), or an intermetallic (e.g., UFe2 or UNi2) based nuclear fuel material. The operation 10106 illustrates providing a nuclear fuel material including at least one of a uranium isotope, a plutonium isotope, or a thorium isotope. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel material may include a fissile nuclear material including, but not limited to, uranium-235 or plutonium-239. By way of another example, the provided nuclear fuel elements 204 may include a non-fissile nuclear material including, but not limited to, thorium-232. While thorium-232 is not by itself fissile, it may be utilized to breed uranium-233, which is fissile in nature. FIG. 102 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 102 illustrates example embodiments where the operation 9430 may include at least one additional operation. Additional operations may include an operation 10202, and/or an operation 10204. The operation 10202 illustrates consolidating the nuclear fuel material into a volume of nuclear fuel material having a surface, the consolidated nuclear fuel material including a plurality of grains, wherein some of the plurality of grains have an interfacial layer including a material different from the material of a grain interior. For example, as shown in FIG. 1J, the grains 104 of the nuclear fuel 100 may include an interfacial layer 154 of a material different than the grain-interiors 110. For instance, the grains 104 may include an oxide-based or carbide-based interfacial layer 154. The operation 10204 illustrates consolidating the nuclear fuel material into a volume of nuclear fuel material having a surface, the nuclear fuel material consolidated to density at or below a theoretical density. For example, as shown in FIGS. 1A through 4, the consolidation process (e.g., casting, compacting, sintering, or the like) used to create the volume 102 of consolidated nuclear fuel material may fabricate a nuclear fuel piece having a selected density, wherein the selected density is less than the theoretical density. For instance, the nuclear fuel material may be consolidated to a density between approximately 65 and 99% of the theoretical density FIG. 103 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 103 illustrates example embodiments where the operation 9430 may include at least one additional operation. Additional operations may include an operation 10302, and/or an operation 10304. The operation 10302 illustrates consolidating the nuclear fuel material into a volume of nuclear fuel material having a surface, the volume of nuclear fuel contained in a geometry maintaining container. For example, as shown in FIGS. 1A through 4, a casting process may consolidate a metallic nuclear fuel material inside a fuel rod, where the molten metallic nuclear fuel material may then solidify. The operation 10304 illustrates consolidating the nuclear fuel material into a solid self-supporting volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be consolidated and formed into a self-supporting geometry. FIG. 104 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 104 illustrates example embodiments where the operation 9430 may include at least one additional operation. Additional operations may include an operation 10402, and/or an operation 10404. The operation 10402 illustrates compacting the nuclear fuel material into a consolidated solid self-supporting volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be placed in a mold and compacted to form a self-supporting fuel pellet. The operation 10404 illustrates sintering the nuclear fuel material into a consolidated solid self-supporting volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be placed in a mold and compacted and sintered to form a self-supporting fuel pellet. FIG. 105 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 105 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 10502, an operation 10504, and/or an operation 10506. The operation 10502 illustrates performing one or more material processing techniques on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the volume 102 of nuclear fuel 100 in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100. Further, the operation 10504 illustrates cold-working the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may be cold-worked in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100. The cold-working process may include, but is not limited to, cold-rolling, extruding at low temperature, bending, compression, or drawing. Further, the operation 10506 illustrates annealing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be annealed. For instance, after being cold-worked, the nuclear reactor fuel 100 may be annealed in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100. FIG. 106 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 106 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 10602, and/or an operation 10604. Further, the operation 10602 illustrates normalizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may be normalized in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100. Further, the operation 10604 illustrates tempering the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may be tempered in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100. FIG. 107 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 107 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 10702, and/or an operation 10704. Further, the operation 10702 illustrates performing one or more mechanical treatment process on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may undergo a mechanical processing technique (e.g., stretching, bending, compression, or the like) in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100 Further, the operation 10704 illustrates chemically treating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may be chemically treated in order to further refine the sizes of the nuclear elements 104 or the boundary network 114 of the nuclear fuel 100. FIG. 108 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 108 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 10802, and/or an operation 10804. Further, the operation 10802 illustrates performing one or more porosity control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may undergo a porosity control process (e.g., annealing or chemical treatment). Further, the operation 10804 illustrates performing one or more grain texture control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may undergo a grain texture control process, such as annealing or chemical treatment (e.g., doping) in order to control the grain texture of the plurality of grains 104 of the nuclear fuel 100. FIG. 109 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 109 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 10902, and/or an operation 10904. The operation 10902 illustrates introducing the consolidated volume of nuclear fuel material into an elevated temperature environment. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 200 may be introduced into a high temperature environment, such as operation within a nuclear reactor. The operation 10904 illustrates irradiating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may be irradiated (e.g., irradiated in nuclear reactor implementation or irradiated via neutron source) in order to refine the sizes of the nuclear fuel elements 104 or the boundary network 114. FIG. 110 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 110 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11002, and/or an operation 11004. The operation 11002 illustrates performing a fission process utilizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel 100 may be utilized in a fission process (e.g., utilized in a nuclear reactor). It is recognized that the sizes of the grains 104 of the nuclear fuel 100 may become more refined and/or the boundary network 114 of the nuclear fuel 100 may become more developed upon implementing the nuclear fuel 100 in a nuclear reactor. The operation 11004 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along a selected dimension of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1G, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along a selected dimension of some grains 104. For instance, in grains having an elongated structure, the grains 104 may have a “thin” dimension smaller than or equal to a selected distance. FIG. 111 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 111 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11102, and/or an operation 11104. The operation 11102 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along a selected direction of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1H, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along a selected direction of some of the grains smaller than or equal to a selected distance. For instance, in grains having an elongated structure, the grains 104 may have a characteristic length 106 along a selected direction 134 with the nuclear fuel 100. For example, the grains may have a selected characteristic length 106 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). The operation 11104 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain an average characteristic length along a selected dimension of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1G, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have an average characteristic length 106 along a selected dimension of some grains 104. FIG. 112 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 112 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11202, and/or an operation 11204. The operation 11202 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain an average characteristic length along a selected direction of some of the grains smaller than or equal to a selected distance. For example, as shown in FIG. 1H, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have an average characteristic length 106 along a selected direction of some of the grains smaller than or equal to a selected distance. For instance, in grains having an elongated structure, the grains 104 may have an average characteristic length 106 along a selected direction 134 with the nuclear fuel 100. For example, the grains may have an average selected characteristic length 106 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). The operation 11204 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a selected statistical distribution of characteristic lengths. For example, as shown in FIGS. 1A through 4, the grains 104 of the nuclear fuel 100 may have a selected statistical distribution of characteristic lengths. For example, the grains 104 of the nuclear fuel 100 may have a grain size distribution having a selected percentage of the grains 104 having a grain size 106 below a selected distance. For instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 65% of the grains have a grain size 106 equal to or less than 4 μm, with an average grain size of 2.5 μm. In another example, the grains 104 of the nuclear fuel 100 may have a selected spatial distribution of characteristic lengths. FIG. 113 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 113 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11302. The operation 11302 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a selected set of statistical distributions of characteristic lengths. For example, as shown in FIGS. 1A through 4, the grains 104 of the nuclear fuel 100 may have multiple statistical distributions of characteristic lengths. For instance, the nuclear fuel 100 of the present invention may have a grain size 106 distribution such that 25% of the grains have a grain size 106 equal to or less than 2 μm, 25% of the grains have a grain size 106 equal to or less than 1 μm, and 10% of the grains are below 0.5 μm. FIG. 114 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 114 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11402, and/or an operation 11404. The operation 11402 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of an operation condition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of an operation condition of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon an operational condition of the nuclear fuel 100. Further, the operation 11404 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of an operational temperature of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of an operation temperature of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the operation temperature of the nuclear fuel 100. FIG. 115 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 115 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11502. Further, the operation 11502 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of a temperature induced pressure of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of a temperature induced pressure of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the temperature induced pressure within the nuclear reactor fuel 100. FIG. 116 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 116 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11602. The operation 11602 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of a chemical composition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of the chemical composition of the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the chemical composition (e.g., type of fissile material(s), types of alloying agents, relative concentration of fissile materials, or the like) of the nuclear reactor fuel 100. FIG. 117 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 117 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11702. The operation 11702 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is a function of a fission product generation rate of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, which is a function of the fission product 108 generation rate within the nuclear fuel 100. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel may depend upon the fission product 108 generation rate of the nuclear reactor fuel 100. Further, the fission product 108 generation rate (e.g., fission gas 118 generation rate) is proportional to the fission rate with the nuclear fuel 100, which in turn is proportional to the power density of the nuclear fuel 100, which in turn is dependent upon the chemical composition of the nuclear fuel 100. FIG. 118 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 118 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11802, and/or an operation 11804. The operation 11802 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the at least one dimension is selected to maximize heat transfer from a grain interior to a grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, wherein the dimension of the grains is selected to maximize heat transfer from the grain interiors 110 to the grain-boundaries 112 of the nuclear fuel 100. For instance, a dimension of the grains 104 to be minimize may be selected in order maximize (or at least improve) heat transfer from the grain-interiors 110 to the grain-boundaries 112. Further, the operation 11804 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel smaller than a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the grains. For instance, in order to maximize diffusion of a fission gas 118 from the grain-interiors 110 to the grain-boundaries 112 a “thin” dimension of the grains 104 may be arranged so as to align substantially perpendicular to the direction of a thermal gradient within the nuclear fuel 100. Conversely, the “thick” dimension of the grains 104 may be aligned so as to align substantially parallel with the direction of the thermal gradient within the nuclear fuel 100. FIG. 119 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 119 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 11902, and/or an operation 11904. The operation 11902 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a selected level. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration within the volume 102 of the nuclear fuel 100 at or below a selected level. For instance, in a general sense, the rate of diffusion from the grain-interiors 110 to the grain-boundaries 112 in the grains 104 may be inversely related to the average grain size 106 of the grains 104 of the nuclear fuel 100. In this sense, as the grain sizes 106 of the grains 104 decrease, the fission gas 118 diffusion rate from the grain-interiors 110 to the grain-boundaries 112 increases. Therefore, the concentration of a fission gas 118 within the grains 104 may be adjusted to fall within acceptable concentration levels by engineering the grain sizes 106 of the grains 104 of the nuclear fuel 100. Further, the operation 11904 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a concentration required for nucleation of the fission product. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration below a concentration level required for nucleation of the fission product 108 within a grain-interior 110. For instance, the concentration of a fission gas 118 within the grains 104 may be adjusted to fall below the concentration level required for fission gas nucleation with the grain-interiors 110 by engineering the grain sizes 106 of the grains 104 of the nuclear fuel 100. FIG. 120 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 120 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 12002. The operation 12002 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a gaseous fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a critical distance required for adequate diffusion of a fission gas 118 (e.g., krypton or xenon) from the grain-interiors 110 to the grain-boundaries 112 of the grains 104. FIG. 121 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 121 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 12102. The operation 12102 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a liquid fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a critical distance required for adequate diffusion of a liquid fission product 119 (e.g., liquid metal) from the grain-interiors 110 to the grain-boundaries 112 of the grains 104. FIG. 122 illustrates alternative embodiments of the example operational flow 9400 of FIG. 94. FIG. 122 illustrates example embodiments where the operation 9440 may include at least one additional operation. Additional operations may include an operation 12202. The operation 12202 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a characteristic length along at least one dimension of some of the grains smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a solid fission product from a grain interior to at least one grain boundary in some of the grains. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the grains 104 of the nuclear fuel 100 to have a characteristic length 106 along at least one dimension of some of the grains 104 of the nuclear fuel 100 smaller than a critical distance required for adequate diffusion of a solid fission product 120 (e.g., tellurium or cesium) from the grain-interiors 110 to the grain-boundaries 112 of the grains 104. FIG. 123 illustrates an operational flow 12300 representing example operations related to a method for fabricating a nuclear fuel. FIG. 123 illustrates an example embodiment where the example operational flow 9400 of FIG. 94 may include at least one additional operation. Additional operations may include an operation 12310, and/or an operation 12312. After a start operation, a providing operation 9410, a dispersing operation 9420, a consolidating operation 9430, and a processing operation 9440, the operational flow 12300 moves to a boundary formation operation 12310. Operation 12310 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a boundary network configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the consolidated volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the volume 102 of nuclear fuel 100 (e.g., fuel rod, fuel pellet, or fuel pebble) in order to form or further develop a boundary network 114 suitable for transporting a fission product 108 from the grain-boundaries 112 to the geometric surface 101 of the nuclear fuel 100. The operation 12312 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a boundary network having at least one transportation pathway configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the transportation pathway is defined by a region between two or more adjacent grains. For example, as shown in FIGS. 1A through 4, one or more processes (e.g., cold-working, annealing, or the like) may be performed on the volume 102 of nuclear fuel 100 (e.g., fuel rod, fuel pellet, or fuel pebble) in order to form or further develop a boundary network 114 having one or more transportation pathways 116 defined by a region two adjacent grains 104 suitable for transporting a fission product 108 from the grain-boundaries 112 to the geometric surface 101 of the nuclear fuel 100. FIG. 124 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 124 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 12402. The operation 12402 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a boundary network having at least one transportation pathway configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein the transportation pathway intersects with the at least one grain boundary. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the volume 102 of nuclear fuel 100 in order to form or further develop a boundary network 114 having one or more transportation pathways 116 intersecting with a grain-boundary 110 of one or more grains 104 suitable for transporting a fission product 108 from the grain-boundaries 112 to the geometric surface 101 of the nuclear fuel 100 FIG. 125 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 125 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 12502, and/or an operation 12504. The operation 12502 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a boundary network having a plurality of interconnected pathways configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the volume 102 of nuclear fuel 100 in order to form or further develop a boundary network 114 having one or more interconnected pathways suitable for transporting a fission product 108 from the grain-boundaries 112 to the geometric surface 101 of the nuclear fuel 100. Further, the operation 12504 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a boundary network having a plurality of interconnected pathways configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality of interconnected transportation pathways is defined by a region between two or more adjacent grains. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the volume 102 of nuclear fuel 100 in order to form or further develop a boundary network 114 having one or more interconnected pathways defined by the region between two or more adjacent grains 104. FIG. 126 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 126 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 12602. Further, the operation 12602 illustrates performing one or more processes on the consolidated volume of nuclear fuel material in order to obtain a boundary network having a plurality of interconnected pathways configured to transport a fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by one or more void regions. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the volume 102 of nuclear fuel 100 in order to form or further develop a boundary network 114 having one or more interconnected pathways defined by one or more void regions within the nuclear fuel 100. FIG. 127 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 127 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 12702, an operation 12704, and/or an operation 12706. The operation 12702 illustrates performing one or more material processing techniques on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be employed to reduce the grain sizes 106 of the grains 104 within the nuclear fuel 100 below a size required for adequate diffusion of a fission product 108. In another example, one or more material processing steps may be employed to form or facilitate the formation of the boundary network 114 within the nuclear reactor fuel 100. Moreover, as the grain sizes 106 decrease within the nuclear fuel 100 the number of potential transportation pathways 116 of the boundary network 114 increases, increasing the interconnection frequency within the boundary network 114 and increasing the number of pathways 116 that intersect with the geometric surface 101 of the nuclear fuel 100. Further, grain size 106 reduction and boundary network 114 formation may be carried out utilizing a single process step or multiple process steps. Further, the operation 12704 illustrates cold-working the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be cold-worked in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. The cold-working process may include, but is not limited to, cold-rolling, extruding a cast nuclear fuel material at low temperature, bending, compression, or drawing. Further, the operation 12706 illustrates annealing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be annealed in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, after being cold-worked, the nuclear reactor fuel 100 may be annealed to a temperature below the recrystallization temperature in order to achieve the desired grain size 106 within the nuclear fuel 100. In another instance, during a casting process, the nuclear reactor fuel 100 may be annealed in order to facilitate the migration of precipitating agents, such as carbon or nitrogen, out of the nuclear fuel material to the grain-boundaries 112 of the nuclear fuel 100. FIG. 128 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 128 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 12802, and/or an operation 12804. Further, the operation 12802 illustrates normalizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a normalizing process in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, after undergoing a cold-working process, the nuclear reactor fuel 100 may be annealed to a temperature above its upper critical temperature. The nuclear fuel 100 may be held at the elevated temperature for a selected amount of time and then cooled to ambient temperatures in air. Further, the operation 12804 illustrates tempering the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a tempering process in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the composition of the nuclear fuel material of the consolidated volume 102 of the nuclear fuel 100 may be suitable for precipitation of a precipitant (e.g., carbon) upon annealing. For example, a tempering process may be utilized to precipitate out a precipitating agent, such as, but not limited to, carbon. The precipitation of this agent into the grain structure of the nuclear fuel 100 may then lead to a reduction in the grain sizes 106 of the grains 104 and/or development of the boundary network 114 of the nuclear fuel 100. FIG. 129 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 129 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 12902, and/or an operation 12904. Further, the operation 12902 illustrates performing one or more mechanical treatment processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a mechanical treatment process (e.g., compression) in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. Further, the operation 12904 illustrates chemically treating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a chemical treatment process in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, a volume of 102 uranium dioxide may undergo an annealing process in the presence of an oxygen reducing gas (e.g., hydrogen-argon mixture or hydrogen-nitrogen mixture) in order to convert a portion of the stoichiometric UO2 phase to a non-stoichiometric oxygen reduced phase, such as UO1.8. The sub-stoichiometric phase has a reduced grain size with respect to the stoichiometric phase. FIG. 130 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 130 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 13002, and/or an operation 13004. Further, the operation 13002 illustrates performing one or more porosity control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a porosity control process. For instance, porosity of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., an annealing process or melting process) or a chemical treatment process. Further, the operation 13004 illustrates performing one or more grain texture control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may undergo a grain texture control process. For instance, grain textures of the grains 104 of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., annealing) or a chemical treatment process (e.g., doping). FIG. 131 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 131 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 13102, and/or an operation 13104. The operation 13102 illustrates introducing the consolidated volume of nuclear fuel material into an elevated temperature environment. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be exposed to a high temperature environment in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the consolidated volume 102 of nuclear fuel material may be implemented in a nuclear reactor setting. The nuclear fuel grain structure may be configured (e.g., cold-worked) to take advantage of the high temperature environment which occurs when the nuclear reactor fuel 100 undergoes fission. The thermal energy produced by the fission of a portion of the nuclear fuel 100 may act to reduce or further reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the thermal energy produced during a fission process of the nuclear fuel 100 may act to facilitate migration of precipitant agents, such as carbon or nitrogen, within the nuclear fuel material. Upon thermal activation, the precipitants may migrate to the grain-boundaries 112 of the nuclear fuel 100, aiding in “open” bubble formation a the grain-boundaries, leading to a development of the boundary network 114. The operation 13104 illustrates irradiating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be irradiated in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the consolidated volume 102 of nuclear fuel material may be implemented in a nuclear reactor setting. Prior to implementation in the nuclear reactor setting, the grain sizes 106 of the nuclear fuel 100 may be engineered to have a size below the critical size necessary for adequate diffusion of a produced fission gases (e.g., xenon or krypton) from the grain-interiors 110 to the grain-boundaries 112 of the nuclear fuel 100. As a result, when implemented in a nuclear reactor setting the fission gases 118 produced during the nuclear fuel 100 fission processes may efficiently nucleate at the grain-boundaries 112 of the nuclear fuel 100. This may facilitate the production of a boundary network 114 suitable for transportation of the fission gases to the geometric surface 101 of the nuclear fuel 100. FIG. 132 illustrates alternative embodiments of the example operational flow 12300 of FIG. 123. FIG. 132 illustrates example embodiments where the operation 12310 may include at least one additional operation. Additional operations may include an operation 13202. The operation 13202 illustrates performing a fission process utilizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 102 of nuclear fuel material may be utilized in nuclear reactor in order to reduce the grain sizes 106 of one or more grains 104 within the consolidated volume 102 and/or develop the boundary network 114 of the nuclear fuel 100. For instance, the elevated radiation environment and/or the high temperatures within the nuclear fuel 100 may lead to the efficient nucleation at the grain-boundaries 112 of the nuclear fuel 100. This may facilitate the production of a boundary network 114 suitable for transportation of the fission gases to the geometric surface 101 of the nuclear fuel 100. FIG. 133 illustrates an operational flow 13300 representing example operations related to a method for fabricating a nuclear fuel. In FIG. 133 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1A through 4, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 4. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 13300 moves to a providing operation 13310. Providing operation 13310 depicts providing a plurality of nuclear fuel elements, some of the plurality of nuclear fuel elements having a characteristic length along at least one dimension smaller than or equal to a selected distance, the selected distance suitable for maintaining adequate diffusion of a fission product from a nuclear fuel element interior to at least one free surface in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204 may be fabricated via a ball milling process such that their average size is smaller than a critical distance suitable for maintaining adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. For instance, a plurality of spherical nuclear fuel particles may be fabricated to have an average radius of 100 nm. Then, dispersing operation 13320 depicts dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles are configured to create preferential fission product occupation sites within the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the plurality of dispersant particles 318 may include, but is not limited to, a powder of particles of a selected material type. These particles may then be intermixed (e.g., dry mixing or wet mixing) with the provided nuclear fuel material. Then, consolidating operation 13330 depicts consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, the provided plurality of nuclear fuel elements 204 (e.g., uranium dioxide powder) and the dispersant particles 318 may be consolidated into a solid volume 202 utilizing a sintering process. FIG. 134 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 134 illustrates example embodiments where the operation 13320 may include at least one additional operation. Additional operations may include an operation 13402, and/or an operation 13404. The operation 13402 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles include a ceramic material. For example, the dispersant particles 318 may include one or more types of ceramic materials. Further, the operation 13404 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles include at least one of an oxide material, a nitride material, or a carbide material. For example, the dispersant particles 318 may include, but are not limited to, one or more oxide particles, nitride particles, or carbide particles. For instance, some of the dispersant particles may include a stable oxide, such as zirconium dioxide. FIG. 135 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 135 illustrates example embodiments where the operation 13320 may include at least one additional operation. Additional operations may include an operation 13502, and/or an operation 13504. The operation 13502 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles include a metallic material. For example, the dispersant particles 318 may include one or more types of metallic materials. Further, the operation 13504 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles include at least one of a metal material, a metal alloy material, or an intermetallic material. For example, the dispersant particles 318 may include, but are not limited to, one or more metal particles, metal alloy particles, or intermetallic particles. FIG. 136 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 136 illustrates example embodiments where the operation 13320 may include at least one additional operation. Additional operations may include an operation 13602, and/or an operation 13604. The operation 13602 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles are dispersed along one or more free surfaces of some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, some of the dispersant particles may be arranged such that they are localized on one or more surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200. The dispersant particles at the surfaces 212 of the nuclear fuel elements 212 of the nuclear fuel 200 may serve as preferential fission gas 118 occupation sites, which may facilitate an interconnected porosity within the nuclear fuel 200, leading to a boundary network 214. The operation 13604 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein some of the dispersant particles have a geometric shape. For example, as shown in FIGS. 1A through 4, the dispersant particles may have a substantially spherical shape. In a general sense, the dispersant particles may have any regular or irregular three dimensional shape. FIG. 137 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 137 illustrates example embodiments where the operation 13320 may include at least one additional operation. Additional operations may include an operation 13702. The operation 13702 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein the plurality of dispersant particles are arranged to form a low density geometric structure within a consolidated volume of the nuclear fuel material. For example, in the case of a cylindrical fuel pellet, the dispersant particles 318 may be distributed throughout the nuclear fuels 100 in a manner which produces low density cylindrical concentric shells. FIG. 138 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 138 illustrates example embodiments where the operation 13320 may include at least one additional operation. Additional operations may include an operation 13802, and/or an operation 13804. The operation 13802 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein the plurality of dispersant particles are dispersed within the nuclear fuel material prior to a solid volume forming process of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the dispersant particles 318 may be intermixed with a nuclear fuel material or a pre-cursor of a nuclear fuel material prior to being pressed. Further, the operation 13804 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein the plurality of dispersant particles are dispersed within the plurality of dispersant particles prior to a sintering process of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the dispersant particles 318 may be intermixed with a nuclear fuel material or a pre-cursor of a nuclear fuel material prior to being pressed sintered. FIG. 139 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 139 illustrates example embodiments where the operation 13320 may include at least one additional operation. Additional operations may include an operation 13902. Further, the operation 13902 illustrates dispersing a plurality of dispersant particles within the plurality of nuclear fuel elements, wherein the plurality of dispersant particles are dispersed within the plurality of dispersant particles prior to prior to a casting process of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the dispersant particles 318 may be dispersed within the volume of a molten nuclear fuel material prior to being cast. FIG. 140 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 140 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14002, an operation 14004, and/or an operation 14006. The operation 14002 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 14004 illustrates performing one or more material processing techniques on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 14006 illustrates performing one or more cold-working processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a cold-working process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The cold-working process may include, but is not limited to, cold-rolling, drawing, bending, or compression. FIG. 141 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 141 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14102. For example, as shown in FIGS. 1A through 4, an annealing process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the nuclear fuel elements 204 may be annealed in the presence of a processing gas, such as an oxygen reducing gas. FIG. 142 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 142 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14202. Further, the operation 14202 illustrates performing one or more normalizing processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a normalizing process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204, as described previously herein. FIG. 143 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 143 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14302. Further, the operation 14302 illustrates performing one or more tempering processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a tempering process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204, as described previously herein. FIG. 144 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 144 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14402. Further, the operation 14402 illustrates performing one or more chemical treatment processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a chemical treatment process may be performed on the provided nuclear fuel elements 204 in order to reduce the nuclear fuel element sizes 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. For instance, an oxygen reducing treatment may be performed on the provided nuclear fuel elements 204, as described previously herein. FIG. 145 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 145 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14502. Further, the operation 14502 illustrates performing one or more mechanical treatment processes on a plurality of nuclear fuel elements in order to achieve a characteristic length along at least one dimension smaller than or equal to a selected distance in some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a mechanical process (e.g., reactive ball milling) may be performed on the provided nuclear fuel elements 204 in order to reduce one or more dimensions of the nuclear fuel elements 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 146 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 146 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14602. Further, the operation 14602 illustrates performing one or more porosity control processes on a plurality of nuclear fuel elements in order to achieve a selected porosity within some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a porosity control process may be performed on the provided nuclear fuel elements 204 in order to achieve a selected porosity in the nuclear fuel elements 204 the nuclear fuel elements 206. For instance, porosity of the nuclear fuel 100 may be controlled via a heat treatment process (e.g., an annealing process or melting process) or a chemical treatment process. FIG. 147 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 147 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14702. Further, the operation 14702 illustrates performing one or more grain texture control processes on a plurality of nuclear fuel elements in order to achieve a selected grain texture within some of the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, a grain texture control process may be performed on the provided nuclear fuel elements 204 in order to achieve a selected grain texture in two or more grains of the nuclear fuel elements 204. For instance, grain textures of the grains of the nuclear fuel elements 204 may be controlled via a heat treatment process (e.g., annealing) or a chemical treatment process (e.g., doping). FIG. 148 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 148 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14802, and/or an operation 14804. The operation 14802 illustrates irradiating a plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, an irradiating process (e.g., exposure to neutron flux) may be performed on the provided nuclear fuel elements 204 in order to reduce one or more dimensions of the nuclear fuel elements 206 to a size below a critical size required for adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. The operation 14804 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along a selected dimension of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIG. 2D, one or more processes may be utilized in order engineer the nuclear fuel elements 204 to have a characteristic length 206 along a selected dimension of some of the nuclear fuel elements 204. For instance, in nuclear fuel elements 204 having an elongated structure, the nuclear fuel elements 204 may have a “thin” dimension that is smaller than or equal to a selected distance FIG. 149 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 149 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 14902, and/or an operation 14904. The operation 14902 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along a selected direction of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIG. 2E, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along a selected direction smaller than or equal to a selected distance. For instance, in nuclear fuel elements having an elongated structure, the nuclear fuel elements 204 may have a characteristic length 206 along a selected direction 134 within the nuclear fuel 200. For example, the nuclear fuel elements may have a selected characteristic length 206 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). The operation 14904 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain an average characteristic length along a selected dimension of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have an average characteristic length 206 along a selected dimension of some nuclear fuel elements 204. FIG. 150 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 150 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15002, and/or an operation 15004. The operation 15002 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain an average characteristic length along a selected direction of some of some of the nuclear fuel elements smaller than or equal to a selected distance. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have an average characteristic length 206 along a selected direction of some of the nuclear fuel elements 204 smaller than or equal to a selected distance. For instance, in nuclear fuel elements 204 having an elongated structure, the nuclear fuel elements 204 may have an average characteristic length 206 along a selected direction 134 with the nuclear fuel 200. For example, the nuclear fuel elements may have an average selected characteristic length 206 along the radial direction within a cylindrically shaped nuclear fuel piece (e.g., fuel pellet). The operation 15004 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a selected statistical distribution of characteristic lengths in the plurality of nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 of the nuclear fuel 200 may have a selected statistical distribution of characteristic lengths 206. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may have a element size distribution with a selected percentage of the nuclear fuel elements 204 having a size 206 below a selected distance. For instance, the nuclear fuel 200 of the present invention may have a nuclear fuel element (e.g., particle) size 206 distribution such that 65% of the nuclear fuel elements 204 have a size 206 equal to or less than 4 μm, with an average size of 2.5 μm. In another example, the nuclear fuel elements 204 of the nuclear fuel 200 may have a selected spatial distribution of characteristic lengths, within the consolidated volume of nuclear fuel 200. FIG. 151 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 151 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15102. The operation 15102 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a selected set of statistical distributions of characteristic lengths. For example, the nuclear fuel elements 204 of the nuclear fuel 200 may have multiple statistical distributions of characteristic lengths 206. For instance, the nuclear fuel 200 of the present invention may have a nuclear fuel element size 206 distribution such that 25% of the nuclear fuel elements 204 have a size equal to or less than 10 μm, 25% of the nuclear fuel elements have a nuclear fuel element size 106 equal to or less than 5 μm, and 10% of the nuclear fuel elements are below 1 μm. FIG. 152 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 152 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15202, and/or an operation 15204. The operation 15202 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of an operation condition of the nuclear fuel material For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel smaller than a selected distance, which is a function of an operation condition of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend upon an operational condition of the nuclear fuel 200. Further, the operation 15204 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of an operational temperature of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of an operation temperature of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend may depend upon the operation temperature of the nuclear fuel 200. FIG. 153 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 153 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15302. Further, the operation 15302 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a temperature induced pressure of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel smaller than a selected distance, which is a function of a temperature induced pressure of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend may depend upon the temperature induced pressure within the nuclear reactor fuel 100. FIG. 154 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 154 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15402. The operation 15402 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a chemical composition of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of the chemical composition of the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements of the nuclear fuel 200 may depend upon the chemical composition (e.g., type of fissile material(s), types of alloying agents, relative concentration of fissile materials, or the like) of the nuclear reactor fuel 200. FIG. 155 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 155 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15502. The operation 15502 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is a function of a fission product generation rate of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is a function of the fission product 108 generation rate within the nuclear fuel 200. For instance, the critical size necessary to ensure adequate diffusion of a fission product 108 from interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200 may depend upon the fission product 108 generation rate of the nuclear reactor fuel 200. Further, the fission product 108 generation rate (e.g., fission gas 118 generation rate) is proportional to the fission rate with the nuclear fuel 200, which in turn is proportional to the power density of the nuclear fuel 200, which in turn is dependent upon the chemical composition of the nuclear fuel 200. FIG. 156 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 156 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15602, and/or an operation 15604. The operation 15602 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the at least one dimension is selected to maximize heat transfer from a nuclear fuel element interior to a free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, wherein the dimension of the nuclear fuel elements is selected in order to maximize heat transfer from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 of the nuclear fuel 200. For instance, a dimension of the nuclear fuel elements 204 to be minimized may be selected in order maximize (or at least improve) heat transfer from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. Further, the operation 15604 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a nuclear fuel element interior in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, wherein the at least one dimension is selected to be substantially parallel with a thermal gradient in a grain interior in some of the nuclear fuel elements. For instance, in order to maximize diffusion of a fission gas 118 from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 a “thin” dimension of the nuclear fuel elements 204 may be arranged so as to align substantially perpendicular to the direction of a thermal gradient within the nuclear reactor fuel 200. Conversely, a “thick” dimension of the nuclear fuel elements 204 may be arranged so as to align substantially parallel with the direction of the thermal gradient within the nuclear reactor fuel 200. FIG. 157 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 157 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15702, and/or an operation 15704. The operation 15702 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level in the plurality of nuclear fuel elements necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a selected level. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 to be smaller than a selected distance, which is selected in order to maintain a selected fission product 108 (e.g., fission gas 118) concentration within the volume 102 of the nuclear fuel 100 at or below a selected level. For instance, the rate of diffusion from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 in the nuclear fuel elements 204 may be inversely related to the average nuclear fuel element size 206 within the nuclear fuel 200. In this sense, as the nuclear fuel element sizes 206 of the nuclear fuel elements 204 decrease, the fission gas 118 diffusion rate from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204 may increase. Therefore, the concentration of a fission gas 118 within the nuclear fuel elements 204 may be adjusted to fall within acceptable concentration levels by engineering the nuclear fuel element sizes 206 of the nuclear fuel elements 204 of the nuclear fuel 200. Further, the operation 15704 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining a diffusion level in the plurality of nuclear fuel elements necessary to maintain a fission product concentration within the volume of a nuclear fuel material at or below a concentration required for nucleation of the fission product. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain a selected fission product 108 concentration below a concentration level required for nucleation of the fission product 108 within an interior 210 of a nuclear fuel element 204. For instance, the concentration of a fission gas 118 within the nuclear fuel elements 204 may be adjusted to fall below the concentration level required for fission gas nucleation within the interiors 210 of the nuclear fuel elements 204 by engineering the nuclear fuel element sizes 206 of the nuclear fuel elements 204 of the nuclear fuel 200. FIG. 158 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 158 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15802. The operation 15802 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a gaseous fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a gaseous fission product (e.g., krypton or xenon) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 159 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 159 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 15902. The operation 15902 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a liquid fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a liquid fission product (e.g., a liquid metal) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 160 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 160 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16002. The operation 16002 illustrates performing one or more processes on a plurality of nuclear fuel elements in order to obtain a characteristic length along at least one dimension of some of the plurality of nuclear fuel elements smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a solid fission product from a nuclear fuel element interior to at least one free surface of a nuclear fuel element in some of the nuclear fuel elements. For example, as shown in FIGS. 1A through 4, one or more processes may be utilized in order engineer the nuclear fuel elements 204 of the nuclear fuel 200 to have a characteristic length 206 along at least one dimension of some of the nuclear fuel elements 204 of the nuclear fuel 200 smaller than a selected distance, which is selected in order to maintain adequate diffusion of a solid fission product (e.g., tellurium or cesium) from the interiors 210 of the nuclear fuel elements 204 to the surfaces 212 of the nuclear fuel elements 204. FIG. 161 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 161 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16102. The operation 16102 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements have an interfacial layer including a material different from an interior of a nuclear fuel element. For example, as shown in FIG. 2G, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include an interfacial layer of a material different from the material within the interiors 210 of the nuclear fuel elements 204. For instance, the nuclear fuel elements 204 may include an oxide-based or carbide-based interfacial layer. FIG. 162 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 162 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16202. The operation 16202 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include two or more grains. For example, as shown in FIG. 2F, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include two or more grains (i.e., the nuclear fuel elements are polycrystalline). FIG. 163 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 163 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16302. The operation 16302 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include a plurality of pathways configured to transport a fission product from at least one portion of a nuclear fuel element interior to at least one free surface of the nuclear fuel element. For example, as shown in FIG. 2F, one or more of the nuclear fuel elements 204 of the nuclear fuel 200 may include one or more internal pathways suitable for transporting fission gas 118 from the nuclear fuel element interior 210 to the nuclear fuel element surface 212. Moreover, as previously described herein, the internal pathways 110 may be defined by a grain-boundary 112 between adjacent grains within a common nuclear fuel element 204. FIG. 164 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 164 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16402. The operation 16402 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements includes a ceramic nuclear fuel material. For example, as shown in FIGS. 1A through 4, some of the nuclear fuel elements 204 may include, but are not limited to a metal oxide (e.g., uranium dioxide, plutonium dioxide, or thorium dioxide) nuclear fuel material, a mixed oxide nuclear fuel material (e.g., blend of plutonium dioxide and depleted uranium dioxide), a metal nitride (e.g., uranium nitride) based nuclear fuel material, or a metal carbide (e.g., uranium carbide) based nuclear fuel material. FIG. 165 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 165 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16502. The operation 16502 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include at least one of a metal, a metal alloy, or an intermetallic nuclear fuel material. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel elements 204 may include a metal based nuclear fuel material. For instance, the plurality of nuclear fuel elements 204 of the nuclear fuel 200 may include, but is not limited to a metal (e.g., uranium, plutonium, or thorium) nuclear fuel material, a metal alloy fuel material (e.g., uranium zirconium, uranium-plutonium-zirconium, or uranium zirconium hydride), or an intermetallic (e.g., UFe2 or UNi2) based nuclear fuel material. FIG. 166 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 166 illustrates example embodiments where the operation 13310 may include at least one additional operation. Additional operations may include an operation 16602. The operation 16602 illustrates providing a plurality of nuclear fuel elements, wherein some of the plurality of nuclear fuel elements include at least one of a uranium isotope, a plutonium isotope, or a thorium isotope. For example, as shown in FIGS. 1A through 4, the provided nuclear fuel elements 204 may include a fissile nuclear material including, but not limited to, uranium-235 or plutonium-239. By way of another example, the provided nuclear fuel elements 204 may include a non-fissile nuclear material including, but not limited to, thorium-232. While thorium-232 is not by itself fissile, it may be utilized to breed uranium-233, which is fissile in nature. FIG. 167 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 167 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 16702, and/or an operation 16704. The operation 16702 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 16704 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes at least one transportation pathway defined by a region between two or more adjacent nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216. FIG. 168 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 168 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 16802. Further, the operation 16802 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes at least one transportation pathway intersecting the at least one free surface. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216 intersecting a surface 212 of one or more nuclear fuel elements 204. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for providing a boundary network 214 having at least one transportation pathway 216 intersecting a surface 212 of one or more nuclear fuel elements 204. FIG. 169 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 169 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 16902, and/or an operation 16904. Further, the operation 16902 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 plurality of interconnected transportation pathways 216 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 16904 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by a region between two or more adjacent nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having a plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements. 204 and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 170 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 170 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 17002. Further, the operation 17002 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of interconnected pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality interconnected transportation pathways is defined by one or more void regions. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a consolidation process, such as, but not limited to, a compacting process, or a sintering process, configured to provide a porosity level within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having a plurality of interconnected transportation pathways 216 defined by one or more void regions and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of interconnected transportation pathways 216 defined by a region between two or more adjacent nuclear fuel elements and configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. FIG. 171 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 171 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 17102, and/or an operation 17104. Further, the operation 17102 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of non-interconnected transportation pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of non-interconnected transportation pathways 214 configured to transport a fission product 118 from the surfaces 212 of the nuclear fuel elements 204 to the geometric surface 201 of the nuclear fuel 200. Further, the operation 17104 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a boundary network configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein the boundary network includes a plurality of non-interconnected pathways configured to transport a fission product from at least one free surface of some of the nuclear fuel elements to the surface of the volume of the nuclear fuel material, wherein at least one of the plurality non-interconnected transportation pathways is defined by a region between surfaces of adjacent and substantially parallel or concentric nuclear fuel elements. For example, as shown in FIGS. 1A through 4, the nuclear fuel elements 204 may be consolidated via a mechanical process configured to provide spatial configuration within the consolidated nuclear fuel 200 suitable for producing a boundary network 214 having plurality of non-interconnected transportation pathways 214 defined by a region between surfaces of adjacent and substantially parallel or concentric nuclear fuel elements 204. FIG. 172 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 172 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 17202, and/or an operation 17204. The operation 17202 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, wherein the nuclear fuel material is consolidated to a density equal to or below a theoretical density. For example, as shown in FIGS. 1A through 4, the consolidation process (e.g., compacting, sintering, or the like) used to create the volume 202 of consolidated nuclear fuel 200 may produce a nuclear fuel piece having a selected density, wherein the selected density is less than the theoretical density. For instance, the nuclear fuel 200 may be consolidated to a density of 95% of the theoretical density. The operation 17204 illustrates consolidating the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface, wherein the nuclear fuel material is contained in a geometry maintaining container. For example, as shown in FIGS. 1A through 4, the plurality of nuclear fuel elements 204 may be compacted into a fuel containing vessel or container suitable for maintaining the shape of the nuclear fuel piece. FIG. 173 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 173 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 17302, and/or an operation 17304. The operation 17302 illustrates consolidating the plurality of nuclear fuel elements into a self-supporting volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a metal oxide powder, such as uranium-dioxide, may be formed into a self-supporting geometry via a compacting and sintering. The operation 17304 illustrates compacting the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a metal oxide powder (e.g., uranium dioxide powder), may be placed in a mold and compacted to form a self-supporting fuel pellet. FIG. 174 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 174 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 17402. The operation 17402 illustrates sintering the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a metal oxide powder (e.g., uranium dioxide powder), may be placed in a mold and compacted and sintered to form a self-supporting fuel pellet. FIG. 175 illustrates alternative embodiments of the example operational flow 13300 of FIG. 133. FIG. 175 illustrates example embodiments where the operation 13330 may include at least one additional operation. Additional operations may include an operation 17502, an operation 17504, an operation 17506, and/or an operation 17508. The operation 17502 illustrates mechanically arranging the plurality of nuclear fuel elements into a volume of nuclear fuel material having a surface. For example, as shown in FIGS. 1A through 4, a plurality of nuclear fuel elements 204, such as a plurality of metal (e.g., thorium) or metal alloy (e.g., uranium alloy) nuclear fuel elements, may be mechanically arranged into a volume 202 of nuclear fuel 200. Further, the operation 17504 illustrates weaving a plurality of linear nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2J, a plurality of nuclear fuel elements 204, such as a plurality of metal (e.g., thorium) or metal alloy (e.g., uranium alloy) nuclear fuel elements, may be woven into a woven structure 224 of nuclear fuel 200. Further, the operation 17506 illustrates rolling a plurality of planar nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2I, a nuclear fuel element 204, such as a metal or metal alloy planar sheet, may be rolled into a cylindrical volume 222. It is further recognized that two or more of the cylindrical rolled volumes 222 may be combined to form a nuclear fuel 200. Further, the operation 17508 illustrates stacking a plurality of planar nuclear fuel elements into a solid volume of nuclear fuel material having a surface. For example, as shown in FIG. 2H, two or more nuclear fuel elements 204, such as a metal or metal alloy planar sheet, may be stacked together in order to form a volume of nuclear fuel 200. FIG. 176 illustrates an operational flow 17600 representing example operations related to a method for fabricating a nuclear fuel. FIG. 176 illustrates an example embodiment where the example operational flow 13300 of FIG. 133 may include at least one additional operation. Additional operations may include an operation 17610, an operation 17612, an operation 17614, and/or an operation 17616. After a start operation, a providing operation 13310, a dispersing operation 13320, and a consolidation operation 13330, the operational flow 17600 moves to a processing operation 17610. Operation 17610 illustrates performing one or more processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more processing techniques may be performed on the volume 202 of nuclear fuel 200 in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. The operation 17612 illustrates performing one or more material processing techniques on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, one or more material processing techniques may be performed on the volume 202 of nuclear fuel 200 in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 17614 illustrates cold-working the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be cold-worked in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. The cold-working process may include, but is not limited to, cold-rolling, extruding at low temperature, bending, compression, or drawing. Further, the operation 17616 illustrates annealing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel material may be annealed. For instance, after being cold-worked, the nuclear reactor fuel 200 may be annealed in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 177 illustrates alternative embodiments of the example operational flow 17600 of FIG. 176. FIG. 177 illustrates example embodiments where the operation 17610 may include at least one additional operation. Additional operations may include an operation 17702, and/or an operation 17704. Further, the operation 17702 illustrates melting a portion of the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, a portion of the consolidated volume 202 of nuclear fuel 200 may be melted. Further, the operation 17704 illustrates normalizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be normalized in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 61 illustrates alternative embodiments of the example operational flow 5900 of FIG. 59. FIG. 61 illustrates example embodiments where the operation 5910 may include at least one additional operation. Additional operations may include an operation 6102, and/or an operation 6104. FIG. 178 illustrates alternative embodiments of the example operational flow 17600 of FIG. 176. FIG. 178 illustrates example embodiments where the operation 17610 may include at least one additional operation. Additional operations may include an operation 17802, and/or an operation 17804. Further, the operation 17802 illustrates tempering the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be tempered in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. Further, the operation 17804 illustrates chemically treating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be chemically treated in order to further refine the sizes of the nuclear elements 204 or the boundary network 214 of the nuclear fuel 200. FIG. 179 illustrates alternative embodiments of the example operational flow 17600 of FIG. 176. FIG. 179 illustrates example embodiments where the operation 17610 may include at least one additional operation. Additional operations may include an operation 17902. Further, the operation 17902 illustrates performing one or more porosity control processes on the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may undergo a porosity control process (e.g., annealing or chemical treatment). FIG. 180 illustrates alternative embodiments of the example operational flow 17600 of FIG. 176. FIG. 180 illustrates example embodiments where the operation 17610 may include at least one additional operation. Additional operations may include an operation 18002, and/or an operation 18004. The operation 18002 illustrates introducing the consolidated volume of nuclear fuel material into an elevated temperature environment. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be introduced into a high temperature environment, such as a operation within a nuclear reactor. The operation 18004 illustrates irradiating the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be irradiated (e.g., irradiated in nuclear reactor implementation or irradiated via neutron source) in order to refine the sizes of the nuclear fuel elements 204 or the boundary network 114. FIG. 181 illustrates alternative embodiments of the example operational flow 17600 of FIG. 176. FIG. 181 illustrates example embodiments where the operation 17610 may include at least one additional operation. Additional operations may include an operation 18102. The operation 18102 illustrates performing a fission process utilizing the consolidated volume of nuclear fuel material. For example, as shown in FIGS. 1A through 4, the consolidated volume 202 of nuclear fuel 200 may be utilized in a fission process (e.g., utilized in a nuclear reactor). It is recognized that the sizes of the nuclear fission elements 204 may become more refined and/or the boundary network 114 of the nuclear fuel 200 may become more developed upon implementing the nuclear fuel 200 in a nuclear reactor 200. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware, software, and/or firmware implementations of aspects of systems; the use of hardware, software, and/or firmware is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. In some implementations described herein, logic and similar implementations may include software or other control structures. Electronic circuitry, for example, may have one or more paths of electrical current constructed and arranged to implement various functions as described herein. In some implementations, one or more media may be configured to bear a device-detectable implementation when such media hold or transmit device-detectable instructions operable to perform as described herein. In some variants, for example, implementations may include an update or modification of existing software or firmware, or of gate arrays or programmable hardware, such as by performing a reception of or a transmission of one or more instructions in relation to one or more operations described herein. Alternatively or additionally, in some variants, an implementation may include special-purpose hardware, software, firmware components, and/or general-purpose components executing or otherwise invoking special-purpose components. Specifications or other implementations may be transmitted by one or more instances of tangible transmission media as described herein, optionally by packet transmission or otherwise by passing through distributed media at various times. Alternatively or additionally, implementations may include executing a special-purpose instruction sequence or invoking circuitry for enabling, triggering, coordinating, requesting, or otherwise causing one or more occurrences of virtually any functional operations described herein. In some variants, operational or other logical descriptions herein may be expressed as source code and compiled or otherwise invoked as an executable instruction sequence. In some contexts, for example, implementations may be provided, in whole or in part, by source code, such as C++, or other code sequences. In other implementations, source or other code implementation, using commercially available and/or techniques in the art, may be compiled/implemented/translated/converted into a high-level descriptor language (e.g., initially implementing described technologies in C or C++ programming language and thereafter converting the programming language implementation into a logic-synthesizable language implementation, a hardware description language implementation, a hardware design simulation implementation, and/or other such similar mode(s) of expression). For example, some or all of a logical expression (e.g., computer programming language implementation) may be manifested as a Verilog-type hardware description (e.g., via Hardware Description Language (HDL) and/or Very High Speed Integrated Circuit Hardware Descriptor Language (VHDL)) or other circuitry model which may then be used to create a physical implementation having hardware (e.g., an Application Specific Integrated Circuit). Those skilled in the art will recognize how to obtain, configure, and optimize suitable transmission or computational elements, material supplies, actuators, or other structures in light of these teachings. The foregoing detailed description has set forth various embodiments of the devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In one embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link (e.g., transmitter, receiver, transmission logic, reception logic, etc.), etc.). In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, and/or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, electro-magnetically actuated devices, and/or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, a Micro Electro Mechanical System (MEMS), etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.), and/or any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, medical devices, as well as other systems such as motorized transport systems, factory automation systems, security systems, and/or communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, and/or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Those skilled in the art will recognize that at least a portion of the devices and/or processes described herein can be integrated into a data processing system. Those having skill in the art will recognize that a data processing system generally includes one or more of a system unit housing, a video display device, memory such as volatile or non-volatile memory, processors such as microprocessors or digital signal processors, computational entities such as operating systems, drivers, graphical user interfaces, and applications programs, one or more interaction devices (e.g., a touch pad, a touch screen, an antenna, etc.), and/or control systems including feedback loops and control motors (e.g., feedback for sensing position and/or velocity; control motors for moving and/or adjusting components and/or quantities). A data processing system may be implemented utilizing suitable commercially available components, such as those typically found in data computing/communication and/or network computing/communication systems. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken limiting. Although a user is shown/described herein as a single illustrated figure, those skilled in the art will appreciate that the user may be representative of a human user, a robotic user (e.g., computational entity), and/or substantially any combination thereof (e.g., a user may be assisted by one or more robotic agents) unless context dictates otherwise. Those skilled in the art will appreciate that, in general, the same may be said of “sender” and/or other entity-oriented terms as such terms are used herein unless context dictates otherwise. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that such terms (e.g., “configured to”) can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B. With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. |
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056278667 | summary | TECHNICAL FIELD This invention relates to the structure of a fuel assembly in a boiling water nuclear reactor vessel and, more particularly, to a fuel assembly structure that utilizes the channel to support its load, thereby eliminating the need for fuel tie rods. BACKGROUND A conventional fuel assembly in a boiling water nuclear reactor vessel includes a lower tie plate, an upper tie plate and a matrix of the sealed fuel rods supported between the upper and lower tie plates. The fuel rods contain nuclear fuel pellets in sealed containment for supporting a required critical reaction for the generation of steam. One or more coolant rods is included in the matrix of the fuel rods and is also supported between the upper and lower tie plates. A channel surrounds the tie plates, fuel rods and coolant rod. This channel is commonly square in cross-section and made of metal (preferably an alloy called Zircaloy). Water passes from the bottom of the fuel assembly to the top of the fuel assembly. Water enters through the lower tie plate within the channel and passes between the upstanding fuel rods. Water and generated steam exit from within the channel between the fuel rods and out through the upper tie plate. The channel confines the required moderator coolant flow to a flow path that is restricted between the tie plates. The lower tie plate and the upper tie plate serve to support the sealed fuel rods in the vertical and upstanding matrix. Typically, the upper tie plate forms an overlying matrix of fuel rod support points. Into about eight of these support points are conventionally placed correspondingly male threaded tie rods and fittings. The tie rods, which contain fuel similar to the fuel rods, are threaded at their lower ends for corresponding attachment to the lower tie plate. The lower tie plate similarly forms an underlying matrix of fuel rod support points. These underlying support points correspond for the most part to the overlying support points of the upper tie plate. Conventionally, about eight of these support points are threaded with female apertures, which correspond to the overlying apertures in the upper tie plates. Into these threaded support points in the lower tie plates are placed the lower threaded ends of the fuel tie rods. Thus, conventionally, the two tie plates are tied together with the fuel tie rods. The tie plates also define a matrix of apertures for permitting fluid flow into and out of the fuel assembly. Specifically, the lower tie plate defines a first matrix of apertures for permitting the in flow of water coolant. This coolant functions in the capacity of moderating or slowing down reaction produced fast neutrons to produce reaction continuing slow or thermal neutrons. At the same time, as the coolant passes upwardly through the fuel assembly within the channel, a portion of the coolant is turned to steam. This steam and the coolant that is not turned into steam and remains in the liquid phase must pass out through the upper tie plate. Consequently, the upper tie plate forms its own matrix of apertures in between its matrix of fuel rod support points. The upper tie plate matrix of apertures permits the out flow of the two phase steam/water mixture from the fuel assembly. The fuel bundle must be periodically replaced and/or inspected during so-called "outages" of a reactor. These outages occur when the central steam generating core of a nuclear reactor has its overlying component removed to provide access through shielding water to the core. During such "outages," sections of the reactor vessel core are removed, inspected and/or replaced. The core, submerged in a radiation quenching bath of water, has the fuel bundles to be replaced for inspection removed by remotely grasping the fuel assembly at a handle. The handle must define, at the top of the fuel assembly, a support point for the entire weight of the fuel assembly in a depending relationship when the assembly is removed from the vessel. Once the fuel assembly is supported at the handle, the entire weight of the fuel assembly is carried through the handle. This weight includes the weight of the fuel and coolant rods, the weight of the upper tie plate, the weight of the lower tie plate and the weight of the surrounding channel (upwards of 600 pounds). Once the fuel assembly is removed from the vessel, the tie plates, fuel rods and coolant rods can be separated from the channel. After separation from the channel, the fuel rods can easily be inspected and/or replaced. Conventionally, however, the threaded end plugs of the fuel tie rods tend to seize in their threaded connections, thus making replacement of the fuel tie rods difficult and time consuming. Moreover, as fuel assembly design lifetimes are extended, corrosion effects weaken the fuel tie rods. This weakening occurs due to corrosion thinning of the material and by a reduction in ductility due to the formation of hydrogen and its absorption. Thus, there is a need to provide a fuel assembly structure that does not include fuel tie rods threadedly connected between the upper and lower tie plates. Moreover, there is a need to utilize a structural load path for the fuel assembly that is less affected by corrosion effects. In general, since corrosion is a surface phenomena, a structure with a high volume to surface area provides more margin in this regard. Without adding additional structure to the general design of boiling water reactor fuel assemblies, the component with the highest volume to surface area is the channel. DISCLOSURE OF THE INVENTION It is therefore an object of the invention to provide a fuel assembly structure that utilizes the channel as the structural member that limits the axial extension between the upper and lower tie plates. Another object of the invention is to provide a fuel assembly structure that provides redundancy in the attaching features of the upper tie plate to the channel as well as between the transition member and the channel and allows lifting of the assembly through a load path that does not utilize fuel tie rods that are threaded into the tie plates. It is yet another object of the invention to provide a method of removing the fuel bundle from the fuel assembly. These and other objects and advantages of the invention are achieved by providing a fuel assembly for a nuclear reactor vessel that includes a plurality of fuel rods; a coolant rod; a channel surrounding the plurality of fuel rods and the coolant rod; and a lower tie plate supporting the plurality of fuel rods and the coolant rod, the lower tie plate being supported by the channel such that the channel carries a load of the fuel assembly. The fuel assembly may further include an upper tie plate disposed inside the channel that laterally supports the plurality of fuel rods and the coolant rod; and connecting structure releasably connecting the upper tie plate and the channel. The upper tie plate may be provided with a bolt aperture adjacent the coolant rod that opens to a spring channel in the upper tie plate. In this regard, the fuel assembly further includes a coolant rod main spring disposed surrounding the coolant rod; a coolant rod main spring support supporting the coolant rod main spring; a bolt inserted in the bolt aperture and extending into the spring channel; a substantially cylindrical member fixed to an end of the bolt and delimiting the spring channel; and a spring disposed surrounding the bolt in the channel between the coolant rod main spring support and an end of the spring channel. The connecting structure can include two opposed extendible and retractable latch pins, wherein the channel has a corresponding two opposed apertures for receiving the latch pins. The upper tie plate preferably has two boss members, which may be integral with the upper tie plate, each having a channel therein, wherein each of the latch pins is movably disposed in the each of the channels, respectively. Two springs may be provided, one each disposed in each of the channels surrounding a respective one of the latch pins. The springs urge the latch pins toward an extended position, the latch pins having a first outer periphery configured to be received in the apertures, respectively, and a second outer periphery, larger than the first outer periphery, configured to sit against the channel in the extended position. In this regard, each of the latch pins may include a spring engaging surface delimiting the channels, respectively, wherein the springs are disposed in the channels between an end surface of the channels and the spring engaging surfaces, respectively. The fuel assembly may further include a transition member partially disposed in the channel and adjacent the lower tie plate. The transition member is attached to the channel, and the lower tie plate rests on the transition member in this regard, a bolt may be provided threadedly secured through the channel and the transition member. Still further, a channel clip may be provided fixed to the channel and inserted into a slot in the transition, wherein the channel clip is partially disposed axially interior of the channel, thereby restricting axial displacement of the transition member relative to the channel. In a preferred embodiment, four channel clips are welded to the channel. In accordance with another aspect of the invention, the connecting structure includes a channel guide fixed to the upper tie plate and disposed outside of the channel. The channel guide includes an ear, and the channel includes a corresponding ear aperture configured to receive the ear. The channel guide may be provided with two arms disposed at substantially 90.degree. and two ears, each fixed to a respective one of the two arms. The channel guide may include two legs extending from ends of the two arms, respectively, wherein the two ears are each fixed to one of the two legs, respectively. In accordance with yet another aspect of the invention, there is provided a fuel assembly for a nuclear reactor vessel that includes a plurality of fuel rods; a coolant rod; a channel surrounding the plurality of fuel rods and the coolant rod; a lower tie plate supporting the plurality of fuel rods and the coolant rod, the lower tie plate being supported by the channel such that the channel carries a load of the fuel assembly; an upper tie plate disposed inside the channel and laterally supporting the plurality of fuel rods and the coolant rod; a transition member partially disposed in the channel and adjacent the lower tie plate, the transition member being attached to the; channel, wherein the lower tie plate rests on the transition member; and connecting structure releasably connecting the upper tie plate and the channel, wherein the connecting structure includes a channel guide fixed to the upper tie plate and disposed outside of the channel, the channel guide including an ear, and the channel including a corresponding ear aperture configured to receive the ear. The connecting structure may further include an extendible and retractable latch pin, and the channel may include a corresponding aperture for receiving the; latch pin. Finally, in accordance with a further aspect of the invention, there is provided a method of removing a fuel bundle from a fuel assembly, the fuel bundle including a plurality of fuel rods, a coolant rod, and a lower tie plate supporting the plurality of fuel rods and the coolant rod, the coolant rod being fixed to the lower tie plate. The fuel assembly includes a channel surrounding the fuel bundle, and the lower tie plate is supported by the channel such that the channel carries a load of the fuel assembly. The method includes detaching the lower tic; plate from being supported by the channel; and removing the fuel bundle by lifting the coolant rod with a lifting tool. The fuel assembly may further include an upper tie plate disposed inside of the channel and laterally supporting the plurality of fuel rods and the coolant rod, and connecting structure releasably connecting the upper tie plate and the channel. In this regard, the detaching step may include removing the connecting structure to detach the upper tie plate and the channel. |
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claims | 1. A system for storing spent nuclear fuel comprising: a lid having a side wall with substantially horizontal ventilation means extending through the side wall, the lid having a bottom surface; a ventilated vertical overpack having a body including lower ventilation ducts, a bottom, a top surface, and a chamber formed by the body and the bottom adapted for receiving a canister of spent nuclear fuel; and wherein the lid is secured to the overpack by a plurality of fasteners so that the bottom surface of the lid and the top surface of the overpack form an interface comprising a substantially horizontal section. 2. A system for storing spent nuclear fuel comprising: a lid having a side wall with substantially horizontal ventilation means extending through the side wall, the lid having a bottom surface; a ventilated vertical overpack having a body including lower ventilation ducts a bottom, a top surface, and a chamber formed by the body and the bottom adapted for receiving a canister of spent nuclear fuel; a lid shear ring having a lower surface attached to the bottom surface of the lid; a body shear ring having an upper surface attached to the top surface of the overpack; and the lid being secured to the overpack by a plurality of fastener so that the upper surface of the body shear ring contacts the bottom surface of the lid forming a first substantially horizontal interface and the lower surface of the lid shear ring contacts the top surface of the overpack forming a second substantially horizontal interface. 3. The system of claim 1 wherein when the canister of spent nuclear fuel is received in the overpack and the lid is secured, air within the chamber is warmed by heat from the spent nuclear fuel, cold air entering through the lower ventilation ducts and warmed air exiting through the ventilation means of the lid. claim 1 4. A method of storing spent nuclear fuel comprising: placing a canister of spent nuclear fuel in a chamber of an overpack of a system comprising a lid having a side wall with substantially horizontal ventilation means extending through the side wall, the lid having a bottom surface, the overpack having a body including lower ventilation ducts, a bottom, and a top surface, the chamber being formed by the body and the bottom, wherein the lid is secured to the overpack by a plurality of fasteners; and securing the lid to the overpack so that the bottom surface of the lid and the top surface of the overpack form an interface comprising at least one substantially horizontal section so that air within the chamber is warmed by heat from the spent nuclear fuel, cold air entering through the lower ventilation ducts and warmed air exiting through the ventilation means. 5. The method of claim 4 wherein the fasteners are bolts and the securing step comprises bolting the lid to the overpack. claim 4 6. The system of claim 1 wherein the lower ventilation ducts are located in the body near the bottom and no upper ventilation ducts are included in the body of the overpack. claim 1 7. The system of claim 2 wherein the lower ventilation ducts are located in the body near the bottom and no upper ventilation ducts are included in the body of the overpack. claim 2 8. The system of claim 2 wherein the body of the overpack is cylindrical. claim 2 9. The system of claim 1 wherein the lid comprises a lid cap and a lid body. claim 1 10. The system of claim 9 wherein the lid body is an annular ring. claim 9 11. The system of claim 1 wherein the body of the overpack is cylindrical. claim 1 12. The system of claim 1 wherein the fasteners are bolts extending through the lid and threadily engaging the overpack. claim 1 13. The system of claim 1 wherein the ventilation means are four ventilation ducts. claim 1 14. The system of claim 2 wherein the lid shear ring engages the body shear ring so as to prevent lateral movement of the lid with respect to the overpack. claim 2 15. The system of claim 2 wherein the lid comprises a lid body and a lid cap. claim 2 16. The system of claim 15 wherein the lid body is an annular ring. claim 15 17. The system of claim 2 wherein the fasteners are bolts extending through the lid and threadily engaging the overpack. claim 2 18. The system of claim 2 wherein the ventilation means are four ventilation ducts. claim 2 19. A method of storing spent nuclear fuel comprising: placing a canister of spent nuclear fuel in a chamber of an overpack of a system comprising a lid having a side wall with substantially horizontal ventilation means extending through the side wall, the lid having a bottom surface and a lid shear ring having a lower surface attached to the bottom surface, the overpack having a body including lower ventilation ducts, a bottom, a top surface, and a body shear ring having an upper surface attached to the top surface, the chamber being formed by the body and the bottom; and securing the lid to the overpack with a plurality of fasteners so that the lower surface of the lid shear ring contacts the top surface of the overpack forming a first substantially horizontal interface, and the upper surface of the body shear ring contacts the bottom surface of the lid forming a second substantially horizontal interface so that air within the chamber is warmed by heat from the spent nuclear fuel, cold air entering through the lower ventilation ducts and warmed air exiting through the ventilation means. 20. The method of claim 19 wherein the fasteners are bolts and the securing step comprises bolting the lid to the overpack. claim 19 |
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048428086 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear fuel rods for use in nuclear reactors and, more particularly, is concerned with a system and method for collating nuclear fuel pellets so as to facilitate subsequent filling of fuel rods with pellets in multiple zones of different nuclear fuel enrichments. 2. Description of the Prior Art In a typical nuclear reactor, the reactor core includes a large number of elongated fuel assemblies. Conventional designs of these fuel assemblies include top and bottom nozzles with a plurality of elongated transversely spaced guide thimbles extending longitudinally between and connected at opposite ends to the nozzles and a plurality of transverse support grids axially spaced along the guide thimbles. Also, each fuel assembly is composed of a multiplicity of elongated fuel elements or rods transversely spaced apart from one another and from the guide thimbles and supported by the transverse grids between the top and bottom nozzles. The fuel rods each contain fissile material in the form of a plurality generally cylindrical nuclear fuel pellets maintained in a row or stack thereof in the rod. The fuel rods are grouped together in an array which is organized so as to provide a neutron flux in the core sufficient to support a high rate of nuclear fission and thus the release of a large amount of energy in the form of heat. A liquid coolant is pumped upwardly through the core in order to extract some of the heat generated in the core for the production of useful work. Up to the present time, generally a given fuel rod has been filled with nuclear fuel pellets of the same enrichment. In filling a typical fuel rod, pellets of the same enrichment are continually fed from supply trays into the fuel rod until the specified depth of fill has been achieved. The supply trays are used up completely. A partial tray left after completing a batch of twenty-five fuel rods is used to start the fill of the next batch. Various systems have been used in the past to fill the fuel rods with fuel pellets. Representative of the prior art are the systems disclosed in U.S. Pat. Nos. to Gerkey (4,158,601)--assigned to the assignee of the present invention--and Gheri (4,495,146), and Japanese Pat. No. 61-4999 to Mitsubishi. Recently, nuclear fuel rods with pellets in multiple zones of different enrichments have been introduced. These new zoned fuel rods contain short lengths of "blanket" pellets at each end. The fuel rods additionally have pellets grouped within from three to five zones of different enrichments between the end zones of blanket pellets. Pellets for filling these enrichment-zoned fuel rods will first need to be assembled into the desired sequences of enrichment zones before filling of the fuel rods can commence. One approach to assembling fuel pellets in the desired sequence of enrichment zones for filling a fuel rod is by manual effort. However, this approach envisions individual handling of each pellet which is highly labor-intensive and time-consuming and subject to human error. A much more desirable approach is to assemble the different zones of fuel pellet by use of a highly-mechanized and automated system which would be substantially free of human error. Consequently, a need exists for an automated system capable of assembling pellets of different enrichments into the desired sequence of enrichment zones of pellets so that specified sequences of pellets will be produced for subsequent loading into fuel rods. SUMMARY OF THE INVENTION The present invention provides a fuel pellet collating system and method designed to satisfy the aforementioned needs. The collating system and method of the present invention is adapted to handle and transfer trays of pellets of various enrichments, to measure, record, and sum the lengths of the various pellet zones, and to load these measured rows or stacks of zoned pellets into a collated storage and transport cabinet so that pellets in the specified sequence of enrichment zones will be subsequently loaded into the fuel rods. In the pellet collating system and method of the present invention, a multiplicity of rows of pellets, for instance twenty-five rows, are handled simultaneously. Accordingly, the present invention is directed to a system of collating nuclear fuel pellets which comprises: (a) means for positioning a plurality of pellet supply trays and a plurality of pellet storage trays, each supply tray being adapted to support in at least one row thereon a plurality of pellets of an enrichment different from the enrichments of pellets on at least some other of said supply trays, each storage tray being adapted to support in at least one row thereon a plurality of pellets of an enrichment different from the enrichments of pellets on at least some other of said storage trays; (b) a pellet collating line including a serially-arranged pellet input station, pellet measuring and collating work station and pellet output station; (c) means for transferring supply trays one at time between the tray positioning means and the input station and for transferring storage trays one at a time between the tray positioning means and the output station; and (d) pellet collating means disposed adjacent the pellet collating line and being operable for moving pellets in the at least one row thereof onto the work station from a given one supply tray on the input station, for measuring a desired length of pellets on the work station and separating the measured length of pellets from the remaining pellets, if there be any, for moving the measured length of pellets from the work station onto a given one storage tray on the output station, and for moving the remaining pellets, if there be any, from the work station back onto the one given supply tray on the input station. More particularly, the pellet collating means includes means in the form of an input sweep head disposed adjacent the input station and being operable for sweeping pellets in the at least one row thereof onto the work station from the one pellet supply tray on the input station, means in the form of a gripping and measuring head disposed adjacent the work station and being operable for measuring the desired length of pellets and separating the measured length of pellets from the remaining pellets, if there be any, and means in the form of an output sweep head disposed adjacent the output station and being operable for sweeping the measured length of pellets from the work station onto the one pellet storage tray on the output station. One of the input and output sweep heads and of the gripping and measuring head is also operable to sweep the remaining pellets, if there be any, from the work station back onto the one pellet supply tray on the input station. Also, the present invention is directed to a method of collating nuclear fuel pellets which comprises the steps of: (a) supporting a plurality of pellet supply trays and a plurality of pellet storage trays at a tray positioning station, each of the supply trays supporting in at least one row thereon a plurality of nuclear fuel pellets of an enrichment different from the enrichments of pellets on at least some other of the supply trays, each of the storage tray being adapted to receive in at least one row thereon a plurality of nuclear fuel pellets of an enrichment different from the enrichments of pellets on at least some other of the storage trays; (b) transferring one supply tray from the tray positioning station and disposing the same on an input station of a pellet collating line; (c) transferring one storage tray from the tray positioning station and disposing the same on an output station of the pellet collating line; (d) sweeping pellets in the at least one row thereof from the one supply tray on the input station onto a work station of the pellet collating line located between the input and output stations thereof; (e) measuring a desired length of pellets in the at least one row thereof on the work station and separating the measured length of pellets from the remaining pellets, if there be any, in the row thereof; (f) sweeping the remaining pellets, if there be any, in the row thereof from the work station back onto the one supply tray on the input station; (g) transferring the one supply tray and remaining pellets, if there be any, thereon from the input station back to the tray support station; (h) sweeping the measured length of pellets from the work station onto the one storage tray on the output station; and (i) transferring the one storage tray and measured length of pellets thereon from the output station back to the tray positioning station. These and other advantages and attainments of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
abstract | There is provided a fluorescent X-ray analysis apparatus in which a detection lower limit has been improved by reducing an X-ray generating subsidiarily and detected. The fluorescent X-ray analysis apparatus is one which possesses an X-ray source irradiating a primary X-ray, and a detector in which a collimator having a through-hole in its center part has been placed in a front face, and in which, by the detector, there is detected a primary fluorescent X-ray which generates from a sample by irradiating the primary X-ray to a sample, and passes through the through-hole of the collimator. The X-ray source and the detector are disposed while adjoining the sample, and an irradiated face of the X-ray source or the detector, to which a primary scattered ray having generated by the fact that the primary X-ray scatters in the sample and the primary fluorescent X-ray having generated from the sample are irradiated, is covered by a secondary X-ray reduction layer reducing a secondary scattered ray and a secondary fluorescent X-ray, which generate by irradiations of the primary scattered ray and the primary fluorescent X-ray. |
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abstract | An internal control rod drive mechanism (CRDM) including an electric motor is disposed in a nuclear reactor and further includes a support surface with sealed electrical connectors electrically connected with the electric motor power the motor. The internal CRDM is disposed on a support element secured inside the nuclear reactor. The support element includes sealed electrical connectors mating with the sealed electrical connectors on the support surface of the internal CRDM to power the electric motor. The sealed electrical connectors may be sealed glass, ceramic, or glass-ceramic connectors welded onto the ends of the MI cables extending from the motor. Springs, are disposed between the mating sealed electrical connectors of the support element and the support surface. A purge line is integrated with each mated connection. |
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050769987 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The system shown in FIG. 1 includes a reactor 2 having a core which contains a neutron detector 4. Detector 4 responds to neutron activity in the core by producing a train of pulses at a rate representative of the level of neutron activity, which is proportional to the reactor power output. The pulse train from detector 4 is delivered to a counter 6 which counts the pulses during successive uniform time intervals T and delivers, at the end of each interval, a representation of the count occurring during that interval. The successive representations produced by counter 6 are supplied to a computing unit 8 which derives power level and rate of change indications. These indications may be supplied to a recorder 10, such as a strip chart, dial gauges 12 and 14, digital displays 16 and 18, and a circuit 20 provided to shut down reactor 2 if an excessive count rate is detected. According to the invention, the count rate signal which provides an indication of the reactor power level is in the form of a first order lag, infinite impulse response digital filter. Specifically, referring to FIG. 2, the signal supplied by counter 6 at the end of each interval, i, is, as provided by function block 30: ##EQU1## where j=ni, and n is a positive integer Counter 6 may be constituted by an apparatus as disclosed in U.S. Pat. No. 4,670,891, which is, in effect, a counter having a variable window. In the apparatus disclosed in that patent, the value for CR is updated at the end of each interval i, but the value of CR is determined by the total number of counts appearing during the previous ni (=j) intervals. The power level indication produced by unit 8 is represented by: EQU CR.sub.if =CR.sub.(i-l)f +(CR.sub.i -CR.sub.(i-l)f).multidot.F, where: CR.sub.if is the count rate resulting from digital filtering; PA1 CR.sub.(i-1)f is the filtered count rate derived during the preceding time interval; and PA1 F is a selected filter factor that controls the response to changes in CR. CR.sub.if is derived by subtracting CR.sub.(i-l)f from CRi in function block 32, multiplying the difference by F in function block 34 and adding the resulting product to CR.sub.(i-1)f in function block 36. The output from block 36 represents CR.sub.if and this output is delayed by T in function block 38 to provide a new value for CR.sub.(i-1)f. The time T corresponds to the duration of interval i. The filter factor F is related to the time constant .tau. of the filter as follows: ##EQU2## Generally, T may be of the order of 0.1 sec. but could have a value of between 0.025 and 0.25 sec., and .tau. may be fixed or variable. Preferably, .tau. is variable and has an inverse relationship to the present count rate. In this way, it is possible to provide a rapid response at high count rates and to filter out higher noise levels occurring at low count rates. Thus, .tau. may have the following form: ##EQU3## .tau. preferably varies between 0.2 and 8 seconds. The value for constant K is selected to establish the desired relation between .tau. and CR. In FIG. 2, F is derived in function block 40, based on CR.sub.if, CR.sub.(i-1)f and separately inputted values for T and K. In further accordance with the invention, a power rate of change indication is derived from the filtered count rate values, CR.sub.f, which indication is particularly valuable during reactor startup. Since the rate of change can vary during startup over a large range, the rate indication is based on the log of the CR.sub.f values. First, at the end of each time interval, an initial rate value, known as a startup rate value, SUR.sub.i, is calculated as follows: ##EQU4## The factor of 60 produces a value in units of decades/minute. There is then derived a filtered startup rate procedures similar to that described above. Specifically, the first filter procedure may be EQU SUR.sub.if =SUR.sub.(i-1)f +(SUR.sub.i -SUR.sub.(i-1)f).multidot.F, where F has the value described above. The next filtering would use the same relation, substituting SUR.sub.if for SUR.sub.i. Preferably, two such filterings are employed to account for the inherently noisy nature of SUR.sub.i. One filtering may be effected with a fixed time constant, the other with a variable time constant having the same value as that employed to obtain the power level indication. These operations are performed in function block 42 of FIG. 2, which may contain, for each filtering, a set of blocks corresponding to blocks 32-38. While the description above refers to particular embodiments of the present invention, it will be understood that many modifications may be made without departing from the spirit thereof. The accompanying claims are intended to cover such modifications as would fall within the true scope and spirit of the present invention. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive, the scope of the invention being indicated by the appended claims, rather than the foregoing description, and all changes which come within the meaning and range of equivalency of the claims are therefore intended to be embraced therein. |
abstract | A collimator module is disclosed for the modular assembly of a collimator for a radiation detector with a multiplicity of absorber elements, which are arranged one behind the other in a collimation direction and held by a carrier. In at least one embodiment, the carrier has at least one alignment device for aligning the collimator module in the collimation direction, which alignment device(s) interact with positioning device(s) in a detector mechanism of the radiation detector when they are integrated into the radiation detector. This provides the preconditions for integrating the collimator module in a fashion that is decoupled from a radiation convertor, and so this allows easy assembly of a collimator and adjustment to a position assumed between a radiation convertor and the collimator. Moreover, a radiation detector with such a collimator module is disclosed. |
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claims | 1. A projection objective configured so that during use the projection objective can direct light from an object plane to an image plane, the projection objective comprising:a last optical element on an image side of the objective, wherein the last optical element is plane on the image side and confines, together with the image plane, an immersion space which is configured to be filled with an immersion liquid, andat least one liquid or solid volume, whichhas plane-parallel interfaces,is configured to be introduced into a beam path of the projection objective andhas an optical thickness that is substantially equal to the optical thickness of the immersion space when the immersion space is filled with the immersion liquid,wherein the projection objective is configured to be used in a microlithographic projection exposure apparatus. 2. The objective of claim 1, wherein only refractive surfaces being plane and extending parallel to the image plane are arranged in a beam path between the image plane and the at least one liquid or solid volume which is furthest away from the image plane. 3. The objective of claim 2, wherein the last optical element on the image side is a plane-parallel terminating plate. 4. The objective of claim 1, wherein the at least one liquid or solid volume has a refractive index which is substantially equal to the refractive index of the immersion liquid, and wherein the overall extension of the at least one liquid or solid volume along an optical axis of the objective is substantially equal to the dimension of the immersion space along the optical axis. 5. The objective of claim 4, wherein the overall extension of the at least one liquid or solid volume along the optical axis differs by less than 10 nm from the dimension of the immersion space along the optical axis. 6. The objective of claim 5, wherein the overall extension of the at least one liquid or solid volume along the optical axis differs by less than 1 nm from the dimension of the immersion space along the optical axis. 7. The objective of claim 6, wherein the overall extension of the at least one liquid or solid volume along the optical axis differs by less than 0.1 nm from the dimension of the immersion space along the optical axis. 8. The objective of claim 1, wherein the at least one liquid or solid volume is a liquid, and wherein the objective comprises a sealable intermediate space arranged between two optical elements having mutually facing plane-parallel interfaces, said intermediate space being configured to receive the liquid. 9. The objective of claim 8, wherein the last optical element on the image side is a plane-parallel terminating plate which forms one of the two optical elements. 10. The objective of claim 9, further comprising an immersion device which is configured to fill and empty the immersion space and to fill and empty the intermediate space with the same liquid. 11. The objective of claim 1, wherein the at least one liquid or solid volume is formed by a plane-parallel plate which has the same refractive index as the immersion liquid. 12. A microlithographic projection exposure apparatus comprising the objective of claim 1. 13. A method for converting a projection objective of a microlithographic projection exposure apparatus from dry operation to immersed operation, the method comprising:a) providing a projection objective comprising a last optical element on an image side of the objective, wherein the last optical element is plane on the image side and confines, together with the image plane, an immersion space which is configured to be filled with an immersion liquid;b) removing at least one liquid or solid volume from a location in the beam path of the objective that is different from the immersion space, wherein the at least one liquid or solid volume has plane-parallel interfaces, and has an optical thickness that is substantially equal to the optical thickness of the immersion space when the immersion space is filled with the immersion liquid; andc) filling the immersion space with an immersion liquid which has the same refractive index as the at least one volume removed in step b). 14. The method of claim 13, wherein the at least one liquid or solid volume is liquid, and wherein the liquid volume is removed in step b) and relocated into the immersion space. 15. The method of claim 13, wherein, after exposures have been carried out in immersed operation, the objective is converted back to dry operation by removing the liquid from the immersion space and re-introducing the at least one volume removed in step b) into the beam path of the objective. 16. A method for the microlithographic production of microstructured components, the method comprising:a) providing a support, on at least some of which a layer of a photosensitive material is applied;b) providing a reticle which contains structures to be projected;c) providing the objective of claim 1; andd) projecting at least a part of the reticle onto a region on the layer with the aid of the objective provided in step c). 17. The method of claim 16, wherein the method is used to provide a microstructured component. 18. A projection objective configured so that during use the projection objective can direct light from an object plane to an image plane, the projection objective compnsing:an optical axis,a space whichis configured to be filled with a liquid andis confined by two at least substantially plane-parallel surfaces extending perpendicularly to the optical axis, andat least one liquid or solid volume, whichhas plane-parallel interfaces,is configured to be introduced into a beam path of the projection objective andhas an optical thickness that is substantially equal to the optical thickness of the space when it is filled with the liquid,wherein the projection objective is configured to be used in a microlithographic projection exposure apparatus. 19. The objective of claim 18, wherein only refractive surfaces being plane and extending perpendicular to the optical axis are arranged between the space and the at least one liquid or solid volume which is furthest away from the space. 20. The objective of claim 18, wherein one of the surfaces confining the space coincides with the image plane, and wherein the liquid is an immersion liquid. 21. The objective of claim 20, wherein the at least one liquid or solid volume has a refractive index which is substantially equal to the refractive index of the immersion liquid, and wherein the overall extension of the at least one liquid or solid volume along the optical axis is substantially equal to the dimension of the space along the optical axis. 22. The objective of claim 21, wherein the overall extension of the at least one liquid or solid volume along the optical axis differs by less than 10 nm from the dimension of the space along the optical axis. 23. The objective of claim 22, wherein the overall extension of the at least one liquid or solid volume along the optical axis differs by less than 1 nm from the dimension of the space along the optical axis. 24. The objective of claim 23, wherein the overall extension of the at least one liquid or solid volume along the optical axis differs by less than 0.1 nm from the dimension of the space along the optical axis. 25. The objective of claim 18, wherein the at least one liquid or solid volume is a liquid, and wherein the objective comprises a sealable intermediate space arranged between two optical elements having mutually facing plane-parallel interfaces, said intermediate space being configured to receive the liquid. 26. The objective of claim 25, wherein the space is an immersion space which is contiguous to the image plane, and wherein the objective comprises a plane-parallel terminating plate which is the last optical element on the image side and which separates the intermediate space from the immersion space. 27. The objective of claim 26, further comprising an immersion device which is configured to fill and empty the immersion space and to fill and empty the intermediate space with the same liquid. 28. The objective of claim 18, wherein the at least one liquid or solid volume is formed by a plane-parallel plate which has the same refractive index as the liquid. 29. A microlithographic projection exposure apparatus comprising the objective of claim 18. 30. A method for operating a projection objective of a microlithographic projection exposure apparatus, the method comprising:a) providing a projection objective comprising an optical axis and an empty space which is configured to be filled with a liquid and is confined by two at least substantially plane- parallel surfaces extending perpendicularly to the optical axis;b) carrying out exposures;c) removing at least one liquid or solid volume from a location in the beam path of the objective that is different from the space, wherein the at least one liquid or solid volume has plane-parallel interfaces, and has an optical thickness that is substantially equal to the optical thickness of the space when the space is filled with the liquid; andd) filling the space with a liquid. 31. The method of claim 30, wherein the at least one liquid or solid volume is liquid, and wherein the liquid volume is removed in step c) and relocated into the space in step d). 32. A method for operating a projection objective of a microlithographic projection exposure apparatus, the method comprising:a) providing a projection objective comprising an optical axis and a space which is filled with a liquid and is confined by two at least substantially plane-parallel surfaces extending perpendicularly to the optical axis;b) carrying out exposures;c) removing the liquid from the space andd) introducing at least one liquid or solid volume into a location in the beam path of the objective that is different from the space, wherein the at least one liquid or solid volume has plane-parallel interfaces, and has an optical thickness that is substantially equal to the optical thickness of the space when the space is filled with the liquid. 33. The method of claim 32, wherein the at least one liquid or solid volume is liquid, and wherein the liquid is removed in step c) and introduced into the space in step d). |
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054065992 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 8 illustrates a typical nuclear fuel assembly 11 that is generally comprised of a plurality of fuel rods 12, grid assemblies or spacer grids 13, guide tubes 14, and end fittings 15. Spacer grids 13 provide support to fuel rods 12 while still allowing maximum surface area contact of fuel rods 12 with coolant flowing through the individual cells of the spacer grids. A typical spacer grid 13 is depicted in FIG. 9 and 9A. Strips 28 are arranged to define a plurality of square or rectangular cells that receive fuel rods 12 as seen in FIG. 8. The center cell is normally used in a fuel assembly to receive an instrumentation tube. A plurality of other cells are used to receive guide tubes 14. As seen in FIG. 9A, strips 28 contain end tabs 27 and weld tabs 31. There are several different types of fuel rod contact points within a spacer grid, such as hardstops, softstops, and guide tube saddles. The spacer grid 13 can be described as a square array of grid cells comprised of combinations of these features. The configurations form three different type of cells, those being the fuel rod cell, guide tube cell, and instrument tube cell. While guide tube cells all have the same configurations and there is only one instrument tube configuration (similar to the guide tube configuration), the fuel rod cell has several different possible configurations. The different fuel rod configurations are: a) Inside normal cell (FIG. 1 and 1A)--There are six contact points, four outboard hardstops 1 at or near the top and bottom of the strip, and two softstops 2. FIG. 2 depicts this type cell fixtured by the invention. The outboard hardstops 1 are fixtured by the upper and lower slotted plates of the invention, and the softstops are positioned by the softstop alignment pin of the invention. PA1 b) Special double Hardstops (FIG. 3 and 3A)--There are six contact points, two outboard hardstops 1 on one wall at the top and bottom of the strip, two inboard hardstops 3 on another wall inboard near the top and bottom of the strip, and two softstops 2. FIG. 4 depicts this type cell fixtured by the invention. The outboard hardstop locations are positioned by the upper and lower slotted plates of the invention. The invention uses several types of alignment pins around the double hardstop locations. The normal softstop alignment pin of the invention is used in adjacent cells, and short alignment pins are used to fixture the inboard hardstops, and a special combination pin is used to simultaneously position the upper inboard hardstop and the softstop. PA1 c) Adjacent to Guide tubes (FIG. 5 and 5A)--There are six contact points, two outboard hardstops 1 on one wall at the top and bottom of the strip, two inboard hardstops 3 on the wall inboard of the guide tube saddles near the top and bottom of the strip, and two softstops 2. FIG. 6 depicts this type cell fixtured by the invention. The outboard hardstops 1 are positioned by the upper and lower slotted plates of the invention. In the guide tube cell, a sizing pin is used to set the cell size and contact points to allow proper insertion and fit-up of the guide tubes. In adjacent cells, normal softstop alignment pins are used in two cell locations, and short alignment pins and combination pins are used in the other two locations surrounding the guide tube cells. It is seen in FIG. 10 that the invention is generally indicated by the numeral 16. Weld fixture 16 is generally comprised of first plate 10, second plate 9, guide rods 19, guide cell pins 5, and alignment pins 6, 7, 8. First and second plates 10 and 9 are each provided with a plurality of intermittent intersecting slots 20 on one side of each plate, as shown in FIGS. 10, 11, and 12, that precisely align and position the fuel rod contact points 1 referred to as outboard hardstops. The slots 20 preferably intersect at right angles to define a plurality of raised polygons 21. As shown in FIG. 7, the intermittent slots are not equally spaced on either side because of the double hardstop strip. In addition, the spacing varies in one direction versus the other to ensure alignment of the stops by accounting for different strip shrinkage on the slotted versus solid side of the spacer grid egg crate. Therefore, D.sub.1 .noteq.D.sub.2 .noteq.D.sub.3 .noteq.D.sub.4 .noteq.D.sub.5 .noteq.D.sub.6. As best seen in FIG. 10, slots 20 preferably have a chamfered upper edge 22. For ease of illustration, chamfered upper edge 22 is only shown on second plate 9. In the preferred embodiment, the chamfer angle is thirty-five degrees to provide a 0.061 inch distance between the inner and outer most portion of the chamfer. Plates 10 and 9 each have a bore 23 therethrough in the polygon defined at each corner. Plates 10 and 9 are also provided with a bore 24 therethrough substantially at the intersections of slots 20. As best seen in FIG. 13, bores 24 are preferably provided with chamfered edge 25 on the sides of plates 10 and 9 opposite slots 20. As indicated by dimensions A and B respectively, chamfered edges 25 have a preferred diameter of 0.500 inch and a chamfer angle of sixty degrees. As seen in FIGS. 11 and 12, plates 10 and 9 are provided with bores 26 therethrough adjacent the outer edges of plates 10 and 9 and centered in slots 20. Bores 26 are drilled in the plates on the opposite side of slots 20. Bores 26 receive end tabs 27 on inner strips 28 of the spacer grid assembly, seen in FIG. 9A. Plates 10 and 9 are respectively provided with bores 29a and 29b therethrough in all recessed milled areas 30 referred to as guide tube cell locations. Bores 29b are sized to receive bolts or screws through plate 9 while bores 29a are threaded for receiving the threaded end of the bolts or screws and holding the plates together after the strips 28 that form the spacer grid are positioned between the plates 10 and 9. The recessed milled areas 30 that are provided with bores 29a,b correspond to component guide tube and instrument tube locations in a spacer grid assembly of a nuclear fuel bundle assembly. The milled recessed areas 30 are cut approximately to the depth of the slots 20. Guide cell pins 5 are sized to be received in the recessed milled areas 30 that have bores 29a,b therein. A selected number of guide cell pins 5, five in the preferred embodiment, are hollow as indicated in FIG. 10. The hollow pins receive the through bolts and are positioned to evenly distribute the pressure when plates 10 and 9 are bolted together on the assembled strips 28. In the preferred embodiment, guide cell pins 5 are slightly taller than strips 28 so that when plates 10 and 9 are bolted together the load is carried by the guide cell pins 5 and not strips 28. The end tabs 27 are received in bores 26 and weld tabs 31 extend into bores 24 without any distortion load being placed on the strips. Guide rods 19 are sized to fit closely through corner bores 23 in plates 10 and 9. This aligns slots 20 on the plates, which causes strips 28 to be forced into proper alignment as second plate 9 is moved towards first plate 10 on guide rods 19. Base 32 is provided on one end of guide rods 19 to serve as a rest for first plate 10 during fixturing of strips 28 into plates 10 and 9. Second plate 9 is provided with counterbores 33a,b in polygons not having bores 23 or 29b. Bores 33a,b, seen in section in FIG. 14 and 15, are sized to receive flat head screws not shown which are threaded into alignment pins 6, 7, and 8 seen in FIG. 2, 4, and 6. Alignment pins 6 are tapered inwardly away from the threaded end and are used to align and position fuel rod contact points referred to as softstops 2. As seen in FIG. 10, alignment pins 7 are shorter than pins 6 and are provided with an index pin 34 that extends radially from one end of the alignment pin. Pins 7 are sized for and positioned in polygons 21 in second plate 9 that correspond to those cells in spacer grid 13 that are used to align and position fuel rod contact points referred to as inboard hardstops 3. Since pins 7 do not extend the full height of spacer grid 13, corresponding pins 8 and bores 33b are provided for first plate 10. Pins 8 in second plate 9 are used for positioning in polygons with bores 33b that require simultaneous positioning of inboard hardstop and softstop contact points, and allow for corresponding pins 7 and bores 33b in first plate 10. As seen in FIG. 14 and 15, each bore 33b is different from bore 33a only in having radial notch 35 that receives index pin 34. The shape of short pin 7 and combination pin 8 are directly related to the position of the hardstops in the grid cells. Placing index pin 34 in notch 35 prevents turning of pins 7 and 8 during threading of the flat head screw and maintains the proper orientation of pins 7 and 8 throughout use. In the preferred embodiment an alignment pin is provided in each polygon not occupied by a guide cell pin in second plate 9, and in each polygon in first plate 10 with bore 33b. In operation, weld fixture 16 is designed to align and position all the critical fuel rod contact points within a spacer grid during the intersection welding process and not to position any of the actual base strip locations. First plate 10 has a guide rod 19 placed through each corner bore 23 such that slots 20 face upward. Plates 10 and 9 mirror each other with regard to slot spacing. However, the spacing between slots 20 is not equal on either side of the plates. This requires a means of insuring that plates 10 and 9 are properly oriented relative to each other to achieve the proper slot orientation between plates. This is achieved by having one corresponding bore 23 in the corner of each plate larger than the remaining bores 23 and the diameter of one of guide rods 19 of a corresponding larger diameter. This insures proper orientation of plates 10 and 9 relative to each other. The spacer grid egg crate, which is a plurality of crisscrossing strips 28, is placed on first plate 10 so that each strip 28 is received in and resting on the bottom of an individual slot 20 and end tabs 27 and weld tabs 31 are respectively received in edge bores 26 and slot intersection bores 24. This assures that all strip contact points 1, 2, 3, and 4 are at the same elevations (relative to each other) and are secure in first plate 10. As seen in FIG. 2, 4, and 6, outboard hardstops 1 at the upper edges of strips 28 are received in slots 20 for alignment. Sizing of the slots 20 to receive the outboard hardstops 1 provides for the proper alignment and positioning of the aforementioned fuel rod contact points. Guide cell pins 5 are then inserted into the locations that correspond to guide tube and instrument tube locations in the nuclear fuel assembly to assure that the saddles in strips 28 are properly oriented for welding. The hollow guide cell pins are preferably positioned as shown in FIG. 10 at the corresponding instrument tube location, center cell, and at four outer locations having bores 29a. Second plate 9 is then received on guide rods 19 such that slots 20 face first plate 10 and is slowly lowered onto the egg crate. Second plate 9 bottoms out on guide cell pins 5 without transferring a load to the egg crate while still receiving the strips and tabs in their corresponding slots as referred to above. During the lowering of second plate 9, alignment pins 6, 7, and 8 each move into an individual cell defined by the egg crate to properly align and position all of the fuel rod contact points not positioned by the features in the first plate 10. It should be understood that pins 6, 7, and 8 will not necessarily contact all of the hardstops or softstops in the spacer grid. Contact in some areas of the spacer grid results in proper alignment in other areas, with all of the pins being required to insure proper alignment and positioning throughout the spacer grid. Bolts are then inserted through plate 9 and hollow pins 5 at selected locations and threaded into bores 29b in first plate 10 to secure the plates and egg crate in position. Guide rods 19 are then removed and weld fixture 16 is placed in a welding frame for welding of the interior intersections of the grid strips. In the welding process a laser beam is directed into each slot intersection bore 24 through chamfered portion 25 onto weld tabs 31 which melt and secure strips 28 together to form the interior section of a spacer grid. Because many varying and differing embodiments may be made within the scope of the inventive concept herein taught and because many modifications may be made in the embodiment herein detailed in accordance with the descriptive requirement of the law, it is to be understood that the details herein are to be interpreted as illustrative and not in a limiting sense. |
description | The field of the disclosure relates generally to radionuclide generators and, more particularly, to systems and methods for an autoclave rack loading system. Radioisotopes used for medical diagnostic purposes may emit high levels of radioactivity. These radioisotopes are typically generated in generators contained within hot cells that prevent the radioactivity from escaping the generator. However, the hot cell prevents operators from accessing the generation process. Accordingly, equipment within the hot cell, such as autoclave sterilizers, must be loaded by automated equipment within the hot cell. Conventional autoclaves include only one sterilization station or cart containing column assemblies. The sterilization process can take a significant amount of time such that he sterilization process may be the rate limiting step in the generation process. More sterilization stations within an autoclave increases the amount of column assemblies which can be sterilized in an autoclave. Because space is limited in autoclaves, more sterilization stations may be added to the autoclave by stacking the stations on top of each other. Raising carts of column assemblies to the stacked sterilization stations may be a challenge. Accordingly, a need exists for reliable automated systems and methods for loading carts of column assemblies into autoclaves with stacked sterilization stations. This Background section is intended to introduce the reader to various aspects of art that may be related to various aspects of the present disclosure, which are described and/or claimed below. This discussion is believed to be helpful in providing the reader with background information to facilitate a better understanding of the various aspects of the present disclosure. Accordingly, it should be understood that these statements are to be read in this light, and not as admissions of prior art. One aspect is a system for manufacturing radionuclide generators including an enclosure defining a radioactive environment, at least one autoclave sterilizer within the enclosure, and at least two loading and unloading elevators. The enclosure includes radiation shielding to prevent radiation within the radioactive environment from moving to an exterior of the enclosure. Each autoclave sterilizer includes a plurality of sterilization stations arranged vertically and at least two autoclave rails. One loading and unloading elevator is configured to load a cart into the autoclave sterilizer and one loading and unloading elevator is configured to unload the cart from the autoclave sterilizer. Each loading and unloading elevator includes at least two cart rails configured to support the cart and a plurality of loading elevator rails coupled to the cart rails. The loading elevator rails are configured to adjust the height of the cart rails. In another aspect, a loading and unloading elevator for an autoclave sterilizer in a radioactive environment including a table top, an elevation system, at least two cart rails, and a plurality of loading elevator rails. The table top separates a processing space from a maintenance space. The elevation system is positioned within the maintenance space. The cart rails are configured to support a cart. The loading elevator rails are coupled to the cart rails. The cart rails are positioned within the processing space. The loading elevator rails extend from the elevation system within the maintenance space through the table top to the processing space. The loading elevator rails are configured to adjust the height of the cart rails. In yet another aspect, a method includes transferring a cart to a first loading and unloading elevator within a radioactive environment. The cart is configured to hold a plurality of column assembly racks. Each column assembly rack is configured to hold a plurality of column assemblies. The first loading and unloading elevator includes at least two cart rails, a motor, and a plurality of loading elevator rails coupled to the cart rails. The cart rails are configured to support the cart. The loading elevator rails are configured to adjust the height of the cart rails. The method also includes adjusting the height of the cart with the motor to a predetermined height corresponding to the height of a sterilization station within an autoclave. The method further includes transferring the cart to the sterilization station. Various refinements exist of the features noted in relation to the above-mentioned aspects. Further features may also be incorporated in the above-mentioned aspects as well. These refinements and additional features may exist individually or in any combination. For instance, various features discussed below in relation to any of the illustrated embodiments may be incorporated into any of the above-described aspects, alone or in any combination. Corresponding reference characters indicate corresponding parts throughout the several views of the drawings. Radioactive material is used in nuclear medicine for diagnostic and therapeutic purposes by injecting a patient with a small dose of the radioactive material, which concentrates in certain organs or regions of the patient. Radioactive materials typically used for nuclear medicine include Technetium-99m (“Tc-99m”), Indium-111m (“In-111”), Thallium-201, and Strontium-87m, among others. Such radioactive materials may be produced using a radionuclide generator. Radionuclide generators generally include an autoclave for terminally sterilizing column assemblies. The autoclave may be located in a hot cell to shield the surrounding environment from radiation. As such, space within the hot cell and the autoclave is limited. In order to maximize the space within the autoclave, racks of column assemblies are stacked vertically within the autoclave. A loading and unloading system loads the racks into the autoclave and unloads the racks from the autoclave after sterilization. FIG. 1 is a schematic view of a system 100 for manufacturing radionuclide generators. The system 100 shown in FIG. 1 may be used to produce various radionuclide generators, including, for example and without limitation, Technetium generators, Indium generators, and Strontium generators. The system 100 of FIG. 1 is particularly suited for producing Technetium generators. A Technetium generator is a pharmaceutical drug and device used to create sterile injectable solutions containing Tc-99m, an agent used in diagnostic imaging with a relatively short 6 hour radiological half-life, allowing the Tc-99m to be relatively quickly eliminated from human tissue. Tc-99m is “generated” via the natural decay of Molybdenum (“Mo-99”), which has a 66 hour half-life, which is desirable because it gives the generator a relatively long two week shelf life. During generator operation (i.e., elution with a saline solution), Mo-99 remains chemically bound to a core alumina bed (i.e., a retaining media) packed within the generator column, while Tc-99m washes free into an elution vial, ready for injection into a patient. While the system 100 is described herein with reference to Technetium generators, it is understood that the system 100 may be used to produce radionuclide generators other than Technetium generators. As shown in FIG. 1, the system 100 generally includes a plurality of stations. In the example embodiment, the system 100 includes a cask loading station 102, a formulation station 104, an activation station 106, a fill/wash station 108, an autoclave loading station 110, an autoclave station 112, an autoclave unloading station 114, a quality control testing station 116, a shielding station 118, and a packaging station 120. The cask loading station 102 is configured to receive and handle casks or containers of radioactive material, such as a parent radionuclide, and transfer the radioactive material to the formulation station 104. Radioactive material may be transported in secondary containment vessels and flasks that need to be removed from an outer cask prior to formulation. The cask loading station 102 includes suitable tooling and mechanisms to extract secondary containment vessels and flasks from outer casks, as well as transfer of flasks to the formulation cell. Suitable devices that may be used in the cask loading station include, for example and without limitation, telemanipulators. At the formulation station 104, the raw radioactive material (i.e., Mo-99) is quality control tested, chemically treated if necessary, and then pH adjusted while diluting the raw radioactive material to a desired final target concentration. The formulated radioactive material is stored in a suitable containment vessel (e.g., within the formulation station 104). Column assemblies containing a column of retaining media (e.g., alumina) are activated at the activation station 106 to facilitate binding of the formulated radioactive material with the retaining media. In some embodiments, column assemblies are activated by eluting the column assemblies with a suitable volume of HCl at a suitable pH level. Column assemblies are held for a minimum wait time prior to charging the column assemblies with the parent radionuclide. Following activation, column assemblies are loaded into the fill/wash station 108 using a suitable transfer mechanism (e.g., transfer drawer). Each column assembly is then charged with parent radionuclide by eluting formulated radioactive solution (e.g., Mo-99) from the formulation station 104 through individual column assemblies using suitable liquid handling systems (e.g., pumps, valves, etc.). The volume of formulated radioactive solution eluted through each column assembly is based on the desired Ci activity for the corresponding column assembly. The volume eluted through each column assembly is equivalent to the total Ci activity identified at the time of calibration for the column assembly. For example, if a volume of formulated Mo-99 required to make a 1.0Ci generator (at time of calibration) is ‘X’, the volume required to make a 19.0Ci generator is simply 19 times X. After a minimum wait time, the charged column assemblies are eluted with a suitable volume and concentration of acetic acid, followed by an elution with a suitable volume and concentration of saline to “wash” the column assemblies. Column assemblies are held for a minimum wait time before performing assays on the column assemblies. The charged and washed column assemblies (or radionuclide generators) are then transferred to the autoclave load station 110, in which assays are taken from each column assembly to check the amount of parent and daughter radionuclide produced during elution. Each column assembly is eluted with a suitable volume of saline, and the resulting solution is assayed to check the parent and daughter radionuclide levels in the assay. Where the radioactive material is Mo-99, the elutions are assayed for both Tc-99m and Mo-99. Column assemblies having a daughter radionuclide (e.g., Tc-99m) assay falling outside an acceptable range calculation are rejected. Column assemblies having a parent radionuclide (e.g., Mo-99) breakthrough exceeding a maximum acceptable limit are also rejected. As described further herein, systems and methods of the present disclosure facilitate assaying elutions of radionuclide generators without the use of transfer vials or other liquid containers that require transfer to a radiation detection device. For example, embodiments of the systems and methods described herein facilitate eluting a radionuclide generator directly into the collection reservoir of a radiation detection device. Following the assay process, tip caps are applied to the outlet port and the fill port of the column assembly. Column assemblies may be provided with tip caps already applied to the inlet port. If the column assembly is not provided with a tip cap pre-applied to the inlet port, a tip cap may be applied prior to, subsequent to, or concurrently with tip caps being applied to the outlet port and the fill port. Assayed, tip-capped column assemblies are then loaded into an autoclave sterilizer located in the autoclave station 112 for terminal sterilization. The sealed column assemblies are subjected to an autoclave sterilization process within the autoclave station 112 to produce terminally-sterilized column assemblies. Following the autoclave sterilization cycle, column assemblies are unloaded from the autoclave station 112 into the autoclave unloading station 114. Column assemblies are then transferred to the shielding station 118 for shielding. Some of the column assemblies are transferred to the quality control testing station 116 for quality control. In the example embodiment, the quality control testing station 116 includes a QC testing isolator that is sanitized prior to QC testing, and maintained at a positive pressure and a Grade A clean room environment to minimize possible sources of contamination. Column assemblies are aseptically eluted for in-process QC sampling, and subjected to sterility testing within the isolator of the quality control testing station 116. Tip caps are reapplied to the inlet and outlet needles of the column assemblies before the column assemblies are transferred back to the autoclave unloading station 114. The system 100 includes a suitable transfer mechanism for transferring column assemblies from the autoclave unloading station 114 (which is maintained at a negative pressure differential, Grade B clean room environment) to the isolator of the quality control testing station 116. In some embodiments, column assemblies subjected to quality control testing may be transferred from the quality control testing station 116 back to the autoclave unloading station 114, and can be re-sterilized and re-tested, or re-sterilized and packaged for shipment. In other embodiments, column assemblies are discarded after being subjected to QC testing. In the shielding station 118, column assemblies from the autoclave unloading station 114 are visually inspected for container closure part presence, and then placed within a radiation shielding container (e.g., a lead plug). The radiation shielding container is inserted into an appropriate safe constructed of suitable radiation shielding material (e.g., lead, tungsten or depleted uranium). Shielded column assemblies are then released from the shielding station 118. In the packaging station 120, shielded column assemblies from the shielding station 118 are placed in buckets pre-labeled with appropriate regulatory (e.g., FDA) labels. A label uniquely identifying each generator is also printed and applied to each bucket. A hood is then applied to each bucket. A handle is then applied to each hood. The system 100 may generally include any suitable transport systems and devices to facilitate transferring column assemblies between stations. In some embodiments, for example, each of the stations includes at least one telemanipulator to allow an operator outside the hot cell environment (i.e., within the surrounding room or lab) to manipulate and transfer column assemblies within the hot cell environment. Moreover, in some embodiments, the system 100 includes a conveyance system to automatically transport column assemblies between the stations and/or between substations within one or more of the stations (e.g., between a fill substation and a wash substation within the fill/wash station 108). In the example embodiment, some stations of the system 100 include and/or are enclosed within a shielded nuclear radiation containment chamber, also referred to herein as a “hot cell”. Hot cells generally include an enclosure constructed of nuclear radiation shielding material designed to shield the surrounding environment from nuclear radiation. Suitable shielding materials from which hot cells may be constructed include, for example and without limitation, lead, depleted uranium, and tungsten. In some embodiments, hot cells are constructed of steel-clad lead walls forming a cuboid or rectangular prism. In some embodiments, a hot cell may include a viewing window constructed of a transparent shielding material. Suitable materials from which viewing windows may be constructed include, for example and without limitation, lead glass. In the example embodiment, each of the cask loading station 102, the formulation station 104, the fill/wash station 108, the autoclave loading station 110, the autoclave station, the autoclave unloading station 114, and the shielding station 118 include and/or are enclosed within a hot cell. In some embodiments, one or more of the stations are maintained at a certain clean room grade (e.g., Grade B or Grade C). In the example embodiment, pre-autoclave hot cells (i.e., the cask loading station 102, the formulation station 104, the fill/wash station 108, the autoclave loading station 110) are maintained at a Grade C clean room environment, and the autoclave unloading cell or station 114 is maintained at a Grade B clean room environment. The shielding station 118 is maintained at a Grade C clean room environment. The packaging stations 120 are maintained at a Grade D clean room environment. Additionally, the pressure within one or more stations of the system 100 may be controlled at a negative or positive pressure differential relative to the surrounding environment and/or relative to adjacent cells or stations. In some embodiments, for example, all hot cells are maintained at a negative pressure relative to the surrounding environment. Moreover, in some embodiments, the isolator of the quality control testing station 116 is maintained at a positive pressure relative to the surrounding environment and/or relative to adjacent stations of the system 100 (e.g., relative to the autoclave unloading station 114). FIG. 2 is a perspective view of the autoclave loading station 110, the autoclave station 112, and the autoclave unloading station 114 of a radionuclide generator that may be produced with the system 10. FIG. 3 is a perspective view of the autoclave station 112. FIG. 4 is a perspective view of the autoclave loading station 110 and the autoclave unloading station 114. FIG. 5 is a perspective view of the autoclave loading station 110 and sterilization stations 222 with autoclave sterilizer 204 removed for clarity. Although not illustrated in FIGS. 2-5, the components of the autoclave loading station 110, the autoclave station 112, and the autoclave unloading station 114 are enclosed within a hot cell. That is, the components of the autoclave loading station 110, the autoclave station 112, and the autoclave unloading station 114 are enclosed within an enclosure constructed of nuclear radiation shielding material designed to shield the surrounding environment from nuclear radiation. Additionally, in some embodiments, the autoclave loading station 110, the autoclave station 112, and the autoclave unloading station 114 is maintained at a Grade B or higher class clean room environment. That is, the autoclave unloading station 114 has a clean room classification of Grade B or higher. As shown in FIG. 2, the autoclave loading station 110 and the autoclave unloading station 114 both include an autoclave loading/unloading elevator 202. The autoclave loading/unloading elevator 202 within the autoclave loading station 110 is positioned on the upstream (i.e., loading) side of an autoclave sterilizer 204. The autoclave loading elevator 202 includes two elevator cart rails 206 that receive a cart 208 containing up to eight racks 210 (with up to eight column assemblies per rack 210) from the fill/wash station 108. The cart 208 includes a plurality of wheels 212 that enable the cart 208 to roll from the autoclave loading station 110 to the autoclave loading station 110 and the autoclave unloading station 114. The cart 208 may be moved from the autoclave unloading station 114 to the autoclave loading station 110 and the autoclave unloading station 114. The autoclave sterilizer 204 includes an outer casing 214, an entrance 216, an exit 218 and two doors 220. The autoclave sterilizer 204 also includes a plurality of sterilization stations 222 arranged vertically such that each sterilization station 222 is at a different height within the autoclave sterilizer 204. Each sterilization station 222 includes autoclave rails 224, each positioned within autoclave sterilizer 204. During the loading operation, the doors 220 open allowing the cart 208 to enter the autoclave sterilizer 204 through the entrance 216. The autoclave rails 224 receive the cart 208 from the autoclave loading station 110. Specifically, the cart 208 is rolled from the elevator cart rails 206 to the autoclave rails 224 using the autoclave loading/unloading mechanism. The doors 220 are closed and the sterilization process begins within the autoclave sterilizer 204. Autoclave sterilization generally includes exposing a column assembly, having a column loaded with parent radionuclide, to a saturated steam, or a steam-air mixture environment. In this embodiment, the autoclave sterilizer 204 includes three sterilization stations 222. However, the autoclave sterilizer 204 is not limited to three sterilization stations 222 and may include any number of sterilization stations 222 that enable the autoclave sterilizer to operate as described herein including, for example and without limitation, up to five sterilization stations 222. Each sterilization station 222 supports one cart 208 and each cart 208 holds eight racks 210. Each rack 210 holds up to eight column assemblies. Thus, the autoclave sterilizer 204 sterilizes up to 192 column assemblies per sterilization process. In other embodiments, each cart can hold up to twenty-four racks 210 and each rack 210 can hold up to eight column assemblies. Thus, the autoclave sterilizer 204 sterilizes up to 1200 column assemblies per sterilization process. The autoclave loading/unloading elevator 202 within the autoclave unloading station 114 is positioned on the downstream (i.e., unloading) side of the autoclave sterilizer 204. The elevator cart rails 206 receive the cart 208 from the autoclave sterilizer 204. The cart 208 may be removed from the autoclave sterilizer 204, and the racks 210 transferred to an autoclave unloading shuttle (not shown) using an autoclave unloading mechanism including, for example and without limitation, manual, automated, or semi-automated transfer mechanisms such as telemanipulators and pneumatic cylinders. FIG. 6 is a perspective view of an underside of the autoclave loading/unloading elevator 202. FIG. 7 is a perspective view of a bellows sleeves. FIG. 8 is a perspective view of a bearing. As shown in FIG. 6, the autoclave loading/unloading elevator 202 includes a table top 226 that separates a maintenance space 228 below the table top 226 from a processing space 230 above the table top 226. The table top 226 includes a plurality of penetrations 232 that enable a plurality of elevator guide rails 234 to extend through the table top 226. The elevator guide rails 234 are coupled to the elevator cart rails 206 and are configured to raise and lower the elevator cart rails 206. In this embodiment, the autoclave loading/unloading elevator 202 includes four elevator guide rails 234. However, the autoclave loading/unloading elevator 202 may include any number of elevator guide rails 234 that enable the autoclave loading/unloading elevator 202 to operate as described herein. The autoclave loading/unloading elevator 202 also includes an elevation system 236 positioned below table top 226 within maintenance space 228. The elevation system 236 is configured to raise and lower the elevator guide rails 234 which are configured to raise and lower the elevator cart rails 206. The elevation system 236 includes a motor 238, a center worm drive 240, two right-angle gear boxes 242, two screws 244, and a platform 246. The elevation system 236 also includes a programmable logic controller (hereinafter “PLC”) system (not shown) configured to control the elevation system 236. In this embodiment, the motor 238 includes a servomotor. Servomotors use resolver feedback and polyurethane cabling, and are immune to high radiation effects. However, the motor 238 can include any motor that enables the elevation system 236 to operate as described herein. In this embodiment, the motor 238 is coupled to the center worm drive 240 that is coupled to the two right-angle gear boxes 242 by a pair of shafts 248. The two right-angle gear boxes 242 are each coupled to a respective screw 244 which each extend downward from the two right-angle gear boxes 242 to the platform 246. The screws 244 each include a raised helical thread (not shown) running around the screw 244 and extending along a length of the screw 244. The platform 246 includes two threaded holes 250 each including a depressed helical thread (not shown) running around the inside of threaded holes 250. The elevator guide rails 234 extend vertically from the platform 246 through the table top 226 to the elevator cart rails 206. During operation, the motor 238 provides the power for elevating the cart 208. The center worm drive 240 translates the motion of the motor 238 through the shafts 248 to the two right-angle gear boxes 242. The two right-angle gear boxes 242 each rotate a respective screw 244. Rotation of the screw 244 causes the complementary helical treading on the screw 244 and in the threaded holes 250 to raise or low the platform 246. The platform 246 raises or lowers the elevator guide rails 234 which raise or lower the elevator cart rails 206 and the cart 208. In one embodiment, motor 238 raises the cart 208 at a speed of about 0.1-1 inch per second, or about 0.2-0.5 inches per second, or about 0.25 inches per second. As shown in FIG. 5, the sterilization stations 222 are arranged vertically within the autoclave sterilizer 204. Each sterilization station 222 has a height that is different from the other sterilization stations 222. As such, the elevation system 236 raises or lowers the cart 208 to a predetermined height corresponding to each sterilization station 222. Only when the elevation system 236 raised or lowered the cart 208 to the predetermined height corresponding to each sterilization station 222 can the cart 208 be rolled on to the autoclave rails 224. The PLC system controls the elevation system 236 such that the elevation system 236 only raises or lowers the cart 208 to the predetermined height corresponding to each sterilization station 222. Additionally, the motor 238 includes a brake (not shown) that prevents the cart 208 from descending if power is no longer supplied to the motor 238. In this embodiment, the lowest sterilization station 222 is approximately 12 inches above table top 226. Each sterilization station 222 above the lowest sterilization station 222 is approximately 8 inches above the previous sterilization station 222 such that there is at least about 0.1-0.5, or about 0.2-0.4 or about 0.25 inch clearance between loaded carts 208 within the autoclave sterilizer 204. Referring still to FIG. 5, the autoclave loading/unloading elevator 202 includes an unloading prevention bar 252 configured to prevent an operator from unloading a cart 208 from the autoclave sterilizer 204 before the elevation system 236 is prepared to receive the cart 208. In this embodiment, the unloading prevention bar 252 includes a bent bar with a wide portion 256 and a narrow portion 258. Each end 261 of the bent bar 252 is coupled to one of the elevator cart rails 206 and the bent bar 252 extends vertically from the elevator cart rails 206. The wide portion 256 is located near the elevator cart rails 206 and is wide enough to allow the cart 208 to pass from the elevator cart rails 206 to the autoclave sterilizer 204. The narrow end 258 is located away from the elevator cart rails 206 and is narrow enough to prevent a cart 208 in the autoclave sterilizer 204 from falling onto a cart 208 on the elevator cart rails 206 when the elevation system 236 is not at the predetermined height corresponding to the sterilization station 222. In some embodiments, the wide end 256 is about 6-10 inches, or about 8 inches tall, and the narrow end 258 is about 15-25 inches tall or about 20 inches tall. The wide end 256 is between about 15-25 inches wide or about 18 inches wide, and the narrow end 258 is about 8-10 inches wide or about 9 inches wide. As shown in FIG. 4, autoclave loading/unloading elevator 202 includes a plurality of bellow sleeves 260 and a plurality of bearings 262. The bellow sleeves 260 surround the elevator guide rails 234 above the table top 226 within the processing space 230. The bellow sleeves 260 extend with the elevator guide rails 234 in the vertical direction as the elevation system 236 raises and lowers the cart 208. The bellow sleeves 260 seal the penetrations 232 and ensure the processing space 230 is separated from the maintenance space 228. This ensures that the processing space 230 can be sanitized with Vaporous Hydrogen Peroxide (VHP) without dirty air leakage from the maintenance space 228 below the table top 226, and without exposure of lubricated elevator shafts within the processing space 230. Bellow sleeves 260 are made of Viton, ethylene propylene diene terpolymer rubber (EPDM), or polyvinyl chloride (PVC). However, the bellow sleeves 260 can be made of any material that enables the elevation system 236 to operate as described herein. The elevator guide rails 234 extend through the table top 226 through the bearings 262. The bearings 262 ensure the elevator guide rails 234 move smoothly up and down, and that the elevator guide rails 234 will not mechanically bind during movement. FIG. 9 is a perspective view of two autoclave loading stations 110, two autoclave stations 112, and two autoclave unloading stations 114. In this embodiment, system 100 includes two autoclave loading stations 110, two autoclave stations 112, and two autoclave unloading stations 114 for redundancy. An example system suitable for carrying out methods of this disclosure includes an autoclave station 112 including at least one autoclave sterilizer 204, an autoclave loading station 110 adjoining the autoclave station 112, and an autoclave unloading station 114 adjoining the autoclave station 112. In some embodiments, a hot cell encloses the autoclave loading station 110, the autoclave station 112, and the autoclave unloading station 114. Additionally, in some embodiments, autoclave loading station 110 and the autoclave unloading station 114 include an elevation system 236 configured to raise and lower a cart 208. Moreover, in some embodiments, the autoclave loading station 110 and the autoclave unloading station 114 include telemanipulators configured to load and unload the carts 208 from the autoclave sterilizer 204. The systems and methods of the present disclosure provide several advantages over known autoclave sterilizer loading and unloading systems. For example, embodiments of the disclosed systems and methods facilitate raising and lowering racks of column assemblies. Additionally, embodiments of the disclosed systems and methods facilitate loading/unloading the racks into and from an autoclave safely, given limited mechanical reach within the hot cell environment. Embodiments of the present disclosure include loading/unloading elevators which raise and lower racks of column assemblies, rails which facilitate rolling carts of column assemblies into the autoclave, a motor to power the loading/unloading elevators, and PLC to control the loading/unloading elevators. The loading/unloading elevators enable the autoclave sterilizer to include additional stacked sterilization stations. The additional sterilization stations allow the autoclave sterilizer to sterilize more column assemblies, improving the throughput of the generator manufacturing process. The elevator design allows completely sealed separation between a clean space above the tabletop that can be sanitized for pharmaceutical manufacturing, and a mechanical space below the tabletop. When introducing elements of the present invention or the embodiment(s) thereof, the articles “a”, “an”, “the” and “said” are intended to mean that there are one or more of the elements. The terms “comprising”, “including” and “having” are intended to be inclusive and mean that there may be additional elements other than the listed elements. As various changes could be made in the above constructions and methods without departing from the scope of the invention, it is intended that all matter contained in the above description and shown in the accompanying drawings shall be interpreted as illustrative and not in a limiting sense. |
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061119288 | claims | 1. A top mount canopy seal mechanical clamp assembly for repair of a leaking canopy seal weld located on a top surface of a nozzle, comprising: an annular hollow housing having a radially inwardly directed flange of such dimension as to permit it to telescopically pass a nozzle flange having said canopy seal weld; insert support halves of semi-annular shape concentric within said annular hollow housing and having an annular shoulder for face-to-face engagement with an axially facing inner surface of said annular hollow housing flange; an annular hollow top plate including an annular slot; said top plate in register concentrically with said annular hollow housing and surrounding said nozzle; a seal seat in said top plate positioned over and radially outward of said canopy seal weld in which a flexible graphite annulus is mounted for engagement with said canopy seal weld; an annular seal support ring positioned within said annular slot and arranged to completely surround said flexible graphite annulus to confine said flexible graphite annulus over said canopy seal weld; cap screws drawing said top plate and said housing together; whereby the clamp assembly introduces a compressive load into said flexible graphite annulus and against said canopy seal weld to create a flexible graphite leak stopping seal at the weld. an annular hollow housing having a radially inwardly directed flange of such dimension as to permit it to telescopically pass a flange of said nozzle having the canopy seal weld; insert support halves of semi-annular shape concentric within said annular hollow housing and having an annular shoulder in face-to-face engagement with an axially facing inner surface of said annular hollow housing flange, said insert support halves each having an upper surface engaging a lower surface of the nozzle; an annular hollow top plate in register concentrically with said annular hollow housing and surrounding the nozzle, said top plate including an annular slot therein; a seal seat in said top plate positioned over and radially outward of said canopy seal weld in which a flexible graphite annulus is mounted for engagement with said canopy seal weld; an annular seal support ring positioned within said annular slot and arranged to completely surround said flexible graphite annulus to confine said flexible graphite annulus over said canopy seal weld; and cap screws drawing said top plate and said housing together and placing a compressive load into said flexible graphite annulus and against said canopy seal weld to create a flexible graphite leak stopping seal at the weld. 2. The clamp assembly of claim 1, further comprising Belleville washers positioned between respective heads of said cap screws and said top plate to provide and maintain a necessary retaining loading force. 3. The clamp assembly according to claim 1, wherein said insert support halves each have an upper surface shaped to correspond with a lower surface of said nozzle to engage said lower surface of said nozzle directly. 4. In combination, a nozzle having a canopy seal weld located on a top surface thereof and a canopy seal mechanical clamp assembly for repairing a leak in said canopy seal weld, comprising: 5. The combination of claim 4, further comprising Belleville washers positioned between respective heads of said cap screws and said top plate to provide and maintain a necessary retaining loading force. |
045483477 | description | DETAILED DESCRIPTION OF THE INVENTION This invention is directed to an automated fuel pin loading system for nuclear reactors. It is directed specifically toward automated fuel pin production with maximized protection against spread of contamination. FIG. 1 shows the general flow path for fuel pin fabrication. The boxes bounded by dashed lines represent components of the potentially contaminated automated fuel pin loading system which is the subject of this disclosure. A fuel pin subassembly as fabricated is shown in FIG. 2. The subassembly consists of a length of cladding 10 having a welded end cap 11, shown as the bottom end of the fuel pin, and internal non-contaminated fuel pin hardware (not shown) adjacent the welded end cap 11. A fuel loading funnel 12 is mounted to the remaining open end of the fuel cladding 10. As shown in FIG. 4, the fuel pin subassembly funnel 12 has an enlarged outer open end that leads to a reduced diameter neck. A conical transition section connects the open end of funnel 12 to its neck. The smaller neck is at least partially inserted into the open axial end of the cladding as a coaxial extension of it. The fit between the funnel neck and the cladding interior should be reasonably close, and the thickness of the funnel neck should be minimal. The funnel 12 is held within the length of cladding 10 by a continuous length of plastic shrink tubing 14. After placement overlapping a portion of both the funnel and cladding, the tubing 14 is shrunk diametrically by proper application of heat. It tightly encircles and grips both the cladding 10 and funnel 12. It maintains them as a unit during reception of fuel into the cladding through the funnel 12. The funnel 12 is designed to be discarded after the length of cladding 10 has been charged with fuel pellets. Removal of the funnel is accomplished by pulling tubing 14 as cladding 10 is retracted axially. Removal is facilitated by providing a slidable collar or ring 13 that surrounds cladding 10 beneath the shrink tubing 14. The coaxial ring 13 presents a rear annular shoulder which can be engaged to pull ring 13, tubing 14 and funnel 12 as a unit. If desired, the shoulder could be presented as part of that portion of funnel 12 enveloped under the shrink tubing 14. The tubing 14 preferably extends along the outer surface of cladding 10 as an outer protective element. It provides a cylindrical surface for engagement by resilient or inflatable seals. Particles embedded in the tubing 14 as a result of such sealing will be discarded with the tubing when it is removed. This minimizes contamination of the cladding itself. Referring to the flow path shown in FIG. 1 and the associated schematic plan view shown in FIG. 3, the general steps of the process carried out by this system can be outlined. The uppermost box 1 in FIG. 1 indicates the step of fabricating the fuel pin subassembly, which is accomplished outside the scope of the present system. Box 2 of the flow diagram represents the loading or feeding of the fabricating fuel pin subassemblies into the system. This is accomplished by a cladding loader 15 comprising a suitable gravitational hopper for individually directing the lengths of fuel cladding 10 into a gravity feed conveyor shown generally at 16. The gravity feed conveyor directs parallel lengths of fuel cladding 10 from one station to the next in this system. The individual lengths of fuel cladding 10 are next loaded with fuel pellets as required by the reactor for which they are designed. This general step is represented by box 3 in the flow diagram. It is accomplished at a pellet loading station 23. The pellet loading station 23 is environmentally isolated from the remainder of the equipment in order to prevent radioactive contamination of the conveying and handling elements which necessarily contact the fuel cladding 10. To assure against contamination, the filled lengths of fuel cladding 10 are cleaned and temporarily plugged, steps represented by box 4 in the flow diagram. These steps are carried out in a cleaning and capping unit 25 at the entrance/exit of the pellet loading station 23. Final assembly of the fuel pin is represented by box 5, and is accomplished in an inerting enclosure 28 and welding station 30. The gas within the fuel cladding 10 is displaced by a desired inert atmosphere, the interior components of the fuel pin assembly are loaded adjacent to its open end, and the open end is sealed by a welded cap. The boxes 6 and 7 in the flow diagram represent inspection and release steps carried out in addition to the steps of this system prior to actual usage of the fuel pins. To complete the discussion of FIG. 3, it further represents a pellet loading tray 24 included within the pellet loading station 23. The loading tray 24 arranges individual pellets in elongated rows for insertion within individual lengths of fuel cladding. Also provided within the pellet loading station 23 is a funnel handling unit 26 which receives the funnels 12 after removal from each length of fuel cladding 10. It maintains the funnels in a guarded environment for subsequent disposal. Interspersed along the length of the gravity feed conveyor 16 are two cladding transports 36 and 37 aligned alongside the pellet loading station 23 and the welding station 30, respectively. Transports 36 and 37 are described in detail below. They basically position individual lengths of fuel cladding 10 for axial movement relative to the gravity feed conveyor 16. They also have the capability of imparting rotational movement to the fuel cladding 10 for rotation about their individual longitudinal axes. Axial and rotational movement of the cladding is coordinated with the functions of the equipment operating at the open end of the length of fuel cladding 10 during the steps carried out within this system. An accumulator 27 is arranged between transport 36 and the inerting enclosure 28. As will be described in greater detail below, the accumulator gathers a relatively large number of loaded lengths of fuel cladding, which are then fed as a batch into the inerting enclosure 28. To complete the system, fuel pin storage facilities 35 are provided downstream from transport 37. They are arranged to receive the completed fuel pins from the system, holding them for subsequent inspection and eventual release. The welding station 30, as shown schematically in FIG. 3, includes a welder 31, an end hardware loader 34, and an end cap feeder 32. Various components which must be directed into or onto the open end of each length of fuel cladding prior to welding of an end cap are moved into place by a barrel loader, generally shown at 33. The details of the gravity feed conveyor 16 are shown in FIGS. 5 through 7 and 9. The gravity feed conveyor, which leads between the cladding loader 15 and the first cladding transport 36, is longitudinally compressed by transverse folds arranged one above the other in a zigzag path. The path is constantly inclinded in a downward direction between loader 15 and transport 36. Conveyor 16 might be defined by a plurality of elongated rods arranged along the intended direction of travel of the cladding 10, by solid plates, or by a combination of rods and plates. As shown in the drawings, the lower reach of the conveyor 16 is illustrated as including top and bottom guide plates 40 which are mirror images of one another, the two plates being designed to alternately support the rolling lengths of fuel cladding 10 as the various folds along the conveyor are traversed. The illustrated conveyor guide plates 40 are bounded along their sides by end guide plates 43 which prevent unwanted axial movement of the fuel cladding as it rolls along the supporting guide plates 40. To accommodate the radially enlarged end of each cladding subassembly presented by funnel 12, block spacers 41 are inserted between adjacent lengths of the cylindrical fuel cladding 10. The dimensions of the spacers 41 are such as to assure clearance between adjacent funnels 12 on the lengths of cladding. The spacers are identical in size and shape to assure parallel positioning and consistent, free rolling of the engaged lengths of fuel cladding 10. The block spacers 41 are preferably made of cotton or other suitable disposable material that will not damage the outer surfaces of the cladding nor substantially interfere with rolling motion of the cladding along the gravity feed conveyor 16. The block spacers 41 slide along recessed spacer troughs 42 which extend along the lengths of the plates 40. At the end of plates 40 adjacent to the enlarged funnels 12 is provided a larger funnel trough 44, which provides clearance between the inner surfaces of guide plates 40 and the funnels 12. Thus, the lengths of cladding 10 roll along their uniform cylindrical surfaces, and are not misguided by rolling contact of the funnels 12. The lower guide plate 40 leads to an upright transverse stop 45 which spans the width of the cladding 10. Each of the spacer troughs 42 includes an aperture 46 which serves as a spacer exit through which the block spacers 41 can freely drop into a receiving spacer collector shown at 47. The block spacers 41 drop into the collectors 47 as each length of cladding 10 is lifted from contact against the stop 45. The dropping of block spacers 41 is illustrated in dashed lines in FIG. 9. The drawings illustrate details of one transport 36, which is designed to accommodate a single length of cladding 10. It is to be understood that multiple lengths of cladding could alternately be handled in a batch processing system by either duplicating the illustrated transport 36 or by designing transport 36 to support multiple lengths of fuel cladding simultaneously. The details of the transports 36 and 37, which are identical in structure, are illustrated in FIGS. 6, 8, and 9. Each transport includes a reciprocable trolley adapted to selectively move in a transverse direction across the gravity feed conveyor 16. The trolley comprises an upper elongated plate 51 and a lower elongated plate 52. Both are horizontal and one is elevationally spaced above the other. The trolley is supported by a pair of horizontal lower guide tracks 54 mounted to a supporting stationary framework and by a similar pair of stationary tracks 57 immediately above plate 51. Slide bearings 55 provided at intervals along the length of trolley 50 engage the lower tracks 54. They are driven by threaded engagement with a transverse lead screw 56. The lead screw 56 provides powered means operably connected to trolley 50 selectively operable for shifting the transport relative to the gravity feed conveyor in a transverse direction. Similar slide bearings 58 engage the upper tracks 57 to guide the upper portions of trolley 50 for reciprocable sliding movement in the transverse direction. Cladding support is provided on trolley 50 by paired sets of lower cladding support rollers 60 properly spaced so as to cradle each length of cladding 10 while it is positioned along a preselected operational axis defined by the longitudinal center axis of the cladding. Moveable upper support rollers 61 selectively engage the cladding 10 in opposition to the lower rollers 60 to thereby fix the cladding position on trolley 50. The upper rollers 61 are mounted on roller frames 62 which include motors (not shown) that power the rollers 61 to impart rotational movement to the lengths of fuel cladding. The cladding can therefore be selectively rotated about its central axis when desired by operation of the powered upper support rollers 61. The roller frames 62 for the upper support roller 61 are yieldably biased to an operational or lowered position by compression spring 63. Springs 63 are interposed between each roller frame 62 and the upper elongated plate 51. The roller frames 62 are guided on the frame of trolley 50 for limited vertical movement between a released or elevated position shown in full lines in FIG. 9 and a lowered or operational position shown in dashed lines. The raising of roller frames 62 is accomplished by small powered cylinders 64 which lift the frames 62 in opposition to the forces of springs 63. The lengths of cladding 10 are placed on the lower support roller 60 of trolley 50 by movable incline means shown as rotatable arms 65 intermittently spaced across the width of trolley 50. Arms 65 are powered for intermittent rotational movement about a transverse center axis along a common powered shaft 66. The outer ends of arms 65 include protrusions which selectively cradle each length of cladding 10 as it rests against the stop 45 across the gravity feed conveyor 16. The lower guide plate 40 of the gravity feed conveyor is provided with open recesses 67 which provide clearance for rotational pivotal movement of the arms 65 as they come upward beneath each stationary length of cladding. After engagement of the cladding, continued rotational movement of the arms 65 permits each length of cladding to roll along the inclined upper surfaces of the arms and into engagement with the opposite protrusions at the opposite or lower arm end. Further rotational movement deposits the cladding between the lower cladding support rollers 60. The lengths of cladding are removed from trolley 50 by a similar set of rotational arms 68 fixed to a separate powered shaft 59. The arms 68 eject each length of fuel cladding 10 and permit it to roll onto the receiving longitudinal rods or support plate surfaces of accumulator 27 (FIG. 9). The accumulator rods or plates are inclinded so as to continue gravitational movement of the cladding along the length of the system equipment. Loading of fuel pellets is accomplished at the pellet loading station 23. Operational details of this station are illustrated in FIGS. 10, 11 and 12. The cladding 10 is inserted through a contamination boundary wall 70 and held stationary by the supporting transport 36. A complementary fuel loading tube 71 is fitted within the interior of funnel 12, minimizing free particle contamination spread. Fuel pellets (not shown) are then guided through the fuel loading tube 71 and funnel 12 to the interior of the cladding 10. The pellet load for each length of cladding 10 can be prearranged along a pellet loading tray 24 (FIG. 3). Any suitable mechanism can be used to push the row of pellets through tube 71 and into the cladding. Contamination by release of particles is prevented by a dual seal shown in greater detail in FIG. 11. This comprises a cyclone valve 74 which surrounds an opening in the wall 70 through which the cladding 10 is projected for loading purposes. The valve 74 has a pressurized gas inlet 79 and an outlet for the resulting underflow at 76. An axial inlet 75 leading into the cyclone valve 74 surrounds the cladding 10. Air or gas outside wall 70 is drawn through inlet 75. It exhausts at 77, which draws air from within the cyclone valve 74 as well. The scrubbing action of the cyclone valve is illustrated by the arrows shown in FIG. 11. Combined with the cyclone valve 74 is an inflatable seal 78. This includes an elastic cylindrical diaphragm 80 which can be inflated and extended radially inward by pressurized air directed to a surrounding annular groove 81. During the loading sequence, the axial position of cladding 10 is such that the adjacent end of the shrink tubing 14 is positioned within the inflatable seal 78 and is overlapped by diaphragm 80. The shrink tubing 14 thereby prevents possible embedding of particles in the outer surface of the cladding 10. After loading of the cladding 10, it is retracted by transport 36, which draws the open end of the cladding through the boundary wall 70. As this is being accomplished, the shrink tubing 14, funnel removal ring 13 and funnel 12 are stripped from the open end of the cladding 10 by a removal clamp 82 that is movably mounted at the inner surface of wall 70 (FIGS. 10, 12). Clamp 82, which is solenoid operated, moves between a position clear of the fuel cladding 10 and a stripping position in which it straddles the shrink tubing 14 inwardly of the funnel removal ring 13. The adjacent shoulder on the funnel removal ring 13 engages the legs of the removal clamp 82 as the cladding is retracted from the pellet loading station 23, causing the shrink tubing 14, funnel removal ring 13 and funnel 12 to drop onto a receiving funnel removal tray 83. These items are then conveyed to the funnel handling unit 26 for disposal purposes. After the open end of the cladding 10 has passed into the cyclone valve 74 during its retraction, the opening in the wall 70 can be positively sealed by a solenoid controlled gate 84. Gate 84 prevents any further release of contaminating particles from the interior of pellet loading station 23. The cyclone valve 74 is used to strip debris from the cladding 10 as the cladding is removed from the pellet loading station 23. This is accomplished by the high velocity turbulent gas flow which surrounds the outer surfaces of the cladding 10 as indicated in FIG. 11. The cyclone valve 74 also prevents particle ejection from the interior of the pellet loading station 23. However, cleaning for smearable contamination is also required. This cleaning operation is accomplished in the cleaning and capping unit 25 (FIG. 3) and involves use of elements essentially illustrated in FIGS. 13 through 19. Referring to FIG. 13, the cleaning and capping unit is located for operation on the open end of cladding 10 between the inlet 75 of cyclone valve 74 and an apertured wall shown at 85. The equipment basically comprises an inside diameter cleaning head 86, an outside diameter cleaning assembly 87, and a capping assembly 88. The inside diameter cleaning head 86 is shown in detail in FIG. 17. The outside diameter cleaning assembly 87 is shown in FIGS. 14, 15 and 16. The capping assembly 88 is shown in FIGS. 18 and 19. The sequence of cleaning for smearable contamination is preferably to first clean the inside diameter, then the outside diameter, and finally the end of the cladding 10. Referring to FIGS. 13 and 17, cleaning of the inside diameter is accomplished by use of a disposable swab 90 supplied from an adjacent hopper 91 by an ejecting solenoid 92. The head 86 is movably supported on a radial arm 93 pivoted about a lead screw 94. Its angular position is controlled by a cylinder 95 operatively connected between arm 93 and a sliding bearing 96 that engages a guide rod 97. Guide rod 97 is parallel to lead screw 94. Cylinder 95 can be operated to selectively move the head 86 between a position aligned with the outlet of hopper 91 and clear of cladding 10, and an operational position coaxially aligned with the open outer end of the fuel cladding 10. To prepare the cleaning head 86 for operational use, it must first receive a swab 90, which is clamped within head 86 by a suitable collet assembly 98. As shown, the collet assembly 98 is mounted in a pivotable bearing having opposed control lever arms 100, 101 connected to a solenoid 102 and an opposing spring 103, respectively. Spring 103 normally maintains the collet 98 in a coaxial position aligned along the center of the fuel cladding 10. Solenoid 102, when activated, moves the swab 90 angularly about the transverse axis of the collet 98. The swab 90 is inserted into the open end of cladding 10 by operation of lead screw 94. After complete insertion, the cladding 10 will be rotated by operation of the upper support rollers 61 on the trolley 50 of transport 36. As the cladding 10 is rotated, the swab 90 will be drawn outwardly. Prior to such outward movement, the solenoid 102 is activated to pivot the swab 90 so as to frictionally engage the inner diameter surfaces of cladding 10. Once free of the cladding, the swab 90 can be ejected from the cleaning head 86 by operation of a solenoid 104. This cleaning operation can be repeated until normal removal of contamination at the interior of the cladding would be expected. The end edge or surface of cladding 10 is basically cleaned in the same manner as the inner diameter, again by use of swabs 90 and the cleaning head 86. The only difference is that the solenoid 102 is activated prior to engagement of the cladding 10 by swab 90. This pivots the outer end of swab 90 into position to engage the end edge of the cladding 10 as the cladding is being rotated. Cleaning of the outside diameter of cladding 10 is accomplished by use of disposable pads 105 supplied from a hopper 109 to the outside diameter cleaning assembly 87. A movable cleaning head 106 is supported for movement parallel to the cladding 10 by a lead screw 107 and parallel guide rod 108. Individual cleaning pads 105 are supplied from the hopper 109 by a solenoid-operated injector 110 which feeds the pads singly over the cladding 10 into a spring clamp assembly 111. The outside diameter cleaning assembly clamp 112 is then moved over the pad 105 by lead screw 107. Clamp 112 is operated by solenoid 113. It encircles the length of cladding 10 and firmly holds the slotted pad 105 about its exterior diameter. Once a cleaning pad 105 is securely gripped within clamp 112, lead screw 107 can be operated to draw the cleaning head 106 outwardly along the cladding as the cladding rotates. After passing beyond the open end of cladding 10, the pad will be discharged by release of solenoid 113 and operation of an ejecting solenoid 114. The cleaning head 106 can then be reciprocated back to its initial position adjacent to wall 85, where it can receive a subsequent pad 105 and the procedure be repeated. This process can be repeated as many times as required to achieve a clean condition. After cleaning of the open end of the fuel cladding 10, suitable monitors (not shown) can be translated along the cladding axis to evaluate its state of cleanliness. Based upon such an examination, a determination as to the need for further hand cleaning can be made. Assuming that hand cleaning is not required, a filter plug 115 is loaded into the open end of fuel cladding 10 by a capping assembly 88. The plug loader 116 of capping assembly 88 is supported on an arm 118 and lead screw 117 in essentially the same manner as previously described with respect to the cleaning head 86. Arm 118 is angularly positionable about the axis of lead screw 117 by operation of a cylinder 119. The cylinder 119 is supported by a slidable bearing 120 movably supported on a parallel guide rod 121. The plugs 115 are supplied from a hopper 122. They are ejected by a solenoid 123. After receiving a plug 115, the plug loader 116 is pivoted into an operational position coaxial with the open outer end of the clean cladding 10. An ejecting solenoid 124 within the plug loader 116 is selectively operational to frictionally insert the filter plug 115 within the open end of cladding 10. With filter plug 115 in place, the fuel pin subassembly is then clean relative to the gravity feed conveyor 16. However, a suitable preferred gas flow path and hood arrangement can enclose the gravity feed system as a secondary contamination boundary. Such a hood is schematically illustrated in FIG. 5 by the dashed lines 125. After the pellet loading process and cladding cleaning steps have been accomplished, the fuel pin subassemblies are individually removed from trolley 50 on transport 36 by operation of rotatable arms 68. They are permitted to roll along the receiving guide supports 126 of gravity feed conveyor 16 to a solenoid operated transverse stop 127 (FIG. 5). Stop 127 acts as an accumulator to gather the requisite number of fuel pin subassemblies for an inerting sequence. Inerting of the fuel pin subassemblies is accomplished in a batch sequence. It is performed within a rectangular pressure vessel schematically shown at 130 (FIG. 5). The inlet to the vessel 130 includes a solenoid actuated sealable door 131. A similar door 132 is provided at its outlet. The gravity feed conveyor, which extends through the vessel 130, maintains a minimum nuclear cross section from a criticality viewpoint. However, other arrangements of the inerting vessel are capable of being substituted in the system, such as a barrel arrangement set to one side of the principal conveyor path. After the predetermined charge of fuel pin subassemblies is contained within the pressure boundaries of vessel 130, the vessel and its contents can be evacuated and backfilled with the desired inerting gas. In a typical operational system, several hundred fuel pins might be inerted in a single batch, and the process might require several hours. Multiple vessels 130 can be interchanged within the system, depending upon production speed requirements. It should be noted that during the evacuation process, contamination by gas removal from the fuel pin subassemblies is prevented by the filter plugs 115. Following the inerting procedures, each fuel pin subassembly is directed to transport 37, which supports the cladding 10 for axial transverse movement relative to the gravity feed conveyor and for rotational movement about the cladding axis. While the cladding is positioned by transport 37, the filter plug 115 is removed, the reflector assemblies (if any) are inserted within the cladding, and an end cap is welded at the open end of the cladding to complete the fuel pin assembly. FIG. 24 shows a first welder capable of being used in the system. It makes use of a gas tungsten arc (GTA) welding system. The general details of the welder 31 are conventional. It should be noted that prior efforts to automate applications of a GTA welder for fuel pin usage have been limited due to electrode tip maintenance, which is sometimes required after welding of each individual fuel pin. This problem is overcome in the present system by providing an electrode drive mechanism 140 (FIG. 24) that can selectively move the tip of electrode 141 radially inward toward the operational axis of the welder 31. A rotatable grinder 142 is also selectively movable in a radial direction along an axis coaxial with the electrode 141. Grinder 142 is shaped to grind the electrode tip to the desired tip configuration as grinder 142 reaches a preselected stop position. Grinder 142 is operated to refurbish the electrode 141 after each fuel pin welding sequence, thereby eliminating the need for manual adjustment or checking of the electrode tip condition. As shown in FIG. 24, the incoming cladding 10 is directed by the supporting transport 37 to a sealed chamber 143 in communication with the inerting vessel 130. The transport 37 first positions the filter plug 115 directly beneath a retractable end plug remover 144. The end plug remover 144 includes a downwardly open C-clamp 145 which is complementary to an annular groove formed about the exterior of the filter plug 115. A solenoid or air cylinder 146 is selectively operable to position C-clamp 145 about the plug 115 with the surfaces of the C-clamp 145 engaging the shoulders of the plug groove. Transport 37 can then be operated to retract cladding 10, which allows plug 115 to fall into a capture tube 147. After removal of filter plug 115 has been completed, transport 37 is operated to axially shift the open end of cladding 10 into the welder 31. Cladding 10 is radially positioned and gripped by a bellows-operated collet 191 rotatably journalled within the welder 31 by bearings shown at 192. Thus, as cladding 10 is rotated by operation of transport 37, the collet 191 clamped to it will be freely rotated about its coaxial axis on the welder while providing axial and longitudinal support adjacent to the open end of cladding 10. After welding, the collet 191 is freed and seal 190 is released. The fuel pin can then be retracted through chamber 143 by operation of the transport 37. The cylindrical surfaces of cladding 10 can be sealed by an inflatable seal 190. The seal 190 includes a yieldable membrane which can be pressurized by air directed about a surrounding groove. Seal 190 is not required during operation of the GTA welder, but is utilized to seal off the interior of the welder 31 and barrel loader 33 from enclosure 28 during inerting operations. This can be accomplished by feeding a dummy pin into the welder and operating seal 190 to thereby close off the entrance into the inerting enclosure 28. By providing each batch of fuel pin assemblies with a final dummy pin for this purpose, the necessity of carrying out the inerting steps throughout the greater volume of the welder and loading equipment is prevented. After inerting has been accomplished, seal 190 can be relaxed. The atmospheres within the welder and loading equipment will be maintained as an inert atmosphere at all times. With the open end of cladding 10 located within the chamber 143, various barrel loaders can be directed through the welder 31 and into chamber 143 for placement of reflectors, tag gas capsules and other internal or external elements required to complete assembly of the fuel pin. General details of a barrel loader for inserting reflectors and end caps are shown in FIG. 21. A rotatable barrel frame 150 is shown supporting a reflector loader 160 and a diametrically opposite end cap loader 170. The details of loaders 160 and 170 are shown respectively in FIGS. 22 and 23. The barrel frame 150 is rotatably supported about an axis parallel to the axes of the loaders 160 and 170. It can be indexed about its axis to coaxially align alternate loaders 160 and 170 along the axis of welder inlet 148. The loader 160 is alternately aligned coaxially with a reflector injector 161 which receives reflectors 162 from a reflector hopper 163. Similarly, the end cap loader 170 can be indexed about the axis of the barrel frame 150 to align it coaxially with an end cap injector 171 which receives end caps 172 from a hopper 173 located to the rear of the view shown in FIG. 21. The reflector loader 160 comprises a tubular guide 164 having a shoulder 165 complementary in size and shape to the interior of the welder inlet 148 as shown in FIG. 22. A coaxial extension 166 protrudes from guide 164 and is insertable within the cladding 10. It includes an outer cavity which frictionally holds a reflector 162 for insertion purposes. A reciprocable plunger 167 is powered by a double acting pneumatic cylinder 168 operatively connected to a source of air mounted on the barrel frame 150. The details of the end cap loader 170 are shown in FIG. 23. It also includes a guide 174 having a shoulder 175 that fits within the welder inlet 148. No extension is provided on guide 174, since the end cap 172 is inserted directly at the open end of cladding 10 prior to welding. The end caps 172 are individually held by a collet including gripping spring fingers 176 at the end of a shaft 177 which is reciprocable within guide 174. Relative rotation between shaft 177 and guide 174 is prevented by interengagement between a radial pin 178 on the shaft 177 and a receiving longitudinal slot 180 formed through the guide 174. The guide 174 acts as a locking sleeve to urge the collet fingers 176 radially inward against the surfaces of end cap 172. This is accomplished by an abutting collar 181 threadably engaged about the outer end of shaft 177. Collar 181 is selectively rotatable by operation of a drive motor 182 (FIG. 21) operably connected to the collar 181. It can be operated to grip and lock an end cap 172, or to alternately release the end cap after the welding step has been completed. Both loaders 160 and 170 are mounted on the barrel frame 150 for free rotation about their individual axes, as well as for reciprocating movement parallel to their respective axes. The guides 164 and 174 include integral supporting bearings 153 which surround them and receive threaded lead screws 154 powered by longitudinal drive motors 155 operably connected to them by suitable drive gears. The bearings 153 also receive stationary guide rods 156 which stabilize their longitudinal movement as imparted by rotation of the respective lead screws 154. Each of the loaders 160 and 170 are movable between a retracted position, as shown in FIG. 21, and an extended operational position at which the shoulders 165 or 175 abut the interior of the welder inlet 148. After the filter plug 115 has been removed from the outer end of cladding 10 by operation of C-clamp 145, the barrel frame 150 is rotated by barrel drive 151 to properly index the reflector loader 160 in a coaxial position aligned with the welder inlet 148. The loader 160 is then shifted along its axis to bring shoulder 165 into engagement with the interior of inlet 148. At this time, the extension 166 will protrude to the open end of cladding 10, previously positioned within the welder 31. Extension 166 locates reflector 162 inwardly from the open end of the cladding, leaving clearance between the reflector and the subsequently added end cap 172. Reflector 162 is then injected into the cladding 10 by operation of cylinder 168 and plunger 167. After retraction of plunger 167, the reflector loader 160 is reciprocated clear of the welder inlet 148, and barrel frame 150 is again indexed by operation of barrel drive 151. The end cap loader 170, which had previously received an end cap 172, is subsequently aligned with the welder inlet 148. It is also reciprocated along its axis by the associated lead screw 154, bringing shoulder 175 into abutment with the interior of welder inlet 148. At this point, the projecting end cap 172 will be partially inserted within the interior of the cladding 10, with both the end cap 172 and cladding 10 properly positioned for operation of welder 31. As welder 31 is operated, the cladding 10 will be rotated about its longitudinal axis by operation of transport 37. End cap 172 will freely rotate in unison with it, allowing a complete circumferential weld to be formed about cladding 10 and end cap 172. When the welding sequence is finished, the collet fingers 176 within the end cap loader guide 174 are selectively released, permitting loader 170 to be retracted. The completed fuel pin assembly is now withdrawn from welder 31 by operation of transport 37. It can be discharged into fuel pin storage 35 for cleaning and subsequent inspection. An alternate form of welder 31 is shown in FIG. 25. It utilizes a ring loader exemplified by the device shown in FIG. 26 and a slightly modified end cap loader illustrated in FIG. 27. The details of the welding sequence are shown in FIGS. 28 and 29. Referring to FIG. 25, an opening is provided in communication with the interior of chamber 143 through an inflatable seal 193. This seal includes a flexible circumferential membrane that is selectively urged radially inward by pressurized air directed about an annular groove in a supporting sleeve 194. Seal 193 is capable of isolating the environment within chamber 143 from the environment within the welder. The welder 31 in this instance comprises a conventional pulsed magnetic welder of a cylindrical nature. It is arranged coaxially about a tube 195 having an inner diameter capable of receiving the cladding 10. In order to effect a weld by magnetic pulse, it is necessary to place a ring of curie metal about the area on cladding 10 which is to be welded. The magnetic material of ring 217 is thereby collapsed about the cladding as a result of the magnetic pulse produced by the welder. The resulting inward movement of the cladding surfaces causes development of a cold weld at the end of the cladding. Ring 217 is loaded on a guide 218 having a tapered nose 220 adapted to center the guide 218 coaxially in the open end of cladding 10. A radial shoulder 221 is adapted to abut the outer end edge of the cladding. The welding ring 217 can then be pushed onto the cladding by an abutting welding ring compressor 223. The welding ring compressor 223 is an extension of a loading rod 224 which is mounted to the barrel loader in the same manner as is the end cap loader and reflector loader. It is movable axially relative to the barrel loader and relative to the welder. One significant advantage of using pulsed magnetic welding techniques for closing the end of the fuel pin is that the welding can take place within a very confined area in tube 195. Because of the relatively small excess volume required in the welding area of this equipment, relatively expensive tag gas for identification of a fuel pin can be directed into tube 195 and the open end of cladding 10 prior to welding. This eliminates the fabrication and handling expenses of the usual tag gas capsules that are typically placed in the fuel pins and subsequently ruptured. The tube 195 includes a tapered shoulder 197 leading to an enlarged diameter groove 198 provided with a connection 200 for entrance of tag gas and a connection 201 for application of vacuum pressure. The connections 200 and 201 permit introduction of any gaseous environment desired, and alternation of the introduction of gas and vacuum pressure. The area within the welder 31 is sealed by a second inflatable seal 202. The diameter of seal 202 is complementary to the outer diameter of the bottom end cap loader used in conjunction with this equipment. The outer end of tube 195 is open and serves as a welder inlet. It is identified by the reference numeral 203. The end cap loader 209 provides for introduction of tag gas directly in the weld area. The guide 204 has an extension sleeve 211 slidably mounted over its outer end biased outwardly by a compression spring 212 wrapped about guide 204. Spring 212 normally maintains the sleeve 211 in the extended position shown in FIGS. 27 and 28. In this position, the fuel pin end cap 213 is retracted within the outer end of sleeve 211. The sleeve 211 includes apertures 214 which permit entrance of tag gas to the interior of sleeve 211. The sleeve 211 and guide 204 are sealed with respect to each other and with respect to shaft 207 by seals shown at 215 and 216. After the end cap loader has received and gripped an end cap 213, it is positioned coaxially with the welder inlet and shifted axially until the outer end of sleeve 211 engages the tapered shoulder 197 within tube 195. Seal 202 can then be engaged about sleeve 211, thereby defining a small volume sealed area between the seal 193 that encircles cladding 10 and the seal 202 that encircles sleeve 211. Tag gas is added to the atmosphere within the welder through the connection shown at 200. After addition of the tag gas, the barrel loader shifts the shaft 207 and end cap 213 into welding position within the open end of the cladding. An outwardly tapering surface formed on the end cap 213 is located radially inward of the ring 217. Ring 217 is collapsed about the cladding 10 and end cap 213 by a strong momentary magnetic pulse. The acceleration of the cladding as it collapses about the tapered surface on end cap 213 creates a cold weld at the outer edge of the cladding, sealing the cladding to the solid metal end cap 213 (FIG. 29). The collet fingers 206 can be released and end cap loader 209 can be retracted from the welder inlet to complete the welding sequence. The foregoing description of the preferred embodiments of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously many modifications and variations are possible in light of the above teaching. The embodiments discussed in detail were chosen and described in order to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto. |
description | This application is a divisional of prior U.S. patent application Ser. No. 10/059,044 filed Jan. 30, 2002, now abandoned which claims priority from prior provisional patent application Ser. No. 60/264,965 filed Jan. 30, 2001, the entire disclosures of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to feedwater spargers in boiling water reactors and, more particularly, to clamps for the end bracket assemblies of feedwater spargers and to methods of preventing separation of feedwater sparger end bracket assemblies. 2. Brief Discussion of the Related Art Conventional boiling water reactors typically include a reactor vessel, a shroud disposed within the reactor vessel and a fuel assembly within the shroud. Feedwater enters the reactor vessel via a feedwater inlet or nozzle and is distributed circumferentially within the reactor vessel by a feedwater sparger disposed in the reactor vessel between the shroud and the reactor vessel wall. The feedwater sparger comprises a ring-shaped pipe or conduit for carrying the feedwater and having an end attached to a sparger end plate via a feedwater sparger end weld, the sparger end plate and conduit end attached thereto defining an end of the feedwater sparger. A feedwater sparger end bracket assembly couples the end of the feedwater sparger to the reactor vessel wall in spaced relation therewith. The feedwater sparger end bracket assembly normally comprises an attachment plate connected to the sparger end plate via a weld, and the structural components of the feedwater sparger end bracket assembly are ordinarily connected to one another via one or more additional welds. The attachment plate and the sparger end plate attached thereto form a sparger/bracket junction by which the conduit of the feedwater sparger is connected to the feedwater sparger end bracket assembly. The feedwater sparger end bracket assembly defines a load path for transferring loads from the feedwater sparger to a reactor vessel attachment fitting attached to the reactor vessel wall and to which the feedwater sparger end bracket assembly is connected. The structural adequacy of feedwater sparger end welds and feedwater sparger end bracket assembly welds has been questioned in light of cracking identified in these welds. In particular, the weld between the attachment plate and the sparger end plate and the weld between the sparger end plate and the conduit end are primarily fillet welds, and reactor coolant can infiltrate or get between the structural components joined by these fillet welds so that the roots of the fillet welds are exposed to reactor coolant. The geometry of the fillet welds presents a crevice where corrosive products can concentrate and accumulate over time, thereby producing stress corrosion cracking. Cracks large enough to allow significant flow of feedwater from the feedwater sparger may result in direct impingement of the relatively colder feedwater on the reactor vessel wall, causing thermal shock and cracking of the cladding on the interior surface of the reactor vessel wall. In addition, the feedwater sparger end bracket assemblies usually carry an installation preload, and this preload is undesirably compromised or lost in the event of cracking of the feedwater sparger end welds and/or the feedwater sparger end bracket assembly welds, especially in the event of cracking which results in complete detachment of the feedwater sparger end bracket assembly from the feedwater sparger. An example of feedwater sparger end bracket assemblies that have an installation preload are those associated with feedwater spargers that are sprung into place during installation, such as to maintain contact between flow baffles of the feedwater spargers and the reactor vessel wall. The flow baffles, which are ordinarily located at the feedwater nozzles in the reactor vessel, must remain essentially in contact with the reactor vessel wall to effectively eliminate thermal shock conditions at the feedwater nozzles. To assure this, the feedwater spargers are sprung into place at installation, resulting in an installation preload on each feedwater sparger end bracket assembly of about eight thousand pounds. In the event of complete weld failure causing the feedwater spargers to become completely detached from the feedwater sparger end bracket assemblies, the installation preload is lost and the feedwater spargers will not perform as designed. Mechanical solutions to the problems of cracked feedwater sparger end welds and feedwater sparger end bracket assembly welds encounter numerous obstacles in that mechanical devices attached to the ends of the feedwater spargers and the feedwater sparger end bracket assemblies must be capable of maintaining the installation preload in the event of weld failure. Mechanical devices must be capable of balancing all loads and moments to which they are subjected, and particularly must react to the loads and moments created when there is a complete through wall crack of the feedwater sparger end welds and/or the welds of the feedwater sparger end bracket assemblies. Another deterrent to the use of mechanical devices to address the problems of weld failure in the ends of feedwater spargers and in feedwater sparger end bracket assemblies is that existing feedwater sparger end bracket assemblies often have different structural dimensions and/or components. The use of mechanical devices with feedwater sparger end bracket assemblies is thusly impeded by the difficulty involved in designing an essentially standard mechanical device for use with different feedwater sparger end bracket assemblies. A further impediment to the use of mechanical devices in response to cracking of feedwater sparger end welds and feedwater sparger end bracket assembly welds is the need for the mechanical devices to be installed using equipment or tooling operated from a location remote from the reactor vessel. Accordingly, it is a primary object of the present invention to overcome the problems associated with cracking of feedwater sparger end welds and feedwater sparger end bracket assembly welds. Another object of the present invention is to utilize a clamp to prevent separation of feedwater sparger end bracket assemblies. A further object of the present invention is to utilize a clamp to prevent separation of a feedwater sparger end from a feedwater sparger end bracket assembly welded to the feedwater sparger end. The present invention has as another object to constrain a feedwater sparger end bracket assembly from separation in horizontal, vertical and radial directions. It is also an object of the present invention to constrain a feedwater sparger end against separation from a feedwater sparger end bracket assembly in horizontal, vertical and radial directions. An additional object of the present invention is to avoid direct impingement of feedwater from a feedwater sparger on the reactor vessel wall in the event of cracking of a feedwater sparger end weld and/or a feedwater sparger end bracket assembly weld. Yet another object of the present invention is to maintain the preload of a feedwater sparger end bracket assembly in the event of failure of a feedwater sparger end weld and/or a feedwater sparger end bracket assembly weld. Still a further object of the present invention is to balance loads and moments on a clamp for a feedwater sparger end bracket assembly in the event of failure of a feedwater sparger end weld and/or a feedwater sparger end bracket assembly weld. The present invention has as an additional object to adapt a clamp for installation on feedwater sparger end bracket assemblies of various structural dimensions and/or components. Moreover, it is an object of the present invention to utilize a clamp to provide an alternate load path for loads from a feedwater sparger to a reactor vessel attachment fitting. Some of the advantages of the present invention are that the clamp can be adjustably tightened on the feedwater sparger end bracket assembly; the clamp encloses the feedwater sparger end bracket assembly; the clamp holds the feedwater sparger and the feedwater sparger end bracket assembly together in the event of weld failure in any of the welds of the feedwater sparger and/or the feedwater sparger end bracket assembly; the clamp incorporates corrosion resistant materials; the clamp does not require welding to the feedwater sparger end, to the feedwater sparger end bracket assembly or to the reactor vessel; thermal shock and cracking of the cladding on the interior surface of the reactor vessel wall are avoided; the moment created on the clamp in the event of through wall cracking of a feedwater sparger end weld and/or a feedwater sparger end bracket assembly weld is balanced; shims and/or spacers can be used to easily adapt the clamp for installation on feedwater sparger end bracket assemblies having different structural dimensions and/or components; clearances between the clamp and the feedwater sparger and/or between the clamp and the feedwater sparger end bracket assembly can be limited or controlled to ensure a tight fit; proper operation of flow baffles of the feedwater spargers is maintained; the clamp may be installed remotely; and the clamp may be used on both originally installed feedwater sparger end bracket assemblies and replacement feedwater sparger end bracket assemblies. These and other objects, advantages and benefits are realized with the present invention as generally characterized in a clamp for installation on a feedwater sparger end bracket assembly connected to a conduit of a feedwater sparger at a sparger/bracket junction in a boiling water reactor vessel. The clamp includes an upper clamp member for being assembled over a top of the feedwater sparger end bracket assembly, a lower clamp member for being assembled over a bottom of the feedwater sparger end bracket assembly and a connector securing the upper and lower clamp members to one another. The upper clamp member includes a compartment receiving an upper portion of the sparger/bracket junction, and the lower clamp member includes a compartment receiving a lower portion of the sparger/bracket junction. Each compartment comprises opposing walls constraining the sparger/bracket junction in a first direction tangential or horizontal to the boiling water reactor vessel. Each clamp member includes an inner shoulder along an inner side of the feedwater sparger end bracket assembly and an outer shoulder along an outer side of the feedwater sparger end bracket assembly for constraining the feedwater sparger end bracket assembly between the inner and outer shoulders in a second direction radial to the boiling water reactor vessel. The upper clamp member includes a lower surface along the top of the feedwater sparger end bracket assembly. The lower clamp member includes an upper surface along the bottom of the feedwater sparger end bracket assembly, and the feedwater sparger end bracket assembly is constrained between the lower surface and the upper surface in a third direction vertical to the boiling water reactor vessel. The upper clamp member further includes a recessed lower surface along the top of the conduit of the feedwater sparger, and the lower clamp member further includes a recessed upper surface along the bottom of the conduit, the conduit being constrained in the third direction between the recessed lower and upper surfaces. Each clamp member includes a shear tab positioned to be disposed between the conduit of the feedwater sparger and a wall of the boiling water reactor vessel with a close fit. The shear tab balances moments and loads to which the clamp is subjected. The upper clamp member comprises an impingement shield extending downwardly therefrom toward the lower clamp member, and the lower clamp member comprises an impingement shield extending upwardly to meet the impingement shield of the upper clamp member. The impingement shields are disposed between the sparger/bracket junction and the wall of the boiling water reactor vessel and serve to isolate the sparger/bracket junction from the wall of the boiling water reactor vessel. A method of preventing separation of a feedwater sparger end bracket assembly connected to a conduit of a feedwater sparger at a sparger/bracket junction in a boiling water reactor vessel is generally characterized in the steps of vertically separating an upper clamp member of a clamp from a lower clamp member of the clamp, locating the upper clamp member over a top of the feedwater sparger end bracket assembly, locating the lower clamp member over a bottom of the feedwater sparger end bracket assembly, moving the upper and lower clamp members toward one another to position an upper portion of the sparger/bracket junction within a compartment of the upper clamp member and to position a lower portion of the sparger/bracket junction within a compartment of the lower clamp member, securing the upper and lower clamp members to one another, and leaving the upper and lower clamp members in place to constrain the sparger/bracket junction in a first direction, to constrain the feedwater sparger end bracket assembly in a second direction, and to constrain the feedwater sparger end bracket assembly in a third direction. A feedwater sparger end bracket assembly that has the clamp installed thereon constitutes a constrained feedwater sparger end bracket assembly. Other objects and advantages of the present invention will become apparent from the following description of a preferred embodiment taken in conjunction with the accompanying drawings, wherein like parts in each of the several figures are identified by the same reference characters. A fragmentary portion of a conventional boiling water reactor 10 is illustrated in FIGS. 1-4 depicting a segment of a wall 11 of reactor vessel 12 and a length segment of feedwater sparger 14 disposed within reactor vessel 12. The reactor vessel wall 11 is generally cylindrical, and the complete reactor vessel wall extends upwardly, downwardly and circumferentially beyond the edges or borders of the wall segment shown in the drawings. A shroud (not shown) is disposed in the reactor vessel 12 in spaced relation with wall 11, and the feedwater sparger 14 is disposed in the circumferential gap or space between the shroud and the reactor vessel wall. The feedwater sparger 14 comprises a hollow conduit or pipe 15 that generally follows the circumferential curvature of the reactor vessel wall 11 and has a terminal end connected to a sparger end plate 16. The complete conduit 15 forms an annular or ring-shaped conduit or pipe in the reactor vessel 12 for carrying feedwater which enters the reactor vessel at one or more feedwater inlets or nozzles in communication with the lumen of conduit 15. The feedwater is distributed circumferentially within the reactor vessel 12 via interiorly directed outlet holes 17 in the conduit 15. A representative conduit 15 has concentric circumferential inner and outer side walls connected by planar top and bottom walls, with the holes 17 being formed in the inner side wall. The conduit 15 has a rectangular cross-sectional configuration with the major cross-sectional dimension thereof extending or oriented vertically in the reactor vessel 12 and the minor cross-sectional dimension thereof extending or oriented radially in the reactor vessel 12. A representative sparger end plate 16 is planar and is of generally rectangular peripheral configuration with the major dimension thereof also extending or oriented vertically to close off the open end of conduit 15. The sparger end plate 16 is typically attached to the end of conduit 15 by a feedwater sparger end weld 18, and the conduit end and the sparger end plate attached thereto define an end of feedwater sparger 14. The sparger end plate 16 is connected to a feedwater sparger end bracket assembly 20 which, in turn, is connected to a reactor vessel attachment fitting 21 connected, typically by welding, to reactor vessel wall 11. The feedwater sparger end bracket assembly 20 thusly couples the end of the feedwater sparger 14 to the reactor vessel 12 in spaced relation with reactor vessel wall 11. Accordingly, there is a circumferential gap or space between the feedwater sparger 14 and the reactor vessel wall 11 as best shown in FIG. 4. As used herein, the terms “top”, “bottom”, “upper”, “lower”, “upward” and “downward” are referenced in a vertical direction; the terms “inner” and “interior” refer to a direction toward a central longitudinal axis of the reactor vessel; the terms “outer” and “exterior” refer to a direction away from the central longitudinal axis of the reactor vessel; the terms “front” and “forward” refer to a direction away from the conduit end; the terms “back” and “rearward” refer to a direction toward the conduit end; the term “vertical” refers to the direction of the central longitudinal axis of the reactor vessel; the term “radial” refers to a direction radial to the reactor vessel wall; and the terms “horizontal” and “tangential” refer to a direction transverse or perpendicular to the vertical and radial directions. As shown in FIGS. 1-5, the feedwater sparger end bracket assembly 20 for a conventional boiling water reactor 10 ordinarily comprises an attachment plate 22 connected to sparger end plate 16 by a weld 23, a side plate 24 connected to attachment plate 22 and extending transversely or perpendicularly therefrom in the forward direction, and upper and lower bracket members 25 and 26, respectively, extending from attachment plate 22 in the forward direction. A representative attachment plate 22 is planar and is of generally rectangular peripheral configuration with the major dimension thereof extending vertically within the reactor vessel 12 and the minor dimension thereof extending radially in the reactor vessel 12. The peripheral configuration of the attachment plate 22 is larger than the peripheral configuration of the sparger end plate 16 such that the sparger end plate does not protrude beyond the periphery of the attachment plate. As best shown in FIG. 4, it is typical for an inner side edge of attachment plate 22 to be spaced interiorly beyond an inner side edge of sparger end plate 16. The attachment plate 22 is parallel to the sparger end plate 16 and has a rearward surface in facing abutment with a forward surface of the sparger end plate, the weld 23 being disposed between the abutting surfaces of the attachment plate and the sparger end plate. The attachment plate and sparger end plate may be provided with through holes to facilitate grasping thereof during installation, the attachment plate 22 being shown with upper and lower through holes aligned with upper and lower through holes of sparger end plate 16. A representative side plate 24 has a straight rearward edge in facing abutment with a forward surface of the attachment plate 22. A weld 27 is disposed between the abutting surfaces of the side plate 24 and the attachment plate 22 and thusly connects the side plate to the attachment plate. The side plate 24 is connected to the attachment plate 22 at a location adjacent or close to the inner side edge of attachment plate 22. The side plate 24 is planar and has a generally rectangular peripheral configuration with the major dimension thereof extending vertically in the reactor vessel 12 and the minor dimension thereof extending perpendicular to the attachment plate. Representative upper and lower bracket members 25 and 26 are also planar, and each bracket member has a straight rearward edge, a beveled forward edge, and parallel inner and outer straight side edges connecting the forward and rearward edges. The rearward edges of the upper and lower bracket members 25 and 26 are each in facing abutment with the forward surface of attachment plate 22, with the bracket members 25 and 26 being perpendicular to the side plate 24 and to the attachment plate 22. Welds 28 connect the abutting surfaces of the upper and lower bracket members 25 and 26, respectively, and the attachment plate 22. The straight inner side edge of each bracket member is in abutting relation with an outer surface of the side plate 24, and welds 29 connect the abutting surfaces of the upper and lower bracket members 25 and 26, respectively, and the side plate 24. The upper and lower bracket members 25 and 26 are parallel to and vertically spaced from one another to closely accommodate the reactor vessel attachment fitting 21 between a lower surface of the upper bracket member 25 and an upper surface of the lower bracket member 26. Accordingly, the inner side edge of upper bracket member 25 is connected to the side plate 24 at a location adjacent or close to an upper edge of side plate 24, and the inner side edge of lower bracket member 26 is connected to the side plate 24 at a location adjacent or close to a lower edge of the side plate 24. The representative feedwater sparger end bracket assembly 20 also includes a pin 30, extending through the upper and lower bracket members 25 and 26 and through the reactor vessel attachment fitting 21, by which the feedwater sparger end bracket assembly is secured or pinned to the reactor vessel attachment fitting. The pin 30 extends in the vertical direction transverse or perpendicular to the upper and lower bracket members 25 and 26, and is inserted through aligned passages in the upper and lower bracket members and the reactor vessel attachment fitting 21. An upper end or head of the pin 30 is disposed above the upper bracket member 25 and is sized and/or configured such that the head cannot pass through the passage in the upper bracket member 25. The passages in the upper and lower bracket members 25 and 36 are ordinarily oval or elliptical in cross-section and have centers, respectively, located centrally between the inner and outer side edges of the corresponding bracket member. A pin retainer 32 is engaged with or secured to the upper bracket member 25 and, as best seen in FIGS. 1 and 2, includes a bridge 33 centered over the head of pin 30 and a pair of legs 34 extending downwardly from opposite ends of bridge 33 to feet 35. The bridge 33 bisects the head of pin 30, and the legs 34 are spaced from one another a sufficient distance to accommodate the head of pin 30 therebetween. The representative pin retainer 32 is located so that the bridge 33 is perpendicular to the attachment plate 22 and parallel to the upper bracket member 25, with the legs 34 extending perpendicular to the bridge 33. The feet 35 extend outwardly from the legs 34, respectively, and are secured to or engaged with the upper bracket member 25 to prevent removal of the pin 30 after it has been inserted sequentially through the passage of the upper bracket member 25, the passage of the reactor vessel attachment fitting 21 and the passage of the lower bracket member 26. In a representative pin retainer 32, the feet 35 extend perpendicular to legs 34 such that the feet 35 are also perpendicular to the attachment plate 22. The representative pin retainer 32 is formed from an elongated, planar strip of material of relatively minimal thickness. A pin bail 36 of the feedwater sparger end bracket assembly 20 has a generally inverted U-shape straddling the head of pin 30. As best seen in FIGS. 3 and 5, pin bail 36 includes a cross-piece 37 extending over bridge 33 perpendicular thereto and spaced, parallel arms 38 extending downwardly from opposite ends of cross-piece 37 to tapered or angled lower ends secured to the head of pin 30. The cross-piece 37 bisects the head of pin 30 along an axis 90° to bridge 33, and the lower ends of arms 38 are secured to the head of pin 30 at locations, respectively, along this axis. The representative pin bail 36 is flat or planar with beveled outside corners joining the arms 38 to opposite ends of cross-piece 37, the pin bail 36 being parallel to attachment plate 22 and perpendicular to both the side plate 24 and the upper bracket member 25. The space between arms 38 defines an opening through which the bridge 33 passes. The pin bail 36 may be used for grasping to facilitate raising and lowering of pin 30 for insertion and/or removal through the aligned passages in the upper and lower bracket members 25 and 26 and the reactor vessel attachment fitting 21 when the pin retainer 32 is not secured in place on the upper bracket member. The reactor vessel attachment fitting 21 includes a base attached to the reactor vessel wall 11 and a nose protruding interiorly from the base in the radial direction. The nose is insertable between the upper and lower bracket members 25 and 26 when the pin 30 is removed from between the bracket members, and has a vertical dimension to fit closely between the upper and lower bracket members. A vertical passage through the nose is aligned with the aligned passages through the upper and lower bracket members 25 and 26, respectively. The pin 30 is inserted in the aligned passages, with insertion of the pin being facilitated by the pin bail 36. The pin 30 is secured in place in the aligned passages, thereby securing the feedwater sparger end bracket assembly 20 to the reactor vessel attachment fitting 21. The pin 30 may be secured in place via a threaded engagement or in any other suitable manner. The reactor vessel attachment fitting 21 can be formed integrally, unitarily with the reactor vessel wall 11 or as a separate component secured to the reactor vessel wall in any suitable manner, such as welding. The base of fitting 21 may have a curved outer end surface in abutting relation with the curved surface of the reactor vessel wall 11, and the curvature of the end surface preferably corresponds to the curvature of the reactor vessel wall. The feedwater sparger end bracket assembly 20 defines a load path for transferring loads from the feedwater sparger 14 to the reactor vessel attachment fitting 21 and the vessel wall 11. An alternative representative feedwater sparger end bracket assembly for use with the clamp of the present invention is illustrated in FIG. 6 at 120. The feedwater sparger end bracket assembly 120 is similar to feedwater sparger end bracket assembly 20 except that the feedwater sparger end bracket assembly 120 includes a shim plate 141 between the sparger end plate 116 and the attachment plate 122. Shim plate 141 is interposed between the forward surface of the sparger end plate 116 and the rearward surface of the attachment plate 122. The representative shim plate 141 is planar and has a generally rectangular peripheral configuration with a major dimension oriented vertically within the reactor vessel, which is not shown in FIG. 6. The peripheral configuration of shim plate 141 is typically larger than the peripheral configuration of both the sparger end plate 116 and the attachment plate 122 so that both the sparger end plate and the attachment plate are disposed within the peripheral configuration of the shim plate. The shim plate 141 has a planar rearward surface in abutting relation with the forward surface of the sparger end plate 116, and a weld 142 disposed between the abutting surfaces of the shim plate and the sparger end plate connects the shim plate to the sparger end plate. The shim plate 141 has a planar forward surface in abutting relation with the rearward surface of the attachment plate 122, and a weld 143 disposed between the abutting surfaces of the shim plate and the attachment plate connects the shim plate to the attachment plate. The additional welds 142 and 143 of the feedwater sparger end bracket assembly 120 present additional potential sites for stress corrosion cracking and concomitant weld failure in the feedwater sparger end bracket assembly 120. From the above, it should be appreciated that the feedwater sparger end bracket assemblies with which the clamp of the present invention may be utilized can include one or more shim plates, such as shim plate 141, to achieve proper fit of the feedwater sparger end bracket assembly with the corresponding reactor vessel attachment fitting. It should be further appreciated that proper fit of the feedwater sparger end bracket assembly with the corresponding reactor vessel attachment fitting may be achieved by varying the thickness of the attachment plate and/or the thickness of the sparger end plate, with or without the use of one or more shim plates. Where a stock thickness for the attachment plate is not sufficient to allow proper fit of the feedwater sparger end bracket assembly with the reactor vessel attachment fitting, one or more shim plates will typically be utilized to make the necessary adjustments. The radial and vertical locations of the attachment plate relative to the sparger end plate may also be varied, as needed, to obtain the proper fit. Since the reactor vessel for a typical boiling water reactor has a plurality of feedwater sparger end bracket assemblies, most typically eight feedwater sparger end bracket assemblies, it can be seen that the combined thickness of the sparger end plate and the attachment plate, including any shim plates, may vary. In other words, the thickness of the sparger/bracket junction, which includes the sparger end plate, the attachment plate and any shim plates, may not be the same for each feedwater sparger end bracket assembly of a boiling water reactor. The clamp of the present invention may be adapted for use on sparger/bracket junctions of different thicknesses, and may be used on originally installed feedwater sparger end bracket assemblies as well as replacement feedwater sparger end bracket assemblies. The feedwater spargers of boiling water reactors may be sprung into place during installation such that there is a shear load on the feedwater sparger end welds and the feedwater sparger end bracket assembly welds. One instance in which feedwater spargers are sprung into place involves the installation of replacement feedwater spargers having flow baffles at the feedwater nozzles of the reactor vessel to eliminate thermal shock conditions at the feedwater nozzles. In order for the flow baffles to work effectively, they must remain essentially in contact with the reactor vessel wall. This is assured by the feedwater spargers being sprung into place at installation, resulting in a load on each feedwater sparger end bracket assembly of approximately 8,000 pounds. In the event of a complete through wall crack of the weld between the conduit end and the sparger end plate and/or the weld between the sparger end plate and the attachment plate, causing the feedwater sparger end bracket assembly to become completely detached from the feedwater sparger, the installation preload would be lost and the feedwater sparger would not perform as designed. A clamp 44 for feedwater sparger end bracket assemblies is illustrated alone in FIGS. 7-12 and installed on the feedwater sparger end bracket assembly 20 in FIGS. 13-20. The clamp 44 includes first and second clamp members 45 and 46, respectively, and a connector 47 such as an externally threaded bolt or screw adjustably connecting the first and second clamp members in spaced relation. Each clamp member is constructed as a single piece or part, such that the clamp 44 consists of the connector 47 and two pieces or parts 45, 46. First clamp member 45 is an upper clamp member, second clamp member 46 is a lower clamp member, and the connector 47 adjustably connects the upper and lower clamp members in vertical spaced relation. The connector 47 extends through vertically aligned bores in the upper and lower clamp members 45 and 46, respectively, and the bores may be threaded to threadably engage a thread on the connector 47. Alternatively or additionally to the bores being threaded, a threaded nut may be provided on an end of the connector 47. Untightening or unthreading connector 47 allows the clamp to be moved from a closed position shown in FIGS. 7 and 8 to an open position in which the vertical spacing between the upper and lower clamp members is increased, thereby allowing installation of the clamp over a feedwater sparger end bracket assembly. If necessary, the connector 47 may be removed entirely from the lower clamp member to facilitate installation of the clamp on the feedwater sparger end bracket assembly as described further below. Tightening or threading connector 47 closes the clamp to decrease or reduce the vertical spacing between the upper and lower clamp members, thereby securing the clamp in the closed position on the feedwater sparger end bracket assembly as described further below. The upper clamp member 45 is a one-piece constructed and comprises a housing having a medial portion 48 between a forward extension 49 and a rearward extension 50 as shown in FIG. 7. The medial portion 48 extends interiorly beyond the forward and rearward extensions 49 and 50, and the bore through which the connector 47 passes extends entirely through the medial portion. The bore extends through the medial portion 48 in the vertical direction and has a central longitudinal axis coaxial with the bore of lower clamp member 46. The forward extension 49 has an L-shaped configuration with a radial extension portion and a horizontal extension portion extending forwardly from the radial extension portion at a right angle. As best shown in FIG. 8, an internal compartment 51 is defined in the upper clamp member 45 and is bounded at the top by a planar internal top wall 52, at the front by a planar internal front wall 53, at the rear by a planar internal rear wall 54 opposing internal front wall 53, at an inner side by a planar internal inner side wall 55, and at an outer side by a planar impingement shield 56 opposing internal inner side wall 55. The internal front and rear walls 53 and 54 are parallel and are perpendicular to internal top wall 52, internal inner side wall 55 and impingement shield 56. The internal inner side wall 55 is parallel to impingement shield 56, and both are perpendicular to internal top wall 52. The compartment 51 is open at the bottom of the upper clamp member, and a mouth or opening along the bottom of the upper clamp member provides communication with the compartment. The top of upper clamp member 45 has a stepped configuration with forward extension 49 protruding above the medial portion 48 and the rearward extension 50. Accordingly, the forward extension 49 has an external top wall 60 spaced upwardly from an external top wall 61 of medial portion 48 and rearward extension 50. The bottom of upper clamp member 45 has a stepped configuration so as to define a downwardly protruding inner shoulder or protrusion 62 extending downwardly from lower surface 63 and recessed lower surface 64 of upper clamp member 45, and a downwardly protruding outer shoulder or protrusion 65 extending downwardly from the lower surface 63. The lower surface 63 is at the bottom of the upper clamp member and is a forward lower surface located to the front of the opening into compartment 51. The recessed lower surface 64 is at the bottom of the upper clamp member and is a rearward lower surface located to the rear of the opening into compartment 51. The forward and rearward lower surfaces 63 and 64 are planar and parallel to one another, and are perpendicular to the internal front and rear walls 53 and 54, the internal inner side wall 55 and the impingement shield 56. Also, the forward and rearward lower surfaces 63 and 64 are parallel to internal top wall 52. The inner shoulder 62 extends along an inner side of upper clamp member 45 in the horizontal or tangential direction and extends between the front and rear walls of compartment 51. The inner shoulder 62 has a planar outer surface 67, coextensive with internal inner side wall 55, perpendicularly joined to the forward lower surface 63 to the front of the opening into compartment 51 and perpendicularly joined to the rearward lower surface 64 to the rear of the opening into compartment 51. The outer surface 67 of inner shoulder 62 extends downwardly from the opening of compartment 51 and is perpendicular to the internal front and rear walls 53 and 54. A shim pad 66 is secured on the outer surface 67 of inner shoulder 62 and extends vertically between the rearward lower surface 64 and a bottom surface of inner shoulder 62. The shim pad 66 extends forwardly from an external rear wall of medial portion 48 to terminate at a forward edge and has a depth or thickness in the radial direction. The forward edge of shim pad 66 is spaced rearwardly from the internal front wall 53 a sufficient distance for the thickness of the sparger/bracket junction to be accommodated in the compartment 51 between the internal front and rear wall 53 and 54. As best shown in FIG. 20 for the lower clamp member 46, a spacer 68 extends radially outwardly from the internal inner side wall 55 and may be secured to the shim pad 66 and/or to the internal inner side wall 55. The distance that the spacer 68 extends radially outwardly from the internal inner side wall 55 is selected such that the spacer limits the clearance between the upper clamp member 45 and the inner side wall of the conduit of the feedwater sparger as explained further below. The outer shoulder 65 extends along the horizontal extension portion of forward extension 49 parallel to inner shoulder 62 and has a planar inner surface 70 extending perpendicularly downwardly from the forward lower surface 63, the inner surface 70 being parallel to the outer surface 67 of inner shoulder 62. The outer shoulder 65 extends along the outer side of the upper clamp member opposite the inner side thereof and has a depth in the radial direction between inner surface 70 and a planar external outer surface 71 of forward extension 49. The outer surface 71 extends in a first direction, i.e. forwardly, from the compartment 51 and forms part of the external outer side wall for the upper clamp member 45. The depth of the outer shoulder 65 allows the outer shoulder to be accommodated between the lower bracket member of a feedwater sparger end bracket assembly and the reactor vessel wall as explained further below. The radial extension portion of forward extension 49 includes a planar external forward wall 72 perpendicularly joined to a planar external inner side wall 73 of the horizontal extension portion at an inside corner. The forward wall 72 has a notch 74 therein adjacent the inside corner at which the forward wall 72 is joined to the inner side wall 73. The forward wall 72 is perpendicularly joined to the forward lower surface 63 along a lower edge of the forward extension, and the notch 74 is located along the lower edge of the forward extension. The notch 74 is open at the front and at the bottom of the radial extension portion. The inner side wall 73 of the horizontal extension portion is parallel to the inner surface 70 of outer shoulder 65 and has a cavity 75 therein of generally rectangular cross-section. The inner side wall 73 is joined to the inner surface 70 of outer shoulder 65 by the forward lower surface 63, the inner side wall 73 being joined to forward lower surface 63 along the lower edge of the forward extension. The cavity 75 is open along the inner side and bottom of the horizontal extension portion, the cavity 75 being open along a lower edge of the horizontal extension portion at which the inner side wall 73 is perpendicularly joined to the forward lower surface 63. The rearward lower surface 64 is recessed upwardly relative to the forward lower surface 63 so that the rearward lower surface 64 is recessed upwardly from the bottom surface of inner shoulder 62 a greater distance than the forward lower surface 63 is recessed upwardly from the bottom surface of the inner shoulder. The rearward extension 50 is connected to the medial portion 48 at a curved inside corner and includes a straight horizontal segment extending rearwardly in the horizontal direction in alignment with the horizontal extension portion of forward extension 49. The horizontal segment of rearward extension 50 has an external outer side wall 77 coextensive with a downwardly protruding shear tab or protrusion 79 which protrudes downwardly from the rearward lower surface 64 along the outer side of the upper clamp member. The outer side wall 77 forms part of the external outer side wall for the upper clamp member and includes a rearward outer side wall segment along shear tab 79 and a forward outer side wall segment extending forwardly from shear tab 79 to impingement shield 56. The forward and rearward outer side wall segments are planar, with the forward outer side wall segment recessed inwardly from the rearward outer side wall segment. The shear tab 79, which is parallel to inner shoulder 62, begins at an external rear wall of the rearward extension 50 and extends forwardly therefrom along part of the horizontal or tangential length of the outer side of rearward extension 50. Accordingly, the shear tab 79 is spaced from compartment 51 in a second direction, i.e. rearwardly, opposite the direction of extension for outer shoulder 65 such that the inner shoulder 62 is located between the outer shoulder 65 and the shear tab 79 but is disposed closer to the outer shoulder than to the shear tab. The shear tab 79 has a planar outer side surface 78 formed by the rearward outer side wall segment and an inner side surface 80 parallel to the outer side surface 78. The inner side surface 80 is parallel to the outer surface 67 of inner shoulder 62 and is joined to the outer side surface 78 at a tapered lower end of the shear tab 79. The shear tab 79 has a depth or thickness in the radial direction between the outer side surface 78 and the inner side surface 80 and has a location rearward of the compartment 51 to fit between the outer side wall of the feedwater sparger conduit 15 and the reactor vessel wall 11 with a close fit as explained below. The impingement shield 56 may be formed integrally, unitarily with the housing of the upper clamp member 45 or as a separate component attached to the housing in any suitable manner so that the upper clamp member 45 is a single piece or part. The impingement shield 56 is illustrated as a separate component bolted or screwed to the housing. The impingement shield 56 extends vertically from the top of the upper clamp member 45 to a straight, horizontal lower end 82 best shown in FIG. 9. The impingement shield 56 is planar and has a depth or thickness between planar and parallel inner and outer shield surfaces, the outer shield surface being flush or substantially flush with the forward outer side wall segment of rearward extension 50. The lower end 82 of the impingement shield is of reduced thickness to overlap an upper end of an impingement shield for the lower clamp member 46 as explained further below. The impingement shield 56 completes the external outer side wall of upper clamp member 45 and extends from the outer side wall 77 of rearward extension 50 to the outer surface 71 of forward extension 49 such that the compartment 51 is closed off along the outer side of the upper clamp member. The impingement shield 56 extends vertically below the bottom surface of inner shoulder 62 a sufficient distance to meet the impingement shield of the lower clamp member in the closed position for clamp 44 as described below. The lower clamp member 46 is essentially a mirror image of upper clamp member 45 and is a one-piece construct comprising a housing having a medial portion 83 between a forward extension 84 and a rearward extension 85 as shown in FIG. 7. The medial portion 83 extends interiorly beyond the forward and rearward extensions 84 and 85, and has an external inner side wall including a top portion and a bottom portion angled downwardly from the top portion in the outward direction to meet the bottom wall of the lower clamp member. The bore in lower clamp member 46 through which the connector 47 passes extends entirely through medial portion 83 in the vertical direction. The forward extension 84 has an L-shaped configuration with a radial extension portion and a horizontal extension portion perpendicular to the radial extension portion thereof. An internal compartment 86, best shown in FIGS. 7 and 8, is defined in the lower clamp member 46 as a counterpart to compartment 51 of the upper clamp member 45. Internal compartment 86 is similar to compartment 51 and is bounded at the bottom by a planar internal bottom wall 87 which is parallel to internal top wall 52, at the front by a planar internal front wall 88 which is co-planar with internal front wall 53, at the rear by a planar internal rear wall 89 which is co-planar with internal rear wall 54, at the inner side by a planar internal inner side wall 90 which is co-planar with internal inner side wall 55, and at the outer side by a planar impingement shield 91 which is co-planar with impingement shield 56. The compartment 86 is open at the top of the lower clamp member and has a mouth or opening along the top of the lower clamp member providing communication with the compartment 86. The opening of compartment 86 is in facing relation to the opening of compartment 51 and is vertically aligned therewith. The bottom of lower clamp member 46 is defined by a planar external bottom wall. The top of lower clamp member 46 has a stepped configuration defining an upwardly protruding inner shoulder or protrusion 94 extending upwardly from planar upper surface 95 and planar recessed upper surface 96, and an upwardly protruding outer shoulder or protrusion 97 extending upwardly from the upper surface 95. The upper surface 95, which is at the top of the lower clamp member, is located to the front of the opening into compartment 86 and is a forward upper surface. The recessed upper surface 96 is located to the rear of the opening into compartment 86 at the top of the lower clamp member and is a rearward upper surface. The forward upper surface 95 corresponds to the forward lower surface 63 and is parallel to and vertically aligned with the forward lower surface 63. The rearward upper surface 96 corresponds to the rearward lower surface 64 and is parallel to and vertically aligned with the rearward lower surface 64. The inner shoulder 94 is similar to the inner shoulder 62 and is vertically aligned with the inner shoulder 62. The inner shoulder 94 extends along the inner side of lower clamp member 46 in the tangential or horizontal direction between the front and rear walls of compartment 86 and has a planar outer surface 98 coextensive with internal inner side wall 90, the outer surface 98 being perpendicular to the internal front and rear walls 88 and 89. The outer surface 98 of inner shoulder 94 is perpendicularly joined to forward upper surface 95 to the front of the opening of compartment 86 and is perpendicularly joined to rearward upper surface 96 to the rear of the opening into compartment 86. The outer surface 98 extends upwardly from the opening of compartment 86 and is co-planar with the outer surface 67 of the inner shoulder 62. A shim pad 66 is secured on the outer surface 98 as described above for the upper clamp member 45 and is located at a location corresponding to the shim pad 66 of the upper clamp member. The shim pad 66 of the lower clamp member 46 thusly extends vertically between the rearward upper surface 96 and a top surface of inner shoulder 94. A spacer 68, shown in FIG. 20, extends radially outwardly from the internal inner side wall 90 as described above for the spacer 68 of the upper clamp member. The spacer 68 for the lower clamp member is disposed at a location corresponding to the location for the spacer of the upper clamp member. The outer shoulder 97 is similar to the outer shoulder 65 and is vertically aligned with the outer shoulder 65. The outer shoulder 97 extends along the horizontal extension portion of forward extension 84 parallel to the inner shoulder 94 and has a planar inner surface 99 extending perpendicularly upwardly from the forward upper surface 95. The inner surface 99 of outer shoulder 97 is parallel to the outer surface 98 of inner shoulder 94, and is co-planar with the inner surface 70 of the outer shoulder 65 of the upper clamp member 45. The outer shoulder 97 extends along the outer side of the lower clamp member in the forward direction from compartment 86, and has a depth in the radial direction between inner surface 99 and a planar external outer surface 100 of the forward extension 84. The outer surface 100 forms part of the external outer side wall for the lower clamp member. The top of outer shoulder 97 and the bottom of outer shoulder 65 may have beveled or chamfered edges. The radial extension portion of forward extension 84 includes a planar external forward wall 101 joined to a planar external inner side wall 102 of the horizontal extension portion by an angled corner wall. The forward wall 101, the corner wall and the inner side wall 102 are perpendicularly joined to the forward upper surface 95 along an upper edge of the forward extension 84. The rearward upper surface 96 is recessed downwardly relative to the forward upper surface 95 so that the rearward upper surface 96 is recessed downwardly from the top surface of inner shoulder 94 a greater distance than the forward upper surface 95 is recessed downwardly from the top surface of the inner shoulder 94. The rearward extension 85 is similar to rearward extension 50 and is vertically aligned with the rearward extension 50. The rearward extension 85 has a horizontal segment corresponding to the horizontal segment of rearward extension 50 and has an external outer side wall 103 coextensive with an upwardly protruding shear tab or protrusion 105. The outer side wall 103 forms part of the external outer side wall for the lower clamp member and includes a rearward outer side wall segment along shear tab 105 and a forward outer side wall segment extending forwardly from shear tab 105 to impingement shield 91, with the forward outer side wall segment being recessed inwardly from the rearward outer side wall segment. Shear tab 105 is similar to shear tab 79 and protrudes upwardly from the rearward upper surface 96, the shear tab 105 being disposed at a location corresponding to the location of shear tab 79. Accordingly, the shear tab 105 is spaced from compartment 86 in the rearward direction, opposite the direction of extension for outer shoulder 97, such that the inner shoulder 94 is located between the outer shoulder 97 and the shear tab 105 but is disposed closer to the outer shoulder 97 than to the shear tab 105. The shear tab 105 has a planar outer side surface 104 formed by the rearward outer side wall segment of outer side wall 103 and has an inner side surface 106 parallel to the outer side surface 104. The inner side surface 106 is parallel to the outer surface 98 of inner shoulder 94 and is joined to the outer side surface 104 at a tapered upper end for the shear tab 105 as described for the shear tab 79. The outer side surface 104 of shear tab 105 is co-planar with the outer side surface 78 of shear tab 79. The inner side surface 106 of shear tab 105 is co-planar with the inner side surface 80 of shear tab 79. The impingement shield 91 is similar to impingement shield 56 and extends in the vertical direction from the bottom of the lower clamp member 46 to a straight, horizontal upper end 107 best shown in FIG. 9. The upper end 107 is of reduced thickness to overlap the lower end 82 of impingement shield 56 when the clamp 44 is in the closed position as shown in FIG. 9. With the clamp in the closed position, the planar inner surface of shield 56 is flush or substantially flush with a planar inner surface of shield 91, and the planar outer surface of shield 56 is flush or substantially flush with a planar outer surface of shield 91. The impingement shield 91 completes the external outer side wall of lower clamp member 46 and extends from the outer side wall 103 of rearward extension 85 to the outer surface 100 of forward extension 84 such that the compartment 86 is closed off along the outer side of the lower clamp member. The impingement shield 91 extends upwardly above the top surface of inner shoulder 94 such that the upper edge 107 engages the lower edge 82 of impingement shield 56 when the clamp 44 is in the closed position. The clamp 44 is preferably fabricated primarily from Austenitic 300 series stainless steel. If additional material strength is needed for certain components, XM-19 stainless steel may be used. In a preferred embodiment, the upper and lower clamp members each have a maximum horizontal or tangential dimension of about 16.60 inches and a maximum radial dimension of about 7.75 inches. The upper clamp member has a maximum vertical dimension or height of about 8.00 inches, and the lower clamp member has a maximum vertical dimension or height of about 6.5 inches. The connector is about 1.25 inches in diameter with seven threads per inch and is about 15 inches long including the head thereof. A method of preventing separation of feedwater sparger end bracket assemblies involves installing the clamp assembly 44 on a feedwater sparger end bracket assembly, such as feedwater sparger end bracket assembly 20, with the clamp being lowered into the reactor vessel 12 using remote tooling. The upper and lower clamp members are moved away from one another and are separated vertically to the extent required to obtain an open position for the clamp, allowing the clamp members to fit over the feedwater sparger and the feedwater sparger end bracket assembly. The upper clamp member 45 is disposed as one piece over the top of the sparger/bracket junction, and the lower clamp member 46 is disposed as one piece below the sparger/bracket junction with the connector 47 located interiorly of the feedwater sparger 14 and the feedwater sparger end bracket assembly 20. Accordingly, the upper and lower clamp members will be disposed in opposition to one another over the feedwater sparger end bracket assembly. An upper portion of the sparger/bracket junction, i.e. upper portions of sparger end plate 16 and attachment plate 22 which protrude upwardly beyond the conduit 15 and the upper bracket member 25, is aligned with the internal compartment 51 of the upper clamp member 45. A lower portion of the sparger/bracket junction, i.e. lower portions of sparger end plate 16 and attachment plate 22 which protrude downwardly beyond the conduit 15 and the lower bracket member 26, is aligned with the internal compartment 86 of lower clamp member 46. Of course, where the sparger/bracket junction includes a shim plate, an upper portion of the shim plate will be aligned with compartment 51 and a lower portion will be aligned with compartment 86. The connector 47 is then threaded into the lower clamp member, and the connector 47 and/or another tool is used to move the respective one-piece constructs of the upper and lower clamp members 45 and 46 toward one another. The upper and lower clamp members 45 and 46 are drawn together to obtain the closed position in which the clamp members enclose the upper and lower portions of the sparger/bracket junction. Since each clamp member is a one-piece construct, only two pieces or parts are required to be moved and drawn together during installation. The upper and lower clamp members are securely tightened on the feedwater sparger 14 and the feedwater sparger end bracket assembly 20, and are held in place via the connector 47. The clamp 44 will then be in a closed position with the lower end 82 of impingement shield 56 in overlapping engagement with the upper end 107 of impingement shield 91. The clamp is then left in place in the reactor vessel 12. FIGS. 13-20 illustrate the clamp 44 installed on the feedwater sparger end bracket assembly 20 to form a constrained feedwater sparger end bracket assembly, it being noted that the upper clamp member 45 is not shown in FIG. 20. With the clamp 44 installed on the feedwater sparger end bracket assembly 20, the upper portions of sparger end plate 16 and attachment plate 22 are disposed within the compartment 51 of the upper clamp member 45, and the lower portions of sparger end plate 16 and attachment plate 22 are disposed within the compartment 86 of lower clamp member 46. The leg 34 and/or foot 35 of pin retainer 32 disposed closest to the attachment plate 22 is accommodated in the notch 74. An outer side of pin bail 36 is accommodated in the cavity 75. The upper portions of sparger end plate 16 and attachment plate 22, i.e. the upper portion of the sparger/bracket junction, are constrained in a first or horizontal or tangential direction between the internal front wall 53 and the internal rear wall 54 of compartment 51 with a close fit, i.e. with minimal or no clearance therebetween. The lower portions of sparger end plate 16 and attachment plate 22, i.e. the lower portion of the sparger/bracket junction, are constrained in the horizontal or tangential direction with a close fit between the internal front wall 88 and the internal rear wall 89 of compartment 86. Separation of the feedwater sparger end bracket assembly 20 from the end of the feedwater sparger 14 in a horizontal or tangential direction is thusly prevented. The sparger/bracket junction is constrained in a second or vertical direction due to constraint of the sparger end plate 16 and the attachment plate 22 between the internal top wall 52 of the upper clamp member 45 and the internal bottom wall 87 of the lower clamp member 46. The upper and lower bracket members 25 and 26 are constrained in the vertical direction with a close fit between the forward lower surface 63 of the upper clamp member 45 and the forward upper surface 95 of the lower clamp member 46. The feedwater sparger conduit 15 is constrained in the vertical direction with a close fit between the rearward lower surface 64 of upper clamp member 45 and the rearward upper surface 96 of lower clamp member 46. Accordingly, separation of the feedwater sparger end bracket assembly is prevented in the vertical direction. The outer shoulder 65 of upper clamp member 45 is disposed between the upper bracket member 25 and the reactor vessel wall 11, with the inner surface 70 of outer shoulder 65 disposed adjacent the outer side edge of the upper bracket member with little or no clearance. The shear tab 79 of the upper clamp member 45 is disposed between the conduit 15 and the reactor vessel wall 11 with a close fit, the inner side surface 80 of the shear tab 79 being adjacent the outer side wall of the conduit with little or no clearance therebetween. The outer shoulder 97 of the lower clamp member 46 is disposed between the lower bracket member 26 and the reactor vessel wall 11, with the inner surface 99 of the outer shoulder 97 adjacent the outer side edge of the lower bracket member with little or no clearance therebetween. The shear tab 105 of the lower clamp member 46 is disposed between the conduit 15 and the reactor vessel wall 11 with a close fit, with the inner side surface 106 of the shear tab 105 adjacent the outer side wall of the conduit with little or no clearance therebetween. The outer surface 67 of inner shoulder 62 of the upper clamp member 45 is adjacent the inner side edge of attachment plate 22 with minimal or no clearance therebetween to establish a tight fit between the upper clamp member and the feedwater sparger end bracket assembly. The shim pad 66 of the upper clamp member 45 fills the gap or space between the outer surface 67 and the inner side wall of conduit 15 to establish a tight fit between the upper clamp member and the feedwater sparger. The spacer 68 of the upper clamp member 45 occupies the gap or space between the outer surface 67 and the inner side edge of the sparger end plate 16 as shown in FIG. 20 for the lower clamp member, thereby further ensuring a tight fit between the upper clamp member and the feedwater sparger. The lower clamp member 46 is installed on the feedwater sparger and bracket assembly in the same manner as the upper clamp member. The outer surface 98 of inner shoulder 94 of the lower clamp member 46 is adjacent the inner side edge of attachment plate 22, and the shim pad 66 of the lower clamp member fills the gap or space between the outer surface 98 and the inner side wall of conduit 15. The spacer 68 of the lower clamp member 46 occupies the gap or space between the outer surface 98 and the inner side edge of the sparger end plate 16 as shown in FIG. 20. Separation of the feedwater sparger end bracket assembly 20 is prevented in the radial direction due to the feedwater sparger end bracket assembly being constrained with a close fit between the inner shoulders 62, 94 and the outer shoulders 65, 97. Also, the feedwater sparger 14 is constrained in the radial direction due to constraint of conduit 15 between the inner shoulders 62, 94 and the outer shoulder 65, 97 via the shim pads 66 and due to radial constraint of the conduit 15 between the inner shoulders 62, 94 and the shear tabs 79, 105 via the shim pads 66. The impingement shields 56, 91 are disposed between the sparger/bracket junction and the reactor vessel wall 11. The impingement shields 56, 91 isolate the sparger/bracket junction from the vessel wall 11 and prevent direct impingement of feedwater flow on the reactor vessel wall 11 should a through wall crack develop between the conduit 15 and the sparger end plate 16. Accordingly, direct impingement of relatively colder feedwater on the reactor vessel wall 11 with concomitant thermal shock and cracking of the cladding on the interior surface of the reactor vessel wall is avoided. The shear tabs 79, 105 assist in carrying moment on the clamp 44 that occurs in the event of a complete through wall crack of the welds between the conduit 15 and the sparger end plate 16 or between the sparger end plate 16 and the attachment plate 22. In the event that the feedwater spargers are sprung into place such that there is a shear preload on the welds, the preload is maintained by the clamp 44 in the event of weld failure. If either of the welds between the conduit 15 and the sparger end plate 16 or between the sparger end plate 16 and the attachment plate 22 fails, the load from the feedwater sparger is transferred to the clamp 44 at location A shown in FIG. 20. The clamp 44 reacts to this load at location B, and the couple created by the loads at locations A and B creates a moment that must be balanced. The shear tabs 79, 105 react, at location C, to the couple created by the loads at locations A and B such that all loads and moments are balanced. Accordingly, the outer shoulders 65, 97 and the shear tabs 79, 105 cooperate with the inner shoulders 62, 94 to place the clamp and the feedwater sparger in equilibrium. Since the installation preload is transferred to and maintained by the clamp 44, flow baffles of the feedwater sparger will remain essentially in contact with the reactor vessel wall. Various shims and spacers can be incorporated in the clamp 44 to achieve a tight fit between the clamp and the feedwater sparger and/or the feedwater sparger end bracket assembly, as needed. Various shims and spacers can be incorporated in the clamp 44 to adapt the clamp for installation on feedwater sparger end bracket assemblies having various dimensions and components. The clamp can be installed on feedwater sparger end bracket assemblies having one or more shim plates, such as feedwater sparger end bracket assembly 120, by forming the internal compartments 51, 86 with a distance between the internal front walls and the internal rear walls sufficient to accommodate the sparger end plate, the attachment plate, and any shim plate. A clamp designed for installation on feedwater sparger end bracket assembly 120 can also be installed on the feedwater sparger end bracket assembly 20 by using shims and/or spacers to achieve a close fit between the feedwater sparger end bracket assembly 20 and the clamp. It should be appreciated, therefore, that the clamp 44 can be used on sparger/bracket junctions of different thicknesses. The shims and/or spacers can be incorporated in the clamp at any suitable locations. The clamp 44 positively secures to the feedwater sparger end bracket assembly and holds the feedwater sparger and the feedwater sparger end bracket assembly together in the event of through wall cracking of any or all feedwater sparger and/or feedwater sparger end bracket assembly welds. The clamp prevents separation of the feedwater sparger end bracket assembly in first, second and third directions, i.e. horizontal or tangential, vertical and radial directions. The clamp prevents separation of the feedwater sparger end bracket assembly from the feedwater sparger and prevents separation of the structural components of the feedwater sparger end bracket assembly. The clamp provides an alternate load path for loads from the feedwater sparger to the reactor vessel attachment fitting, and the outer shoulders and shear tabs provide an alternate path for transferring loads from the feedwater sparger to the reactor vessel attachment fitting. The clamp can be installed remotely from a refueling bridge using long-handled tooling. The clamp incorporates corrosion resistant materials and does not require welding to the feedwater sparger, the feedwater sparger end bracket assembly or to the reactor vessel. In as much as the present invention is subject to many variations, modifications and changes in detail, it is intended that all subject matter discussed above or shown in the accompanying drawings be interpreted as illustrative only and not be taken in a limiting sense. |
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description | 1. Field of the Invention The present invention relates to a radiation image capturing apparatus for capturing a radiation image of a subject by applying a radiation emitted from a radiation source to the subject and detecting the radiation that has passed through the subject with a radiation detector. 2. Description of the Related Art In the medical field, there have widely been used radiation image capturing apparatus, known as mammographic apparatus, which apply a radiation emitted from a radiation source to a breast of a subject and detect the radiation that has passed through the breast with a radiation detector. One known radiation detector for use in the radiation image capturing apparatus includes a solid-state detector in a laminated structure comprising a matrix of charge collecting electrodes formed on an insulating substrate and a radiation conductor disposed on the charge collecting electrodes for generating electric charges depending on the radiation that is applied. The electric charges generated by the radiation conductor and representing radiation image information are collected by the charge collecting electrodes and temporarily stored in an electric storage unit. The collected electric charges are converted into an electric signal, which is output from the solid-state detector. Other known radiation detectors include a radiation detector comprising a charge-coupled device (CCD) and a radiation detector comprising a combination of amorphous silicon and a scintillator. Furthermore, a stimulable phosphor panel which, when exposed to an applied radiation (X-rays, α-rays, β-rays, γ-rays, electron beams, ultraviolet radiation, or the like), stores part of the energy of the radiation, and, when subsequently exposed to applied stimulating light such as laser beam, visible light, or the like, emits stimulated light in proportion to the stored energy of the radiation, may also be used as a radiation detector. In order to obtain a high-quality radiation image captured by a radiation detector, as shown in FIGS. 7 and 8 of the accompanying drawings, a grid 6 is disposed in front of a radiation detector 2 for preventing scattered rays of a radiation X that are generated in a subject 4 from entering the radiation detector 2, as disclosed in Japanese laid-open patent publication No. 2005-13344. As well known in the art, the grid 6 is a convergent grid comprising an alternate assembly of radiation-permeable members 8 made of aluminum or the like which pass the radiation X therethrough and radiation-impermeable members 10 made of a material including lead or the like, the radiation-impermeable members 10 being inclined parallel to the direction in which the radiation X is applied to the grid 6. On the mammographic apparatus, it is customary to capture various radiation images of the breast in different directions, e.g., vertically, horizontally, and obliquely. Depending on the size of the breast, the breast may not be properly positioned in a prescribed position on the radiation detector 2. If the breast is not properly positioned in the desired position on the radiation detector 2, then the position of the radiation source 12 is changed into alignment with the position of the breast for appropriately irradiating the breast with the radiation X. When the position of the radiation source 12 is changed, however, since the direction in which the radiation X is applied and the direction in which the radiation-impermeable members 10 of the grid 6 are inclined are brought out of alignment with each other, part of the radiation X may possibly be vignetted by the radiation-impermeable members 10. In recent years, efforts have been made to perform tomosynthesis and stereoscopic imaging using mammographic apparatus. According to these imaging processes, the radiation source 12 is turned around the breast 4 in the directions indicated by the arrow α as shown in FIG. 7 to acquire a three-dimensional image or a desired sectional image of the breast 4. As the radiation source 12 is turned around the breast 4, the radiation X emitted from the radiation source 12 falls upon the grid 6 in constantly changing directions. Therefore, the radiation X is partly vignetted by the radiation-impermeable members 10 during the imaging process. In order to avoid the vignetting, the grid 6 may be turned in the directions indicated by the arrows β as shown in FIG. 7 in synchronism with the turning of the radiation source 12. Consequently, an additional mechanism is required to move the grid 6, and also an additional space for moving the grid 6 therein is required to allow the grid 6 to be turned in synchronism with the turning of the radiation source 12. Another problem is that the quality of the generated image of the breast 4 tends to be lowered because the positional relationship between the grid 6 and the radiation detector 2 varies as the grid 6 moves. It is a general object of the present invention to provide a radiation image capturing apparatus which is of a highly simple structure capable of capturing a high-quality radiation image by avoiding an image quality degradation due to the movement of a radiation source with respect to a subject. A major object of the present invention is to provide a radiation image capturing apparatus which is capable of capturing a high-quality radiation image by moving a radiation source only without the need for moving a grid. Another object of the present invention is to provide a radiation image capturing apparatus which is capable of capturing a high-quality radiation image free of shadows of a grid when the grid is movable. Still another object of the present invention is to provide a radiation image capturing apparatus which is capable of capturing a high-quality radiation image when the radiation image capturing apparatus is applied to tomosynthesis and stereoscopic imaging. The above and other objects, features, and advantages of the present invention will become more apparent from the following description when taken in conjunction with the accompanying drawings in which preferred embodiments of the present invention are shown by way of illustrative example. FIG. 1 shows in perspective a mammographic system 20 according to an embodiment of the present invention, to which a radiation image capturing apparatus according to the present invention is applied. As shown in FIG. 1, the mammographic system 20 includes an upstanding base 26, a vertical arm 30 fixed to a horizontal swing shaft 28 disposed substantially centrally on the base 26, a radiation source housing unit 34 storing a radiation source 22 (see FIG. 3) for applying a radiation X to a breast 44 (see FIG. 2) to be imaged of a subject 32 and fixed to an upper end of the arm 30, an image capturing base 36 housing a solid-state detector (radiation detector) 24 (see FIGS. 2 and 3) for detecting a radiation X that has passed through the breast 44 and a grid 23 and fixed to a lower end of the arm 30, and a compression plate 38 for compressing and holding the breast 44 against the image capturing base 36. When the arm 30, to which the radiation source housing unit 34 and the image capturing base 36 are secured, is angularly moved about the swing shaft 28 in the directions indicated by the arrow A, an image capturing direction with respect to the breast 44 of the subject 32 is adjusted. The radiation source housing unit 34 is coupled to the arm 30 by a hinge 35 and is angularly movable in the directions indicated by the arrow A independently of the image capturing base 36. The compression plate 38 that is coupled to the arm 30 is disposed between the radiation source housing unit 34 and the image capturing base 36. The compression plate 38 is vertically displaceable along the arm 30 in the directions indicated by the arrow B. To the base 26, there is connected a display control panel 40 for displaying image capturing information including an image capturing region, an image capturing direction, etc. of the subject 32, the ID information of the subject 32, etc., and setting these items of information, if necessary. FIGS. 2 and 3 show internal structural details of the image capturing base 36 of the mammographic system 20. In FIG. 2, the breast 44 of the subject 32 is shown as being placed between the image capturing base 36 and the compression plate 38. The reference numeral 45 represents the chest wall of the subject 32. The grid 23 is disposed over an upper front surface of the solid-state detector 24 and faces the radiation source 22. The grid 23 serves to remove scattered rays of the radiation X that are generated in the breast 44. The grid 23 comprises an assembly of radiation-permeable members 46 made of aluminum or the like which pass the radiation X therethrough and radiation-impermeable members 48 made of a material including lead or the like. The radiation-permeable members 46 and the radiation-impermeable members 48 extend substantially parallel to each other and also to the chest wall 45 of the subject 32 positioned against the image capturing base 36 along the directions indicated by the arrow A in which the radiation source 22 is angularly movable. The radiation-permeable members 46 and the radiation-impermeable members 48 are disposed alternately in a direction away from the chest wall 45. The radiation-impermeable members 48 are inclined to the horizontal plane of the grid 23 at respective angles θ that are progressively smaller away from the chest wall 45 in alignment with the direction in which the radiation X is applied from the radiation source 22. Therefore, the grid 23 serves as a convergent grid whose focal point is located at the radiation source 22. The grid 23 is reciprocatingly movable in the directions indicated by the arrow C which are perpendicular to the directions in which the radiation-permeable members 46 and the radiation-impermeable members 48 extend. The solid-state detector 24 comprises a two-dimensional matrix of photoelectric transducers made of amorphous selenium (a-Se) or the like. The solid-state detector 24 converts the radiation X applied to the photoelectric transducers into an electric signal and stores radiation image information represented by the radiation X as electric charge information. FIG. 4 shows in block form a control circuit of the mammographic system 20. As shown in FIG. 4, the mammographic system 20 includes a setting console 54 for setting subject information with respect to the age, sex, body type, subject identification number, etc. of the subject 32, image capturing conditions and an image capturing process for capturing a radiation image, etc., a radiation source controller 56 for controlling the radiation source 22 according to the set image capturing conditions including a tube current, a tube voltage, the types of a target and a filter in the radiation source 22, a calculated irradiation dose of the radiation X, a calculated irradiation time, etc., a shutter controller 60 for actuating a shutter 58 to block the radiation X when the grid 23 is reversed in its movement, a grid controller 62 for controlling the reciprocating movement of the grid 23 in the directions indicated by the arrow C (see FIGS. 2 and 3), and an image processor 64 for processing the radiation image of the breast 44 which is acquired from the solid-state detector 24. The mammographic system 20 according to the present embodiment is basically constructed as described above. Operation of the mammographic system 20 will be described below with reference to a flowchart shown in FIG. 5. Using the setting console 54 of the mammographic system 20, the operator, who is typically a radiological technician, sets subject information, image capturing conditions, an image capturing process, etc. (step S1). The subject information includes information as to the age, sex, body type, subject identification number, etc. of the subject 32, and can be acquired from an ID card or the like owned by the subject 32. The image capturing conditions include a tube current, a tube voltage, the types of a target and a filter, an irradiation dose of the radiation X, etc. for acquiring a suitable radiation image depending on the breast 44 which is a region to be imaged of the subject 32. The image capturing process represents information including a region to be imaged that is specified by the doctor, an image capturing direction that is specified by the doctor, etc. These items of information can be displayed on the display control panel 40 of the mammographic system 20 for the radiological technician to confirm. If the mammographic system 20 is connected to a network, these items of information can be acquired from a higher-level apparatus through the network. Then, the radiological technician places the mammographic system 20 into a certain imaging posture according to the specified image capturing process (step S2). For example, the breast 44 may be imaged as a cranio-caudal view (CC) taken from above, a medio-lateral view (ML) taken outwardly from the center of the chest, or a medio-lateral oblique view (MLO) taken from an oblique view. Depending on the information of a selected one of these image capturing directions, the radiological technician turns the arm 30 about the swing shaft 28. In FIG. 1, the mammographic system 20 is set to an imaging posture for taking a cranio-caudal view (CC) of the breast 44. Then, the radiological technician positions the breast 44 of the subject 32 with respect to the mammographic system 20. For example, the radiological technician places the breast 44 on the image capturing base 36, and thereafter lowers the compression plate 38 toward the image capturing base 36 to hold the breast 44 between the image capturing base 36 and the compression plate 38, as shown in FIG. 2 (step S3). FIG. 6 shows the manner in which the mammographic system 20 is set to an imaging posture for taking a medio-lateral oblique view (MLO) of the breast 44, and the breast 44 is fixed between the image capturing base 36 and the compression plate 38 for imaging a medio-lateral oblique view (MLO) thereof. In FIG. 6, since the position of the breast 44 is limited by the upper ends of the image capturing base 36 and the compression plate 38, the breast 44 may possibly be displaced to the lower end of the image capturing base 36, rather than being positioned centrally on the image capturing base 36, depending on the size of the breast 44. Depending on the position of the breast 44 fixed between image capturing base 36 and the compression plate 38, the radiological technician moves the radiation source housing unit 34 in one of the directions indicated by the arrow A to positionally adjust the radiation source 22 into substantial alignment with the center of the breast 44 (step S4). Since the radiation source 22 moves along the direction in which the radiation-impermeable members 48 extend, the radiation X emitted from the radiation source 22 will not be vignetted by the radiation-impermeable members 48. Then, the grid controller 62 actuates the grid 23 to reciprocate in the directions indicated by the arrow C (FIGS. 2 and 3) which are perpendicular to the direction in which the radiation-impermeable members 48 extend (step S5). At this time, the grid controller 62 should actuate the grid 23 to reciprocate within a range that is kept in the imaging zone for the breast 44 and in which the radiation X can reach the solid-state detector 24 without being vignetted by the radiation-impermeable members 48. Then, the radiation source controller 56 controls the tube voltage, the tube current, and the irradiation time of the radiation source 22 according to the image capturing conditions set in step S1 to energize the radiation source 22 for applying the radiation X to the breast 44 to capture a radiation image thereof (step S6). The radiation X that has passed through the compression plate 38, the breast 44, and the moving grid 23 is applied to the solid-state detector 24, which records a radiation image as electric charge information. The radiation image recorded in the solid-state detector 24 is then acquired by the image processor 64 (step S7). While the radiation X is being applied to the solid-state detector 24, the grid 23 reciprocates in the directions indicated by the arrow C. At the ends of the stroke of the grid 23, the speed of the grid 23 is nil. If the radiation X is applied to the solid-state detector 24 at the ends of the stroke of the grid 23, shadows of the radiation-impermeable members 48 will be formed in the radiation image. To avoid the drawback, the shutter controller 60 controls the shutter 58 to block the radiation X emitted from the radiation source 22 when the grid 23 approaches the stroke ends, i.e., nearly when the speed of the grid 23 becomes nil. Consequently, shadows of the radiation-impermeable members 48 are prevented from being formed in the radiation image. According to a modification, the shutter 58 is dispensed with, and the radiation source controller 56 supplies the radiation source 22 with tube current pulses at a frequency in phase with the frequency of reciprocating movement of the grid 23, for example, such that the radiation source 22 is turned off to interrupt the radiation X nearly when the speed of the grid 23 becomes nil. When the mammographic system 20 is applied to tomosynthesis, the radiation source 22 is moved through a predetermined angle in one of the directions indicated by the arrow A (step S8). Then, the image capturing cycle from step S6 is repeated until the imaging process is finished (step S9). Inasmuch as the radiation-impermeable members 48 extend along the directions indicated by the arrow A in which the radiation source 22 moves, the radiation X emitted from the radiation source 22 will not be vignetted by the radiation-impermeable members 48 when the radiation source 22 changes its position. Consequently, the mammographic system 20 is capable of generating a high-quality radiation image free of shadows of the radiation-impermeable members 48. The radiation image acquired while the radiation source 22 is moving in the directions indicated by the arrow A is processed by the image processor 64 to produce a sectional radiation image or a three-dimensional radiation image (step S10). The mammographic system 20 may employ a stimulable phosphor panel instead of the solid-state detector 24. Although certain preferred embodiments of the present invention have been shown and described in detail, it should be understood that various changes and modifications may be made therein without departing from the scope of the appended claims. |
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description | This application is a continuation of PCT International Application No. PCT/JP2013/060921 filed on Apr. 11, 2013, which claims priority under 35 U.S.C. 35 §119(a) to Japanese Patent Application No. 2012-096277 filed on Apr. 20, 2012. Each of the above applications is hereby expressly incorporated by reference, in its entirety, into the present application. 1. Field of the Invention The present invention relates to a radiation image detecting device having a dose detection sensor for performing exposure control of a radiographic image, and a radiation imaging system using the radiation image detecting device. 2. Description Related to the Prior Art In a medical field, a radiation imaging system, for example, an X-ray imaging system using X-rays as a kind of radiation is known. The X-ray imaging system is constituted of an X-ray generating device for producing the X-rays, and an X-ray imaging device for taking an X-ray image of an object by receiving the X-rays passed through the object (patient). The X-ray generating device has an X-ray source for emitting the X-rays to the object, a source control device for controlling the operation of the X-ray source, and an emission switch for inputting to the source control device a command to operate the X-ray source. The X-ray imaging device has an X-ray image detecting device for detecting the X-ray image based on the X-rays passed through the object, and a console, for controlling the operation of the X-ray image detecting device and saving and displaying the X-ray image. As the X-ray image detecting device, a flat panel detector (FPD) for detecting the X-ray image as an electric signal has come into widespread use. The FPD includes a detection panel having an imaging surface having a matrix of pixels for accumulating electric charge in accordance with an X-ray dose incident thereon and circuitry for operating the detection panel. The detection panel accumulates the signal charge on a pixel-by-pixel basis and converts the accumulated electric charge into a voltage signal by a signal processing circuit, to detect the X-ray image of the object and output the X-ray image as digital image data. Also there is a portable type X-ray image detecting device called electronic cassette, which contains the FPD in a cassette housing. In the X-ray imaging, there are cases where a scattered radiation removing member called grid is used for the purpose of reducing the influence of scattered radiation that is produced by the X-rays in passing through the object. The grid is disposed between the object and the X-ray image detecting device. The grid is composed of X-ray absorbing portions and X-ray transmitting portions arranged alternately. Each X-ray absorbing portion is made of an X-ray absorbing and opaque material such as lead into the shape of a slender strip. Each X-ray transmitting portion is made of an X-ray transparent material such as aluminum into the shape of a slender strip. Since the X-ray absorbing portions and the X-ray transmitting portions are arranged alternately in one direction, the X-ray absorbing portions and the X-ray transmitting portions form a stripe pattern. Such a grid is disposed between the imaging surface of the detection panel and the object. The use of the grid facilitates obtaining a high-contrast image with reduced influence of the scattered radiation, because most of the scattered radiation is absorbed by the X-ray absorbing portions in the grid before reaching the imaging surface. The grid is attached to an imaging stand or a housing of the X-ray image detecting device in use. One of items representing the type of the grid is a grid density that represents the number of the X-ray absorbing portions per unit width. There are various grid densities within the confines of 26/cm 100/cm, for example. Taking the case of a grid density of 40/cm (4/mm) as an example, a grid pitch, being the sum of the width of a pair of X-ray absorbing portion and X-ray transmitting portion is 250 μm. Some X-ray image detecting devices have an automatic exposure control (AEC) function, which stops an X-ray emission from the X-ray source at the instant when an X-ray dose emitted from the X-ray source has reached a predetermined emission stop threshold value, in order to perform exposure control of the X-ray image (refer to U.S. Pat. No. 6,952,465 corresponding to Japanese Patent Laid-Open Publication No. 2004-167075 and U.S. Pat. No. 6,944,266 corresponding to Japanese Patent Laid-Open Publication No. 2004-166724, for example). Such X-ray image detecting devices have dose detection sensors, which detect an X-ray dose passed through the object and output a signal corresponding to the detected dose. The U.S. Pat. No. 6,952,465 describes an X-ray image detecting device that is provided with stripe-shaped dose detection sensors having a length of 500 pixels in an imaging surface of a detection panel, besides pixels. According to the U.S. Pat. No. 6,952,465, the dose detection sensors are disposed such that a longitudinal direction (stripe direction) of the striped dose detection sensors is not in parallel with (for example, orthogonal to) a stripe direction of a grid. Accordingly, even if misalignment occurs in geometrical disposition being the positional relation between the grid and the dose detection sensors, stable AEC is performed by reducing variation in an output value of a signal outputted from the dose detection sensor. In other words, since the dose detection sensors are disposed in the imaging surface, the misalignment occurs in the geometrical disposition between the grid and the dose detection sensors owing to an attachment backlash of the grid, a manufacturing error of the grid, and the like. Each of X-ray absorbing portions and X-ray transmitting portions of the grid has a width of the order of micrometers. Thus, the attachment backlash or the manufacturing error of the grid easily causes the misalignment of the order of one X-ray absorbing portion or one X-ray transmitting portion between the grid and the dose detection sensors. The misalignment in the geometrical disposition between the grid and the dose detection sensors causes variation in the amount of X-rays incident upon the dose detection sensors, even if an X-ray emission amount is the same, and hence results in variation in the output value of the dose detection sensor. A variation range of the output value of the dose detection sensor is maximized in a case where the stripe direction of the dose detection sensors is in parallel with the stripe direction of the grid. For example, in a case where the stripe direction of the dose detection sensors is in parallel with the stripe direction of the grid, the striped dose detection sensors may be hidden behind the X-ray absorbing portions throughout its longitudinal direction, or contrarily situated behind the X-ray transmitting portions. In a case where the entirety of the dose detection sensor is hidden behind the X-ray absorbing portion, the X-ray incident amount is reduced throughout the dose detection sensor, and hence the output value of the dose detection sensor is minimized. On the contrary, even if the X-ray emission amount is the same, in a case where the entirety of the dose detection sensor is situated behind the X-ray transmitting portion, the X-ray incident amount is increased throughout the dose detection sensor, and hence the output value of the dose detection sensor is maximized. As described above, paralleling the stripe direction of the dose detection sensors to the stripe direction of the grid increases the variation range of the output value of the dose detection sensor caused by the misalignment in the geometrical disposition between the grid and the dose detection sensors. Thus, according to the U.S. Pat. No. 6,952,465, the striped dose detection sensors are disposed not in parallel with the stripe direction of the grid, so that a part of the dose detection sensor is always disposed behind the X-ray absorbing portion and another part of the dose detection sensor is always disposed behind the X-ray transmitting portion, even if the geometrical disposition between the grid and the dose detection sensors is misaligned. Thereby, the X-ray incident amount is relatively low at a part of the dose detection sensor, while relatively high at another part of the detection sensor, so the output value of the dose detection sensor is leveled. Therefore, as compared with the case of paralleling the stripe direction of the dose detection sensors to the stripe direction of the grid, it is possible to prevent the variation in the output value of the dose detection sensor caused by the misalignment in the geometrical disposition between the grid and the dose detection sensors, and carry out the stable AEC. In an embodiment of the U.S. Pat. No. 6,952,465, a pixel is of a size of 105 μm×105 μm. The dose detection sensor has a size of 500 pixels, and hence a length of the order of 105 μm×500=52500 μm (approximately 50 mm). The dose detection sensor is substituted for the pixels, or disposed in space between the adjoining pixels. In the case of disposing the dose detection sensor in the space between the pixels, the pixels adjoining to the dose detection sensor are downsized, to secure space for the dose detection sensor. A plurality of the dose detection sensors is disposed in a predetermined area. Also, the U.S. Pat. No. 6,944,266 describes an X-ray image detecting device in which detection pixels (referred to as AEC pixels in the U.S. Pat. No. 6,944,266) functioning as dose detection sensors are substituted as some of pixels, instead of providing the striped dose detection sensors. The U.S. Pat. No. 6,944,266 uses so-called non-destructive readout pixels from which an output value is read out while the pixels keep holding accumulated electric charge, and the detection pixels are also read out in a non-destructive manner. In the U.S. Pat. No. 6,944,266, since the detection pixels are disposed in an imaging surface, misalignment in the geometrical disposition between a grid and the detection pixels causes variation in an output value of the detection pixel, just as with the U.S. Pat. No. 6,952,465. The U.S. Pat. No. 6,944,266 deals with the problem of variation in the output value of each detection pixel by calibration, which calibrates the output value of each detection pixel. To be more specific, according to the U.S. Pat. No. 6,944,266, the X-rays are evenly applied to the imaging surface having the grid attached, to obtain a gain image representing the output value of each detection pixel in the imaging surface. In the gain image, variation in the output values of the detection pixels in a state of attaching the grid is reflected. In AEC of actual imaging, the output value of each detection pixel is calibrated with the gain image to correct the output value of each detection pixel. The output value of each detection pixel varies with not only the type of the grid specifying the grid density, but also an imaging condition including an X-ray dose and X-ray quality depending on a tube voltage. Even if the type of the grid and the imaging condition are the same, an attachment backlash of the grid or a manufacturing error causes misalignment in the geometrical disposition between the grid and the detection pixels, so that the gain image is obtained whenever imaging is carried out. In the U.S. Pat. No. 6,952,465, the space for the striped dose detection sensors is secured by substituting or downsizing the pixels, so the obtained X-ray image has low density at portions corresponding to the dose detection sensors. The difference in density is large between the portion corresponding to the dose detection sensor and other portions adjoining thereto, and hence manifests itself as a strip of density step. Since the striped dose detection sensors have a length of approximately 50 mm, being a size visible to a human eye, the density step of the X-ray image is also of a size visible to the human eye and very conspicuous. To eliminate such a density step, the U.S. Pat. No. 6,952,465 discloses performing a defect correction with regarding the dose detection sensors as defect pixels, but a defect correction requires an effort at preparing correction data. Furthermore, the dose detection sensors are large, being approximately 50 mm. Thus, it is difficult to completely eliminate the defect by the defect correction to the extent of being invisible, and it is feared that the image quality of the X-ray image may be degraded. In the case of the U.S. Pat. No. 6,944,266, some of the pixels are used as the detection pixels and the output values of the detection pixels are read out in a non-destructive manner, so no density step occurs in the X-ray image, in contrast to the U.S. Pat. No. 6,952,465. Therefore, there is no problem of efforts at the defect correction and no problem of degradation in the image quality of the X-ray image. However, in the case of the X-ray image detecting device described in the U.S. Pat. No. 6,944,266, the gain image has to be obtained whenever imaging is carried out, and hence there is another problem that obtaining the gain image requires time and effort. Furthermore, if the geometrical disposition between the grid and the detection pixels is misaligned after obtainment of the gain image and before carrying out actual imaging, the correction cannot be performed appropriately with the obtained gain image. An object of the present invention is to provide a radiation image detecting device and a radiation imaging system that can perform stable AEC that is insusceptible to misalignment in geometrical disposition between a grid and dose detection sensors with having little fear of degradation in image quality and requiring no time and effort. A radiation image detecting device according to the present invention includes a detection panel and a plurality of dose detection sensors, and carries out imaging by using a scattered radiation removing grid having radiation absorbing portions for absorbing radiation and radiation transmitting portions for transmitting the radiation alternately and periodically arranged in a first direction. The detection panel has an imaging surface provided with a plurality of pixels for converting the radiation into an electric signal, and detects a radiographic image of an object. The plurality of dose detection sensors are provided for performing exposure control of the radiographic image. The plurality of dose detection sensors is disposed in the imaging surface periodically with leaving space in the first direction, for detecting a dose of the radiation passed through the object and outputting a signal in accordance with the dose. An arrangement period of the radiation absorbing portions is different from an arrangement period of the plurality of dose detection sensors in the first direction in the imaging surface. It is preferable that the arrangement period of the dose detection sensors be not an integral multiple of the arrangement period of the radiation absorbing portions. It is preferable that each of the arrangement period of the dose detection sensors and the arrangement period of the radiation absorbing portions have a length in unit of the number of the pixels, and the arrangement periods be co-prime numbers. It is preferable that an arrangement period of the dose detection sensors in a second direction orthogonal to the first direction be also different from the arrangement period of the radiation absorbing portions. The arrangement period of the dose detection sensors in the second direction is preferably the same as the arrangement period of the dose detection sensors in the first direction. A minimum size of the dose detection sensor is preferably the same as the size of the pixel in the imaging surface. It is preferable that the dose detection sensors be detection pixels as which some of the pixels are utilized. In a case where a plurality of the detection pixels are disposed with being shifted by one or more rows and one or more columns in each of a row direction corresponding to the first direction and a column direction corresponding to the second direction, the arrangement period in the first direction is preferably a length in the row direction, and the arrangement period in the second direction is preferably a length in the column direction. The dose detection sensor is preferably a detection pixel group composed of a plurality of the detection pixels adjoining each other. It is preferable that the dose detection sensor output the signal in accordance with the dose per unit of time, and the radiation image detecting device further include an automatic exposure control section for integrating an output value of the dose detection sensor, and comparing an integral value with an emission stop threshold value set in advance, and stopping emission of the radiation from a radiation source upon the integral value reaching the emission stop threshold value. The automatic exposure control section preferably calculates an average of the output values of a plurality of the dose detection sensors, and obtains the integral value by integrating the calculated average. It is preferable that the scatter radiation removing grid be detachably attached. A radiation imaging system according to the present invention includes a radiation generating device having a radiation source for emitting radiation and a radiation image detecting device for detecting a radiographic image, and carries out imaging by using a scattered radiation removing grid having radiation absorbing portions for absorbing radiation and radiation transmitting portions for transmitting the radiation alternately and periodically arranged in a first direction. The radiation image detecting device includes a detection panel and a plurality of dose detection sensors. According to the present invention, the arrangement period of the plurality of dose detection sensors disposed in the imaging surface with leaving space is different from the arrangement period of the radiation absorbing portions in the grid. Accordingly, not all of the plurality of dose detection sensors is situated behind the radiation transmitting portions or the radiation absorbing portions. Therefore, it is possible to perform the stable AEC that is insusceptible to misalignment in the geometrical disposition between the grid and the dose detection sensors with having little fear of degradation in the image quality and requiring no time and effort. As shown in FIG. 1, an X-ray imaging system 10 according to the present invention is constituted of an X-ray generating device 11 for generating X-rays and an X-ray imaging device 12 for taking an X-ray image from the X-rays passed through a patient M, being an object. The X-ray generating device 11 includes an X-ray source 13 for emitting the X-rays, a source control device 14 for controlling the X-ray source 13, and an emission switch 15 for commanding the start of X-ray emission. The X-ray imaging device 12 includes an electronic cassette 16 being a portable X-ray image detecting device, a console 17 for controlling the electronic cassette 16, and an imaging stand 30. The source control device 14, the electronic cassette 16, and the console 17 are connected wiredly or wirelessly so as to be communicated with each other. An upright type imaging stand is used as the imaging stand 30 in this embodiment, but a bed type imaging stand may be used instead. The electronic cassette 16 is detachably loadable in the imaging stand 30. The electronic cassette 16 is composed of an image detector 35 (see FIG. 4) having a detection panel 35a formed with an imaging surface 36, and a portable flat housing (not shown) containing the image detector 35. The shape of the electronic cassette 16 in plane is, for example, a square having vertical and horizontal sides of equal length. The electronic cassette 16 is detachably set in a holder 30a of the imaging stand 30 and held in such a position that the imaging surface 36 (see FIG. 4) of the detection panel 35a is opposed to the X-ray source 13. Note that, the electronic cassette 16 is sometimes used by itself in a state of being put on a bed under the patient M lying or held by the patient M himself/herself, instead of being set in the imaging stand 30. The X-ray imaging device 12 can perform imaging with the use of a scattered radiation removing grid (hereinafter called grid) 18 for removing scattered radiation produced at the time when the X-rays pass through the patient M. The grid 18 is a thin plate of approximately the same size as the electronic cassette 16. The grid 18 is detachably attached to the holder 30a of the imaging stand 30 together with the electronic cassette 16. The grid 18 is disposed in the holder 30a in such a position as to be opposed to the imaging surface 36 of the electronic cassette 16. Thus, the grid 18 is disposed between the patient M and the electronic cassette 16 during the imaging. Since the grid 18 is detachable from the holder 30a, the grid 18 may be exchanged or detached from the holder 30a in X-ray imaging according to an imaging purpose. There is no mechanism for swinging the grid 18, and hence the grid 18 is a so-called static grid fixed in a set position. Note that, the grid 18 may be detachably attached to the electronic cassette 16. Also in this case, the grid 18 may be exchanged or detached in the X-ray imaging according to the imaging purpose. The X-ray source 13 has an X-ray tube 13a for radiating the X-rays and a radiation filed limiter (collimator) 13b for limiting an irradiation field of the X-rays that the X-ray tube 13a radiate. The X-ray tube 13a has a cathode made of a filament for emitting thermoelectrons, and an anode (target) that radiates the X-rays by collision of the thermoelectrons emitted from the cathode. The radiation field limiter 13b is composed of, for example, four X-ray shielding lead plates disposed on each side of a rectangle so as to form an irradiation opening in its middle through which the X-rays propagate. A shift of the lead plates varies the size of the irradiation opening to limit the irradiation field. As shown in FIG. 2, the source control device 14 includes a high voltage generator 20, a controller 21, and a communication I/F 22. The high voltage generator 20 generates a high tube voltage by multiplying an input voltage by a transformer, and supplies the tube voltage to the X-ray source 13 through a high voltage cable. The controller 21 controls the tube voltage for determining an energy spectrum of the X-rays emitted from the X-ray source 13, a tube current for determining an emission amount per unit of time, and an emission time of the X-rays. The communication I/F 22 mediates transmission of primary information and signals from and to the console 17. To the controller 21, the emission switch 15, a memory 23, and a touch panel 24 are connected. The emission switch 15 is a switch operated by an operator e.g. a radiological technician at the start of imaging, and is a two-step push switch, for example. Upon a first step push of the emission switch 15, a warm-up start signal is issued to start warming up the X-ray source 13. Upon a second step push, an emission start signal is issued to make the X-ray source 13 start the X-ray emission. These signals are inputted to the source control device 14 through a signal cable. Upon receiving the emission start signal from the emission switch 15, the controller 21 starts electric power supply from the high voltage generator 20 to the X-ray source 13. The memory 23 stores in advance several types of imaging conditions each including the tube voltage, the tube current, and the emission time or a tube current-time product (mAs). The imaging condition is set manually by the operator through the touch panel 24. The source control device 14 starts controlling the X-ray emission based on the set imaging condition, including the tube voltage, the tube current, and the emission time or the tube current-time product. The electronic cassette 16, having an AEC function, detects a dose of the X-rays applied from the X-ray source 13 per unit of time. At the instant when it is detected that an integral dose of the X-rays has reached an adequate target value, the AEC function stops the X-ray emission even if actual emission time or actual tube current-time product is equal to or less than the value set in the source control device 14. Note that, a value having an adequate margin is set as the emission time or the tube current-time product in the source control device 14, for the purpose of preventing a dose shortage, more specifically preventing a situation that the X-ray emission is completed, before the integral dose reaches the target value and AEC function judges the stop of the X-ray emission. A maximum value of the emission time, which is determined in accordance with a body part to be imaged under safety regulations, may be set in the source control device 14. Note that, another imaging condition that is transmitted from the console 17 through the communication I/F 22 may be set. An emission signal I/F 25 is connected to the electronic cassette 16 in the case of using the AEC function of the electronic cassette 16. In this case, upon receiving the warm-up start signal from the emission switch 15, the controller 21 transmits an emission start request signal, which queries whether or not the X-ray emission can be started, to the electronic cassette 16 thorough the emission signal I/F 25. In response to the emission start request signal, the electronic cassette 16 performs preparation processing. After the completion of the preparation processing and standing ready to perform imaging, the electronic cassette 16 transmits an emission permission signal to the source control device 14. Upon receiving the emission permission signal from the electronic cassette 16 at the emission signal I/F 25 and further receiving the emission start signal from the emission switch 15, the controller 21 starts electric power supply from the high voltage generator 20 to the X-ray source 13 to emit the X-rays. Moreover, the controller 21 stops the electric power supply from the high voltage generator 20 to the X-ray source 13 to stop the X-ray emission, upon receiving an emission stop signal from the electronic cassette 16 at the emission signal I/F 25. Furthermore, a timer is embedded in the controller 21 to stop the X-ray emission when the set emission time has elapsed, in addition to the function of stopping the X-ray emission upon receiving the emission stop signal. As shown in FIG. 3, the grid 18 has approximately the same size and shape as the electronic cassette 16. The grid 18 has strip-shaped X-ray transmitting portions 32 and X-ray absorbing portions 33 shown with hatching, which extend in a Y1 direction corresponding to a second direction of the present invention. The X-ray transmitting portions 32 and the X-ray absorbing portions 33 are arranged alternately and periodically in an X1 direction, which corresponds to a first direction of the present invention, orthogonal to the Y1 direction. The X-ray transmitting portion 32 is made of an X-ray transparent material, e.g. aluminum or the like. The X-ray absorbing portion 33 is made of an X-ray absorbing material with high X-ray shieldability, e.g. lead, a molybdenum alloy, a tantalum alloy, or the like. The grid 18 absorbs at the X-ray absorbing portions 33 the scattered radiation that the X-rays produce in passing through the patient M, and thereby prevents reduction in contrast of the X-ray image caused by the scattered radiation. The grid 18 is attached to the holder 30a such that the X1 direction, being an arrangement direction of the X-ray transmitting portions 32 and the X-ray absorbing portions 33, coincides with a row direction X2 (see FIG. 4) of pixels 45. There are various types of grids 18 within the confines of a grid density, which represents the number of the X-ray absorbing portions 33 per unit width in the arrangement direction (X1 direction) of the X-ray transmitting portions 32 and the X-ray absorbing portions 33, of 26/cm to 100/cm, for example. This embodiment uses a grid having a grid density of 40/cm (4/mm), which is generally used in the X-ray imaging. A grid pitch refers to the sum of the widths (lengths in the X1 direction) of the X-ray transmitting portion 32 and the X-ray absorbing portion 33, and corresponds to an arrangement period F of the X-ray absorbing portions 33. In the case of the grid density of 4/mm, the arrangement period F of the X-ray absorbing portions 33 is 250 μm. As shown in FIG. 4, the electronic cassette 16 contains an antenna 37, a battery 38, and the image detector 35 having the detection panel 35a in the above-described housing. The electronic cassette 16 wirelessly communicates with the console 17 by using the antenna 37 and the battery 38. The antenna 37 transmits and receives a radio wave for the wireless communication to and from the console 17. The battery 38 supplies the electric power to operate each component of the electronic cassette 16. The battery 38 is of relatively small so as to be contained in the slim electronic cassette 16. The battery 38 can be taken out of the electronic cassette 16 and mounted on a specific cradle for recharging. The battery 38 may be recharged by a wireless power feeder. The electronic cassette 16 is provided with a socket 39, in addition to the antenna 37. The socket 39 is used for establishing wired communication with the console 17, in such a case where the wireless communication between the electronic cassette 16 and the console 17 is disabled due to a shortage of the battery 38 or the like. Connecting a cable of the console 17 to the socket 39 enables the wired communication with the console 17. At this time, the console 17 may feed the electric power to the electronic cassette 16 through the cable connected to the socket 39. The antenna 37 and the socket 39 are provided in a communication unit 40. The communication unit 40 mediates the transmission and reception of various types of information and signals including image data among the antenna 37 or the socket 39, a controller 41, and a memory 42. The image detector 35 is composed of the detection panel 35a and circuitry for controlling the operation of the detection panel 35a. The detection panel 35a includes a TFT (thin film transistor) active matrix substrate and the imaging surface 36 provided on the TFT active matrix substrate. The imaging surface 36 has an array of the plurality of pixels 45 for accumulating signal charge in accordance with the amount of the X-rays incident thereon. The plurality of pixels 45 are arrayed into a matrix of n rows (X2 direction)×m columns (Y2 direction) at a predetermined pitch in two dimensions. The detection panel 35a is a square in plane. The size of the imaging surface 36 is 430 mm×430 mm, for example. The pixel number is 2880×2880, for example. The pixel 45 is a square pixel having vertical and horizontal sides of equal length, and has a size of 150 μm×150 μm, for example. The horizontal and the vertical lengths of the pixel 45 correspond to a pixel pitch Δ in the X2 and the Y2 directions, respectively. The detection panel 35a is of an indirect conversion type having a scintillator (phosphor, not shown) for converting the X-rays into visible light. The pixels 45 perform photoelectric conversion of the visible light produced by the scintillator. The scintillator is made of CsI:Tl (thallium activated cesium iodide), GOS (Gd2O2S:Tb terbium activated gadolinium oxysulfide), or the like, and is opposed to the entire imaging surface 36 having the matrix of pixels 45. The scintillator and the TFT active matrix substrate may adopt either a PSS (penetration side sampling) method or an ISS (irradiation side sampling) method. In the PSS method, the scintillator and the substrate are disposed in this order from an X-ray incident side. In the ISS method, the scintillator and the substrate are disposed in reverse order. Note that, a direct conversion type detection panel, which uses a conversion layer (amorphous selenium or the like) for directly converting the X-rays into the electric charge without using the scintillator, may be used instead. The pixel 45 is composed of a photodiode 46 being a photoelectric conversion element for producing the electric charge (electron and hole pairs) upon entry of the visible light, and the TFT 47 being a switching element. The photodiode 46 is composed of a semiconducting layer (of a PIN (p-intrinsic-n) type, for example) for producing the electric charge, and upper and lower electrodes disposed on the top and bottom of the semiconducting layer. The lower electrode of the photodiode 46 is connected to the TFT 47. The upper electrode of the photodiode 46 is connected to a bias line 48. The number of the bias lines 48 coincides with the number of rows (n rows) of the pixels 45 in the imaging surface 36, and all the bias lines 48 are bound into a bus 49. The bus 49 is connected to a bias power source 50. The bias power source 50 applies a bias voltage Vb to the upper electrodes of the photodiodes 46 through the bus 49 and the bias lines 48. Since the application of the bias voltage Vb produces an electric field in the semiconducting layer, the electric charge (electron and hole pairs) produced in the semiconducting layer by the photoelectric conversion is attracted to the upper and lower electrodes, one of which has positive polarity and the other has negative polarity. Thereby, the electric charge is accumulated in the photodiode 46. A gate electrode of the TFT 47 is connected to a scan line 51. A source electrode of the TFT 47 is connected to a signal line 52. A drain electrode of the TFT 47 is connected to the photodiode 46. The scan lines 51 and the signal lines 52 are routed into a lattice. There are provided the scan lines 51 of a number of the rows (n rows) of the pixels 45 in the imaging surface 36 and the signal lines 52 of a number of the columns (m columns) of the pixels 45. The scan lines 51 are connected to a gate driver 53, and the signal lines 52 are connected to a signal processing circuit 54. The circuitry for controlling the operation of the detection panel 35a includes the controller 41, the gate driver 53, the signal processing circuit 54, and the like. The controller 41 makes the detection panel 35a carry out an accumulation operation in which the pixels 45 accumulate the signal charge in accordance with the amount of the X-rays incident thereon, a readout (actual reading) operation in which the signal charge is read out from the pixels 45, and a reset (idle reading) operation, by driving the TFTs 47 through the gate driver 53. In the accumulation operation, while the TFTs 47 are turned off, the pixels 45 accumulate the signal charge. In the readout operation, the gate driver 53 sequentially issues gate pulses G1 to Gn to drive the TFTs 47 of the same row at a time. Thereby, the scan lines 51 are activated one by one to turn on the TFTs 47 connected to the activated scan line 51 on a row-by-row basis. The duration of turn-on is defined by a pulse width of the gate pulse, and the TFT 47 is returned to a turn-off state after a lapse of time defined by the pulse width. Upon turning on the TFT 47, the electric charge accumulated in the photodiode 46 of the pixel 45 is read out to the signal line 52, and inputted to the signal processing circuit 54. The signal processing circuit 54 includes integrating amplifiers 60, a multiplexer (MUX) 61, an A/D converter (A/D) 62, and the like. The integrating amplifier 60 is connected to each signal line 52 on a one-by-one basis. The integrating amplifier 60 is composed of an operational amplifier 60a and a capacitor 60b connected between input and output terminals of the operational amplifier 60a. The signal line 52 is connected to one of the input terminals of the operation amplifier 60a. The other input terminal of the operational amplifier 60a is connected to a ground (GND). A reset switch 60c is connected in parallel with the capacitor 60b. The integrating amplifier 60 integrates the electric charge inputted from the signal line 52. The integrating amplifiers 60 convert the electric charge into voltage signals D1 to Dm, and output the voltage signals D1 to Dm. To the output terminal of the operational amplifier 60a of each column, the MUX 61 is connected through another amplifier 63 and a sample-hold (S/H) circuit 64. The A/D converter 62 is connected to an output of the MUX 61. The MUX 61 sequentially selects one of the plurality of integrating amplifiers 60 connected in parallel, and inputs the voltage signals D1 to Dm outputted from the selected integrating amplifiers 60 in series to the A/D converter 62. The A/D converter 62 converts the inputted voltage signals D1 to Dm into digital data, and outputs the digital data to the memory 42 contained in the electronic cassette 16. Another amplifier may be connected between the MUX 61 and the A/D converter 62. After the MUX 61 reads out the voltage signals D1 to Dm of one row from the integrating amplifiers 60, the controller 41 outputs a reset pulse RST to the integrating amplifiers 60 to turn on the reset switches 60c. Thereby, the signal charge of one row accumulated in the capacitors 60b is discharged and reset. Upon the reset of the integrating amplifiers 60, the gate driver 53 outputs the gate pulse of the next row to start reading out the signal charge from the pixels 45 of the next row. By sequential repetition of this operation, the signal charge is read out from the pixels 45 of every row. After the completion of the readout from every row, the image data representing the X-ray image of one frame is stored to the memory 42. This image data is read out of the memory 42, and outputted to the console 17 through the communication unit 40. Thereby, the X-ray image of the patient is detected. Dark charge occurs in the semiconducting layer of the photodiode 46 irrespective of the presence or absence of entry of the X-rays. Due to the application of the bias voltage Vb, the dark charge is accumulated in the photodiode 46 of the pixel 45. The dark charge occurring in the pixels 45 becomes a noise component of the image data, and therefore the reset operation is carried out to remove this. The reset operation is an operation in which the dark charge produced in the pixels 45 is discharged through the signal lines 52. The reset operation adopts a sequential reset method, for example, by which the pixels 45 are reset on a row-by-row basis. In the sequential reset method, just as with the readout operation of the signal charge, the gate driver 53 sequentially issues the gate pulses G1 to Gn to the scan lines 51 to turn on the TFTs 47 of the pixels 45 on a row-by-row basis. While the TFT 47 is turned on, the dark charge flows from the pixel 45 through the signal line 52 into the capacitor 60b of the integrating amplifier 60. In the reset operation, in contrast to the readout operation, the MUX 61 does not read out the electric charge accumulated in the capacitors 60b. In synchronization with the issue of each of the gate pulses G1 to Gn, the controller 41 outputs the reset pulse RST to turn on the reset switches 60c. Thereby, the electric charge accumulated in the capacitors 60b is discharged, and the integrating amplifiers 60 are reset. Instead of the sequential reset method, a parallel reset method or all pixels reset method may be used. In the parallel reset method, a plurality of rows of pixels are grouped together, and sequential reset is carried out in each group, so as to concurrently discharge the dark charge from the rows of the number of the groups. In the all pixels reset method, the gate pulse is inputted to every row to discharge the dark charge from every pixel at a time. Adoption of the parallel reset method and the all pixels reset method allows speeding up the reset operation. Upon receiving the emission start request signal from the controller 21 of the source control device 14, the controller 41 makes the detection panel 35a carry out the reset operation, and sends the emission permission signal back to the source control device 14. After that, upon receiving the emission start signal, the controller 41 shifts the operation of the detection panel 35a from the reset operation to the accumulation operation. The detection panel 35a is provided with a plurality of detection pixels 65 connected to the signal lines 52 in a short manner without passing through the TFTs 47, besides the normal pixels 45 connected to the signal lines 52 through the TFTs 47 as described above, in the same imaging surface 36. The detection pixel 65 functions as a dose detection sensor for detecting the dose of the X-rays incident upon the imaging surface 36 through the patient M. In this embodiment, one detection pixel 65 composes one dose detection sensor. The detection pixels 65 occupy an order of several % of the pixels 45 in the imaging surface 36. In the detection pixel 65 according to this embodiment, the photodiode 46 and the like have exactly the same fundamental structure as those of the pixel 45. Therefore, the detection pixels 65 and the pixels 45 can be formed in approximately the same manufacturing process. Accordingly, the size of one detection pixel 65 is 150 μm×150 μm, just as with the size of the pixel 45. Since the detection pixel 65 is connected to the signal line 52 directly without passing through the TFT 47, the signal charge produced in the detection pixel 65 immediately flows into the signal line 52, irrespective of the turn on and off of the TFT 47. The same goes if the normal pixels 45 in the same row have the TFTs 47 turned off and are in the accumulation operation for accumulating the signal charge. Thus, the electric charge produced in the photodiode 46 of the detection pixel 65 always flows into the capacitor 60b of the integrating amplifier 60 in the signal line 52 connected to the detection pixel 65. During the accumulation operation of the detection panel 35a, the electric charge from the detection pixel 65 is accumulated in the capacitor 60b, and outputted through the MUX 61 to the A/D converter 62 as a voltage value at a predetermined sampling period. The A/D converter 62 outputs the voltage value to the memory 42 as a dose detection signal of each detection pixel 65. The dose detection signal represents the dose of the X-rays applied per unit of time. The dose detection signal outputted at the predetermined sampling period is sequentially outputted to the memory 42. As shown in FIG. 5, the detection pixels 65 are arrayed at constant arrangement periods S1 and S2 in the X2 and Y2 directions so as to be uniformly distributed over the imaging surface 36, without being localized in the imaging surface 36. The positions of the detection pixels 65 are already known in manufacturing the detection panel 35a, and an internal memory (not shown) of the controller 41 stores in advance coordinate information representing the position of each detection pixel 65 in the imaging surface 36. The dose detection signal of each detection pixel 65 outputted from the A/D converter 62 is recorded to the memory 42 in correspondence with the coordinate information. The operation of an AEC section (automatic exposure control section) 67 is controlled by the controller 41. The AEC section 67 reads out the dose detection signal of each detection pixel 65 from the memory 42, and carries out AEC based on the read dose detection signal. As shown in FIG. 6, the AEC section 67 has a measurement area selection circuit 75, an integration circuit 76, a comparison circuit 77, and a threshold value generation circuit 78. The measurement area selection circuit 75 chooses one or more of the plurality of detection pixels 65 distributed over the imaging surface 36 whose dose detection signals are to be used in the AEC, based on the information of a measurement area set in the imaging condition. The integration circuit 76 calculates an average of the outputted values of the dose detection signals from the detection pixels 65 chosen by the measurement area selection circuit 75. For example, in a case where areas Aa and Ab represented by chain double-dashed lines in FIG. 5 are chosen as the measurement areas, the integration circuit 76 calculates an average of output values of dose detection signals of nine detection pixels 65a to 65i in each of the measurement areas Aa and Ab. Otherwise, the integration circuit 76 calculates an average of output values of eighteen detection pixels 65 in total in the measurement areas Aa and Ab. The average is calculated whenever the dose detection signals are sampled. Then, the integration circuit 76 obtains an integral value by integrating the average. The integral value represents the integral dose of the applied X-rays. The comparison circuit 77 compares the integral value of the dose detection signals from the integration circuit 76 with an emission stop threshold value applied from the threshold value generation circuit 78. Upon the integral value reaching the threshold value, the comparison circuit 77 outputs the emission stop signal. The communication unit 40 is provided with an emission signal I/F 80, in addition to the antenna 37 and the socket 39 described above. To the emission signal I/F 80, the emission signal I/F 25 of the source control device 14 is connected. The emission signal I/F 80 performs reception of the emission start request signal, transmission of the emission permission signal in response to the emission start request signal, reception of the emission start signal, and transmission of the emission stop signal outputted from the comparison circuit 77. As shown in FIG. 7, in grid imaging using the grid 18, the grid 18 is disposed in front (on the X-ray incident side) of the imaging surface 36, so that the grid 18 is overlaid on the detection pixels 65. The arrangement period S1 of the detection pixels 65 in the X2 direction is different from the arrangement period F of the X-ray absorbing portions 33 of the grid 18 (S1≠F). Also, the arrangement period S1 of the detection pixels 65 is not an integral multiple of the arrangement period F of the X-ray absorbing portions 33. In other words, “arrangement period S1≠N·arrangement period F” (N is an integer) holds true. In this embodiment, as shown in FIG. 8, the detection pixels 65 are arranged every three pixels with leaving space of two pixels 45 between the two detection pixels 65. In the case of arranging the detection pixels 65 every three pixels, the arrangement period S1 of the detection pixels 65 has a length of three pixels. Since the pixel pitch Δ=150 μm, the arrangement period S1 is 150 μm×3=450 μm. On the other hand, since the grid density of the grid 18 is 4/mm in this embodiment, the arrangement period F is 250 μm. Accordingly, the arrangement period S1 (450 μm) does not coincide with the arrangement period F (250 μm). Also, dividing the arrangement period S1 by the arrangement period F results in 450/250=1.8, so the arrangement period S1 is not an integral multiple of the arrangement period F. As described above, since “arrangement period S1≠N·arrangement period F” (N is an integer) holds true, as shown in FIG. 8, not all the detection pixels 65 are disposed behind the X-ray transmitting portions 32 or the X-ray absorbing portions 33. There are a detection pixel 65J situated behind the X-ray transmitting portion 32, a detection pixel 65K situated behind the entire X-ray absorbing portion 33, and a detection pixel 65L situated behind a part, not entire, of the X-ray absorbing portion 33. Therefore, not all of the plurality of detection pixels 65 has maximum output values or minimum output values. Taking FIG. 8 as an example, the dose detection signal of the detection pixel 65J has a maximum output value, and the dose detection signal of the detection pixel 65K has a minimum output value. The dose detection signal of the detection pixel 65L takes an output value between the maximum output value and the minimum output value. As described above, the output value of the dose detection signal of each detection pixel 65 is distributed within the confines between the maximum output value and the minimum output value. Also, as shown in FIG. 9, if the geometrical disposition between the grid 18 and the detection pixels 65 is shifted from a state shown in FIG. 8 by a size of one pixel 45 due to an attachment backlash or a manufacturing error of the grid 18, the detection pixels 65J to 65L exist. Thus, the output value of each detection pixel 65 is distributed in the confines between the maximum output value and the minimum output value. Also in this case, not all the detection pixels 65 have the maximum output values or the minimum output values. Since the output values of the plurality of detection pixels 65 are distributed in the confines between the maximum output value and the minimum output value, a variation range of the average of the output values is smaller than the difference between the maximum output value and the minimum output value. As described above, the AEC section 67 uses the average of the output values of the detection pixels 65 in the measurement areas Aa and Ab shown in FIG. 5 for judgment of the AEC. Reduction in the variation range of the average translates into stability of the output value in accordance with the dose incident upon the measurement areas Aa and Ab, even in a case where the geometrical disposition between the grid 18 and the detection pixels 65 is misaligned. Therefore, the stable AEC can be carried out without being affected by misalignment in the geometrical disposition between the grid 18 and the detection pixels 65. On the other hand, in a comparative example shown in FIGS. 10 and 11, an arrangement period S1a=the arrangement period F holds true. In this case, the variation range of the output value of each detection pixel 65 is a maximum output difference being the difference between the maximum output value and the minimum output value. In FIGS. 10 and 11, a letter symbol Δa represents a pixel pitch of the pixels 45. The detection pixels 65 are disposed every other pixel, and the arrangement period S1a of the detection pixels 65 is a length of the two pixels 45. FIG. 11 shows a case in which the geometrical disposition between the grid 18 and the detection pixels 65 is shifted from a state shown in FIG. 10 by a length of one pixel 45. If the arrangement period S1a=the arrangement period F, there are cases in which only the detection pixels 65K are present each of which is disposed behind the entire X-ray absorbing portion 33 as shown in FIG. 10, or only the detection pixels 65J are present each of which is disposed behind the X-ray transmitting portion 32 as shown in FIG. 11. Thus, the output value of each detection pixel 65 is not distributed, and every detection pixel 65 has the same output value. Only the detection pixels 65K are present in FIG. 10, so every detection pixel 65 outputs the minimum output value. Only the detection pixels 65J are present in FIG. 11, on the contrary, so every detection pixel 65 outputs the maximum output value. A shift of the geometrical disposition between the grid 18 and the detection pixels 65 by a length of just one pixel 45 varies the output value of each detection pixel 65 from the minimum output value to the maximum output value. In the comparative example, since every detection pixel 65 takes the same output value, the average of the output values of the detection pixels 65 is equal to the output value of the one detection pixel 65. Accordingly, the variation range of the average is the maximum output difference, being the difference between the maximum output value and the minimum output value. The large variation range of the average hinders the stable AEC, because the misalignment in the geometrical disposition between the grid 18 and the detection pixels 65 affects severely. FIG. 12 shows a comparative example in which the arrangement period S1a of FIGS. 10 and 11 is modified to a length three times as large as the arrangement period F (S1a=F×3). Also in a case where the arrangement period S1a is an integral multiple of the arrangement period F, just as with FIGS. 10 and 11, the detection pixels 65 are unified into one type of the detection pixels 65J to 65L (detection pixels 65K are shown as an example in FIG. 12), and the plurality of detection pixels 65 has the same output values. Therefore, just as with the comparative example shown in FIGS. 10 and 11, the variation range of the output value of each detection pixel 65 is the maximum output difference, and hence the stable AEC cannot be carried out. The pixel 45 is a square pixel, and the pixel pitch of the pixels 45 in the Y2 direction is A, which is the same as the pixel pitch in the X2 direction. The arrangement period S2 of the detection pixels 65 in the Y2 direction is the same as the arrangement period S1. Therefore, in a case where the arrangement period S1 is different from the arrangement period F and not an integral multiple of the arrangement period F, the arrangement period S2 is different from the arrangement period F too and not an integral multiple of the arrangement period F neither. In other words, as for the arrangement period S2, “the arrangement period S2≠N·arrangement period F” (N is an integer) holds true just as with the arrangement period S1. Thus, as shown in FIG. 13, in the case of using the grid 18 in a state of being turned 90° from a state shown in FIG. 7, the average of the output values of the plurality of detection pixels 65 has a smaller variation range (smaller than the maximum output difference), as compared with the comparative examples shown in FIGS. 9 to 11, so that the stable AEC can be carried out. As shown in FIG. 14, the console 17 is composed of a console body 17a, a display 17b, and a keyboard 17c. The console 17 is wiredly or wirelessly connected to the electronic cassette 16 in a communicatable manner, and controls the operation of the electronic cassette 16. To be more specific, the console 17 not only transmits the imaging condition to the electronic cassette 16 to set conditions (gain of the amplifier for amplifying a voltage corresponding to the accumulated signal charge, and the like) for the AEC and signal processing of the signal processing circuit 54, but also controls the turn on and off of the electronic cassette 16, mode switching to a power saving mode or an imaging ready state, and the like. The console 17 applies various types of image processing such as an offset correction, a gain correction, and a defect correction to the X-ray image data transmitted from the electronic cassette 16. In the defect correction, pixel values of the column having the detection pixel 65 are interpolated with pixel values of the adjoining columns without having the detection pixel 65. The X-ray image after being subjected to the image processing is displayed on the display 17b, and its data is recorded to data storage such as a storage device 87 or a memory 86 contained in the console body 17a, or an image storage server connected to the console 17 through a network. Note that, the electronic cassette 16 may perform each type of the above-described image processing. The console 17 receives an input of an examination order, which includes information about the sex and the age of a patient, a body part to be imaged, and a purpose of imaging, and displays the examination order on the display 17b. The examination order is inputted from an external system, e.g. an HIS (hospital information system) or an RIS (radiography information system), that manages patient data and examination data related to radiography, or inputted manually by the operator. The examination order includes the body part to be imaged e.g. head, chest, abdomen, or the like, and an imaging direction e.g. anterior, medial, diagonal, PA (X-rays are applied from a posterior direction), or AP (X-rays are applied from an anterior direction). The operator confirms the contents of the examination order on the display 17b, and inputs the imaging condition corresponding to the contents through an operation screen of the console 17. As shown in FIG. 15, imaging conditions, which are different from one body part to another, can be set in the console 17. The imaging condition includes the tube voltage, the tube current, the measurement area of the detection pixels 65, the emission stop threshold value used for comparison with the integral value of the dose detection signals of the detection pixels 65 to judge the stop of the X-ray emission, and the like. This information about the imaging conditions is stored to the storage device 87. The same imaging condition is set manually by the operator to the source control device 14 with referring to the imaging condition of the console 17. The measurement area, which represents an area of the detection pixels 65 to be used in the AEC, corresponds to a region of interest to be most noticed in a diagnosis in each body part, and is set at an area from which the dose detection signals are stably obtained. In a case where the body part to be imaged is a chest, for example, portions of lung fields are assigned as the measurement areas, shown as the measurement areas Aa and Ab enclosed by the chain double-dashed lines in FIG. 5. The measurement area is represented by X and Y coordinates. In the case of a rectangular measurement area, like this embodiment, the X and Y coordinates of two points connected by a diagonal line are stored. The X and Y coordinates correspond to the positions of the pixels 45 (including the detection pixels 65) in the imaging surface 36 of the electronic cassette 16. The X and Y coordinates are represented under the condition that an X axis extends in a direction parallel to the scan lines 51, and a Y axis extends in a direction parallel to the signal lines 52, and the coordinates of the most upper left pixel 45 are assigned as an origin point (0, 0). As shown in FIG. 14, the console body 17a is provided with a CPU 85, the memory 86, the storage device 87, a communication I/F 88, and an input and output I/F 89. These components are connected each other through a data bus 90. The display 17b and the keyboard 17c are connected to the console body 17a through the input and output I/F 89. Note that, a mouse, a touch panel, and the like may be used instead of the keyboard 17c. The storage device 87 is a hard disk drive (HDD), for example. The storage device 87 stores a control program and an application program (hereinafter called AP) 92. The AP 92 is a program that makes the console 17 perform various functions related to radiography, such as display processing of the examination order and the X-ray image, image processing of the X-ray image, and a setup of the imaging condition. The memory 86 is a work memory that the CPU 85 uses in executing processing. The CPU 85 loads the control program stored on the storage device 87 into the memory 86, and runs the program for centralized control of each part of a computer. The communication I/F 88 is a network interface for performing wireless or wired transmission control from/to an external device such as the RIS, the HIS, the image storage server, and the electronic cassette 16. As shown in FIG. 16, by running the AP 92, the CPU 85 of the console 17 functions as a storing and retrieving processing unit 95, an input and output controller 96, and a main controller 97. The storing and retrieving processing unit 95 performs storing processing of various types of data to the storage device 87, and retrieval processing from the various types of data stored in the storage device 87. The input and output controller 96 reads out drawing data corresponding to an operation on the keyboard 17c from the storage device 87 through the input and output I/F 89, and outputs to the display 17b various types of operation screens of GUIs based on the read drawing data. The input and output controller 96 receives input of operation commands from the keyboard 17c through the operation screens. The main controller 97 includes a cassette controller 98 for controlling the operation of the electronic cassette 16, and performs centralized control of the operation of each part of the console 17. The cassette controller 98 receives the information about the measurement area and the information about the emission stop threshold value according to the imaging condition from the storing and retrieving processing unit 95, and supplies the electronic cassette 16 with the information. Note that, in the console 17, an image processor for performing various types of imaging processing such as the offset correction, the gain correction, and the defect correction described above and a communicator for mediating communication with the source control device 14 and the electronic cassette 16 are established in the CPU 85, in addition to the components described above. Note that, each component may be established by specific hardware, instead of actualizing the function of each component by software, as with this embodiment. The electronic cassette 16 may perform all or a part of the image processing including the offset correction, the gain correction, the defect correction, and the like. Next, an X-ray imaging procedure by the X-ray imaging system 10 will be described with referring to a flowchart of FIG. 17. Firstly, while the patient M stands in a predetermined position in front of the imaging stand 30, the height and the horizontal position of the electronic cassette 16 set in the imaging stand 30 are adjusted for positioning with the patient's body part to be imaged. In accordance with the position of the electronic cassette 16 and the size of the body part to be imaged, the height and the horizontal position of the X-ray source 13 and the size of the irradiation field are adjusted. Then, the electronic cassette 16 is turned on. The imaging condition is inputted with the keyboard 17c. The imaging condition and the measurement area, the emission stop threshold value, and the like according to the imaging condition are outputted through the cassette controller 98 to the electronic cassette 16. In a like manner, the imaging condition is also set in the source control device 14. After the completion of preparation for imaging, the operator performs the first-step push of the emission switch 15. Thereby, the warm-up start signal is transmitted to the source control device 14 to start warming up the X-ray source 13. After a lapse of predetermined time, upon the second-step push of the emission switch 15, the emission start signal is transmitted to the source control device 14 to start the X-ray emission (S10). The X-rays radiating from the X-ray source 13 produces the scattered radiation in passing through the patient M. This scattered radiation is removed by the grid 18. Before the start of the X-ray emission, the detection panel 35a carries out the reset operation. Upon receiving the emission start signal from the source control device 14, the reset operation is shifted to the accumulation operation. In parallel with the accumulation operation of the detection panel 35a, the AEC section 67 performs the AEC based on the dose detection signals of the detection pixels 65 in the electronic cassette 16. The measurement area selection circuit 75 chooses the dose detection signals outputted from the detection pixels 65 existing in the measurement areas based on the information about the measurement areas supplied by the console 17, out of the dose detection signals of the plurality of detection pixels 65 inputted from the A/D converter 62, and outputs the chosen dose detection signals to the integration circuit 76 (S11). The integration circuit 76 integrates the average of the output values of the dose detection signals. The relation between the arrangement period S1 of the detection pixels 65 in the X2 direction and the arrangement period F and between the arrangement period S2 of the detection pixels 65 in the Y2 direction and the arrangement period F satisfies “arrangement period S1, S2≠N·arrangement period F” (N is an integer). Thus, the output value of the dose detection signal of each detection pixel 65 is distributed. Accordingly, if the geometrical disposition between the grid 18 and the detection pixels 65 is misaligned, the average of the output values of the detection pixels 65 has the small variation range, and therefore the stable AEC can be carried out without being affected by the geometrical disposition between the grid and the detection pixels 65. The threshold value generation circuit 78 produces the emission stop threshold value provided by the cassette controller 98, and outputs the emission stop threshold value to the comparison circuit 77. The comparison circuit 77 compares the integral value of the dose detection signals integrated by the integration circuit 76 with the emission stop threshold value (S13). In a case where the integral value reaches the emission stop threshold value (YES in S14), the emission stop signal is outputted. The emission stop signal outputted from the comparison circuit 77 is transmitted through the emission signal I/F 80 to the emission signal I/F 25 of the source control device 14 (S15). Upon receiving the emission stop signal by the emission signal I/F 25, the controller 21 stops the electric power supply from the high voltage generator 20 to the X-ray source 13 in the source control device 14, and therefore the X-ray emission is stopped (S16). At the instant when an emission stop detecting circuit of the AEC section 67 detects the stop of the X-ray emission, the detection panel 35a stops the accumulation operation and shifts to the readout operation, so that the X-ray image is outputted to the memory 42. After the readout operation, the detection panel 35a restarts the reset operation. The X-ray image is transmitted through the communication unit 40 to the console 17. The X-ray image is subjected to the various types of image processing, and displayed on the display 17b by the input and output controller 96 (S17). According to the present invention, the variation range of the output value of the detection pixel 65 is reduced by determining the arrangement periods S1 and S2 of the detection pixels 65 in relation to the arrangement period F of the grid 18. Therefore, in contrast to the U.S. Pat. No. 6,944,266, it is not necessary to obtain the gain image whenever the imaging is performed, and correct the output value of each detection pixel based on the obtained gain image. In the case of the U.S. Pat. No. 6,944,266, if the geometrical disposition between the grid 18 and the detection pixels 65 is misaligned by an impact or the like after the obtainment of the gain image, the AEC cannot be performed appropriately due to an improper correction of the output values. However, according to the present invention, the AEC can be appropriately carried out if the geometrical disposition is misaligned. According to the above embodiment, one detection sensor is composed of the one detection pixel 65 of the same size as the pixel 45. Thus, in contrast to the U.S. Pat. No. 6,952,465 having the striped dose detection sensors of 500 pixels, no density step that is visible to a human eye occurs in the X-ray image, and there is little fear of degradation in the image quality of the X-ray image. Also, the small-sized detection pixels 65 facilitate the defect correction. Note that, the detection pixel 65 is treated as the defect pixel, and interpolated with the pixel values of the pixels 45 in the vicinity thereof. In this interpolation (defect correction), correction accuracy is increased with reduction in size of the detection pixel 65, so the smaller the size of the detection pixel 65 the better in terms of the image quality. On the other hand, the smaller the size of the detection pixel 65, the severer the effect of the positions of the X-ray absorbing portions 33 of the grid 18 becomes. In other words, focusing attention on the one detection pixel 65, there is a demerit that the misalignment in the geometrical disposition with the grid 18 increases the variation range of the output value. However, according to the present invention, by making the arrangement period S1 differ from the arrangement period F, the output values of the plurality of detection pixels 65 can be distributed even if the output value of each individual detection pixel 65 has the large variation range. Furthermore, not all the detection pixels 65 have a maximum output value Dmax or a minimum output value Dmin. In the AEC, the output values of the plurality of detection pixels 65 are averaged. Therefore, if reduction in size of each detection pixel 65 increases the variation range of each individual output value, the stable AEC can be carried out. Moreover, since the detection pixels 65 are arrayed at the constant period, an algorithm for defect correction processing is easily simplified as compared with the case of an aperiodic arrangement. Also, the detection pixels 65 are easily formed in manufacturing. This brings about the merit of reduced manufacturing costs. Also, the arrangement periods S1 and S2 of the detection pixels 65 in the X2 and Y2 directions are equalized, and both the relation between the arrangement period S1 and the arrangement period F and between the arrangement period S2 and the arrangement period F satisfies “arrangement period S1, S2≠N·arrangement period F” (N is an integer). Thus, the stable AEC can be carried out in either of cases where the grid 18 is used in such a position that the arrangement direction X1 of the grid 18 coincides with the X2 direction of the imaging surface 36 and where the grid 18 is used in such a position that the arrangement direction X1 of the grid 18 coincides with the Y2 direction of the imaging surface 36. In a case where the electronic cassette 16 is in the shape of a square in plane, just as with this embodiment, it is difficult to recognize at sight whether the electronic cassette 16 is in a vertical position in which the X2 direction is in parallel with a horizontal direction or a horizontal position in which the Y2 direction is in parallel with the horizontal direction. Applying the striped dose detection sensors described in the U.S. Pat. No. 6,952,465 to such a square electronic cassette 16 and grid 18 impairs usability, because it is necessary to carefully confirm that the stripe direction of the dose detection sensors is not in parallel with the stripe direction of the grid 18. Especially, since the position cannot be confirmed in a state of setting the electronic cassette 16 in the holder 30a, the electronic cassette 16 has to be taken out of the holder 30a and this further impairs the usability. However, according to this embodiment, making neither of the arrangement periods S1 and S2 of the detection pixels 65 in the X2 and Y2 directions coincide with the arrangement period F improves convenience, because of eliminating the need for carefully confirming the position of the electronic cassette 16 and the position of the grid 18. Note that, the arrangement periods S1 and S2 of the detection pixels 65 are not necessarily the same. This is because as long as each of the arrangement periods S1 and S2 is different from the arrangement period F, the average of the detection pixels 65 has the small variation range and the stable AEC can be carried out. However, it is preferable that the arrangement period S2 and the arrangement period S1 be the same. This is because whether or not the stable AEC can be carried out depends on the relation between the arrangement period S1, S2 and the arrangement period F. Thus, depending on the type (grid density) of the grid 18, the arrangement period S1, S2 may possibly coincide with the arrangement period F and the stable AEC may not be carried out with the grid 18. Therefore, in performing grid imaging with the electronic cassette 16, it is necessary to investigate on a grid-type by grid-type basis whether or not the stable AEC can be carried out by combination with the electronic cassette 16. Such an investigation operation is performed based on the arrangement periods S1 and S2 and the grid density of the grid 18. Equalizing the arrangement periods S1 and S2 eliminates the need for performing the investigation as to each of the arrangement periods S1 and S2, and hence improves convenience. Also, if the arrangement periods S1 and S2 are different from each other, whether or not the stable AEC can be carried out may depend on the position of the electronic cassette 16 even with the use of the same grid 18. Therefore, the usability deteriorates as compared with the case of equalizing the arrangement periods S1 and S2. Also, if the arrangement periods S1 and S2 are different from each other, for example, it is conceivable that the number of the detection pixels 65 included in each of the measurement areas Aa and Ab of the same size may vary depending on the vertical and horizontal position of the electronic cassette 16. In this case, it becomes necessary to change an algorithm for calculating the integral value by using the detection pixels 65 in accordance with the orientation of the electronic cassette 16. Equalizing the arrangement periods S1 and S2 can share the algorithm, because the number of the detection pixels 65 included in the measurement area is invariable if the size of the measurement area is the same. For these reasons, the arrangement periods S1 and S2 are preferably equal to each other. The electronic cassette 16 and the grid 18 in the shape of a square in plane are described in the above embodiment, but the electronic cassette and the grid may be in the shape of a rectangle in plane. As the electronic cassette in the shape of a rectangle in plane, for example, there is an electronic cassette of a size compatible with the International Standard ISO 4090:2001, just as with a film cassette and an IP (imaging plate) cassette. Also in the case of the electronic cassette in the shape of a rectangle in plane, it is preferable that the arrangement periods S1 and S2 be equalized. The rectangular electronic cassette is sometimes used in such a manner that in imaging a chest of a patient of typical physique, the electronic cassette is disposed such that a longitudinal direction of the electronic cassette is along a height direction of the patient, while in imaging a chest of a patient of corpulent physique, the electronic cassette is disposed in a state of being turned 90° such that the longitudinal direction is along a width direction of the patient's body. If the striped dose detection sensors according to the U.S. Pat. No. 6,952,465 are used in this rectangular electronic cassette, the 90° turn of the electronic cassette brings about the coincidence between the stripe direction of the dose detection sensors and the stripe direction of the grid, even though the stripe direction of the dose detection sensors is orthogonal to the stripe direction of the grid in imaging of the patient of the typical physique. However, making the arrangement period S1, S2 of the detection pixels 65 in the X2 or Y2 direction differ from the arrangement period F, as described in this embodiment, prevents the occurrence of this problem. In a second embodiment shown in FIGS. 18 and 19, each of the arrangement period S1 and the arrangement period F is represented as a length in unit of the number of the pixels 45. In FIG. 18, the arrangement period S1 is 7 because there is space of six pixels 45 between the two detection pixels 65. The arrangement period F has a length of four pixels 45, so a conversion value, being the number of pixels 45 into which the arrangement period F is converted, is 4. In an example shown in FIG. 19, the arrangement period S1 is 6, and the arrangement period F is 4. Both of FIGS. 18 and 19 satisfy the relation of arrangement period S1≠the arrangement period F, and hence represent embodiments included in the present invention. However, comparing between FIGS. 18 and 19, an example of FIG. 18 is more preferable. The reason is as follows. In the example of FIG. 18, since the arrangement period S1 of the detection pixels 65 is 7 and the arrangement period F is 4, the arrangement period S1 and the arrangement period F are co-prime numbers. On the contrary, in the example of FIG. 19, since the arrangement period S1 of the detection pixels 65 is 6 and the arrangement period F is 4, the arrangement period S1 and the arrangement period F are not co-prime numbers. As shown in FIG. 8, how the X-ray absorbing portions 33 overlap with the plurality of detection pixels 65 varies in accordance with the position of each detection pixel 65, such that one detection pixel 65 has a large overlap amount with the X-ray absorbing portion 33, while anther detection pixel 65 has a small overlap amount. However, since both of the X-ray absorbing portions 33 and the detection pixels 65 are arrayed periodically, an overlap state is in cycles with an overlap period C. The overlap period C is a least common multiple of the arrangement period S1 and the arrangement period F. The number of the detection pixels 65 included in the overlap period Cis increased with the length of the overlap period C. The larger the number of the detection pixels 65 included in the overlap period C, the more the output value of each detection pixel 65 is distributed. Therefore, the output value of each detection pixel 65 is leveled, and the stable AEC can be carried out. In the case of the example of FIG. 19, the overlap period C is the least common multiple (12) of the arrangement period S1 (6) and the arrangement period (4). In the example of FIG. 19, since the arrangement period S1 (6) and the arrangement period F (4) are not co-prime numbers, the overlap period C is less than the product (4×6=24) of the arrangement period S1 (6) and the arrangement period (4). In the case of the example of FIG. 19, the number of the detection pixels 65 included in the overlap period C is 2, given by dividing the least common multiple (12) by a pixel period (6) of the arrangement period S1. On the other hand, in the case of the example of FIG. 18, since the arrangement period S1 (7) and the arrangement period F (4) are co-prime numbers, the least common multiple is 7×4=28, and hence the overlap period C is 28. Thus, the overlap period C is equal to the product (7×4=28) of the arrangement period S1 (7) and the arrangement period F (4). In the example of FIG. 18, the number of the detection pixels 65 included in the overlap period C is 4, given by dividing the least common multiple (28) by a pixel period (7) of the arrangement period S1. Comparing the examples of FIGS. 18 and 19, since the overlap period C (28) of the co-prime numbers shown in FIG. 18 is longer than the overlap period (12) of the not co-prime numbers shown in FIG. 19, the number (4) of the detection pixels 65 included in the overlap period C of the example of FIG. 18 is larger than the number (2) of the detection pixels 65 included in the overlap period C of the example of FIG. 2. The overlap period C is repeated, so the larger the number of the detection pixels 65 included in the overlap period C, the more the output value of the detection pixel 65 is distributed. Thus, the variation range of the average of the output values of the detection pixels 65 is easily reduced. Accordingly, it is preferable that the arrangement period S1 and the arrangement period Fare co-prime numbers, just as in the case of FIG. 18. Other embodiments of the detection panel 35a will be hereinafter described. In each embodiment, the same reference numerals as those of the first and second embodiments indicate the same components as those of the first and second embodiments, and detailed description thereof will be omitted. In a third embodiment shown in FIGS. 20 to 22, an array of the detection pixels 65 in which the average of the dose detection signals of the detection pixels 65 is calculated is grouped into one set 200. The sets 200 are periodically arranged in the same or different rows, so as to arrange the sets 200 over the entire imaging surface 36 in a distributed manner. The set 200 is a minimum unit of a group of the detection pixels 65 in which the AEC section 67 calculates the average of the dose detection signals. Note that, the sets 200 may be arranged intensively in a region corresponding to the measurement area set in advance, such as left and right lung fields, for example, instead of being arranged over the entire imaging surface 36. A set 200a shown in FIG. 20A is an example in which a plurality of detection pixels 65 (four detection pixels 65 in this example) is arrayed in one row extending in the X2 direction at an arrangement period S1=5, and the sets 200a are regularly arranged at equal intervals, for example. The set 200a is a minimum unit of a group of the detection pixels 65 in which the AEC section 67 calculates the average of the dose detection signals. Accordingly, a block 201a that is formed with the two sets 200a and has the eight detection pixels 65 as shown in FIG. 20B, or a block 201b that is formed with three sets 200a and has the twelve detection pixels 65 as shown in FIG. 20C may be established as a minimum unit for calculating the average. The intervals between the sets 200a or between the blocks 201a or 201b may be irregular. Also, as shown in FIG. 21, a set 200b in which the detection pixels 65 are arrayed in the Y2 direction may be used. In the set 200b, each of the pixel pitch Δ and the arrangement period S2 of the detection pixels 65 is a length in the Y2 direction. Also, the sets 200a each having the detection pixels 65 arrayed in the X2 direction as shown in FIG. 20A and the sets 200b each having the detection pixels 65 arrayed in the Y2 direction may be mixed in the imaging surface 36. Selectively using either the sets 200a or the sets 200b in accordance with the attachment position of the grid 18 makes it possible to carry out the stable AEC, irrespective of the attachment position. The set 200a shown in FIG. 20A or the set 200b shown in FIG. 21 is a set that is composed of a plurality of detection pixels 65 arrayed in one row or one column. However, as a set 200c shown in FIG. 22, a set may be composed of a plurality of the detection pixels 65 arranged with being shifted in the X2 and Y2 directions. In the set 200c shown in FIG. 22, the plurality of detection pixels 65 is arrange in the different rows, and as for the X2 direction at an arrangement period S1=5 with leaving space of four columns. Although the detection pixels 65 are situated in the different rows, the output value of each detection pixel 65 is distributed as long as the arrangement period 51 in the X2 direction is different from the arrangement period F, so the average has a reduced variation range. As described above, even in a case where the detection pixels 65 are situated in the different rows, the arrangement period S1 of the detection pixels 65 can be obtained just as in the case of arranging a plurality of detection pixels 65 in one row, and the arrangement period S1 corresponds to a length in the X2 direction (row direction). As for the X2 direction, the pixel pitch ΔX and the arrangement period S1 of the detection pixels 65 in the set 200c coincide with the pixel pitch Δ and the arrangement period S1 of the detection pixels 65 in the set 200a shown in FIG. 20A. Thus, the average of the output values of a group of the detection pixels 65 in the set 200c is approximately equal to the average of the output values of the group of the detection pixels 65 in the set 200a, and therefore the set 200c may be substituted for the set 200a. Also, as for the Y2 direction, the plurality of detection pixels 65 in the set 200c is arranged at an arrangement period S2=5, though being situated in the different columns. Although the detection pixels 65 are situated in the different columns, the output value of each detection pixel 65 is distributed as long as the arrangement period S2 in the Y2 direction is different from the arrangement period F, so the average has a reduced variation range. As for the Y2 direction, the pixel pitch ΔY and the arrangement period S2 of the detection pixels 65 in the set 200c coincide with the pixel pitch Δ and the arrangement period S2 of the detection pixels 65 in the set 200b shown in FIG. 21. Thus, the set 200c may be substituted for the set 200b. As described above, even in a case where the detection pixels 65 are situated in the different columns, the arrangement period S2 of the detection pixels 65 can be obtained just as in the case of arranging a plurality of the detection pixels 65 in one column, and the arrangement period S2 corresponds to a length in the Y2 direction (column direction). The set 200c is usable instead of both the set 200a shown in FIG. 20A and the set 200b shown in FIG. 21. Thus, providing the sets 200c allows carrying out the stable AEC irrespective of the attachment position of the grid 18, just as in the case of providing the sets 200a and the sets 200b in a mixed manner. Furthermore, in the case of mixing the sets 200a and 200b, it is necessary to selectively use the sets 200a or 200b in accordance with the attachment position of the grid 18. However, using the sets 200c eliminates the need for the selective use in accordance with the attachment position of the grid 18. Also, the use of the sets 200c can reduce the number of the detection pixels 65 in half, as compared with the case of mixing the sets 200a and 200b. As the detection pixel 65 of this example, a short circuit between the photodiode 46 and the signal line 52 causes a continuous flow of the electric charge of the detection pixel 65 through the signal line 52. Thus, even if the detection pixels 65 are situated in the different rows, the electric charge of the detection pixels 65 flows into the integrating amplifiers 60 of the signal processing circuit 54 at approximately the same time. Therefore, there is a merit that the dose detection signals of the detection pixels 65 in the set 200c can be read out at the same time. Note that, in the set 200c of this example, a shift amount (five pixels) of the detection pixels 65 is the same in the X2 and Y2 directions, but may be different between the X2 direction and the Y2 direction. In each of the above embodiments, one detection pixel composes one dose detection sensor. However, one dose detection sensor may be composed of a detection pixel group 66 including a plurality of adjoining detection pixels 65, as shown in FIG. 23. In a case where the dose detection sensor is composed of the detection pixel group 66, the arrangement period S1 corresponds to an arrangement period of the plurality of detection pixel groups 66 that is periodically arranged with leaving space. FIG. 23 shows a state of viewing the grid 18 and the imaging surface 36 of the detection panel 35a from a side (Y1 and Y2 directions), just as with FIG. 8 and the like. The individual detection pixel 65 is used as the dose detection sensor in an imaging surface 36a, while the detection pixel group 66 including the two adjoining detection pixels 65 is used as the dose detection sensor in an imaging surface 36b. The detection pixels 65 and the X-ray absorbing portions 33 are hatched for the purpose of distinction. For example, the width of the X-ray transmitting portion 32 of the grid 18 in the X1 direction is 200 μm, and the width of the X-ray absorbing portion 33 thereof is 50 μm. The width of the detection pixel 65 in the same direction is 150 μm. The ratio of the X-ray transmitting portion 32 to an area of the detection pixel 65 in the imaging surface 36a is in a range of 2/3 to 1. The ratio of the X-ray transmitting portion 32 to an area of the detection pixel group 66 in the imaging surface 36b is in a range of 4/6 to 5/6. According to the area ratios, provided that the X-rays of the same dose is detected, the maximum output difference between the maximum output value and the minimum output value of the detection pixel 65 in the imaging surface 36a is 1÷2/3=1.5, while the maximum output difference of the detection pixel group 66 in the imaging surface 36b is 5/6÷4/6=1.25. As described above, using the detection pixel group 66 as the dose detection sensor reduces the maximum output difference of the dose detection signal, as compared with the case of using the one detection pixel 65 as the one dose detection sensor. The smaller the maximum output difference, the smaller the variation range of the output value becomes. Accordingly, if the geometrical disposition between the grid 18 and the detection panel 35a is misaligned, the variation range of the output value of each individual detection pixel group 66 is smaller than that of the one detection pixel 65. Thus, the output value becomes stable, and the stable AEC can be carried out without being affected by the misalignment in the geometrical disposition. Also, the detection pixel group 66 has an increased signal amount of the dose detection signal because the detection pixel group 66 is larger than the detection pixel 65 in size, and an S/N ratio is improved. Note that, in a case where the one detection pixel group 66 composes the one dose detection sensor, the detection pixel group 66 preferably includes a number of pixels that are at an invisible level after the defect correction, and more preferably on the order of ten pixels. This size of dose detection sensor is much smaller than the striped dose detection sensors of five hundred pixels according to the U.S. Pat. No. 6,952,465, and does not cause degradation in the image quality of the X-ray image. Note that, the detection pixel group 66 is composed of the plurality of detection pixels 65 adjoining in the X2 direction in an example of FIG. 23, but in a like manner, the detection pixel group 66 may be composed of a plurality of detection pixels 65 adjoining in the Y2 direction or both the X2 and Y2 directions. In a case where the one detection pixel group 66 composes the one dose detection sensor, as shown in FIG. 23, the arrangement period S1, S2 corresponds to a distance between the two detection pixel groups 66. In the case of FIG. 23, the arrangement period S1 is 4. In each of the above embodiments, the pixels 45 for image detection and the detection pixels 65 functioning as the dose detection sensors are independent of each other and the detection pixels 65 are read out in a destructive manner, so portions of the detection pixels 65 become so-called point defects. However, since the size of one pixel is small enough, it is known as a result of experiment that performing interpolation processing by which pixel values of a column having the detection pixel 65 are interpolated with pixel values of adjoining columns without having the detection pixel 65 makes the defect hard to see by a human eye, and hence there is no substantial problem. However, it is best to prevent the occurrence of the point defects, and hence adopting a detection panel 100 having structure as shown in FIG. 24 can eliminate the need for performing the interpolation processing as the defect correction. In FIG. 24, the detection panel 100 includes first pixels 101 for specific use in image detection and second pixels 102 for use in both the image detection and the AEC. The first and second pixels 101 and 102 are arranged into a matrix at an appropriate ratio, just as with the pixels 45 and the detection pixels 65 of the above embodiments. An arrangement period of the second pixels 102 is different from the arrangement period of the X-ray absorbing portions 33 of the grid 18. Each of the first and second pixels 101 and 102 has two photodiodes 103 and 104. The photodiodes 103 and 104 of the first pixel 101 are connected in parallel, and one end is connected to the signal line 52 through the TFT 47. In the second pixel 102, on the other hand, one end of the photodiode 103 is connected to the signal line 52 through the TFT 47, just as with that of the first pixel 101, but the photodiode 104 is directly connected to the signal line 52 without passing through the TFT 47. In other words, the photodiode 104 of the second pixel 102 has the same structure as the detection pixel 65 of the above embodiments. From the first pixel 101, electric charge accumulated in the two photodiodes 103 and 104 is read out. From the second pixel 102, on the other hand, electric charge accumulated only in the photodiode 103 is read out. In the second pixel 102, electric charge produced in the photodiode 104 is used for the AEC and does not contribute production of the X-ray image. Thereby, provided that the photodiodes 103 and 104 have the same opening size, the amount of accumulated electric charge of the second pixel 102 is approximately half of that of the first pixel 101 under the same incident dose. However, it is possible to prevent degradation in the image quality of the X-ray image, as compared with the above embodiments in which no pixel value is obtained from the positions of the detection pixels 65 and the interpolation processing is absolutely necessary. A coefficient or the like that converts a pixel value of the second pixel 102 into a value corresponding to a pixel value of the first pixel 101 by multiplication is calculated in advance based on the opening size and the like of the photodiodes 103 and 104. Multiplying an output of the second pixel 102 by the coefficient can produce the X-ray image without performing the interpolation processing, and almost completely eliminate an adverse effect on the image equality of the X-ray image that is caused by using a part of the pixel for the AEC. In the above first embodiment, the detection pixel 65 that is directly connected to the signal line 52 without passing through the TFT 47 is used as the dose detection sensor. However, as shown in FIG. 25, for example, a detection pixel 110 may be connected to a TFT 113 driven by a gate driver 111 and a scan line 112 that are different from the gate driver 53 and the scan lines 51 of the normal pixels 45. Electric charge accumulated in the detection pixel 110 can be read out independently of the normal pixels 45. Alternatively, with taking advantage of the fact that an electric current that is based on electric charge produced in the pixel 45 flows through the bias line 48 for supplying the bias voltage Vb to each pixel 45, an electric current flowing through the bias line 48 connected to the specific pixel 45 may be monitored to detect the radiation dose. In another case, the radiation dose may be detected based on a leak current leaked from the pixel 45 in a state where all the TFTs 47 are turned off. Furthermore, a dose detection sensor for independent AEC having different structure may be provided in the same plane as the imaging surface 36, separately from the pixels 45. The same plane includes a case where the dose detection sensor is stacked on the TFT active matrix substrate having the TFTs 47 as another layer for the AEC, a case where the dose detection sensor is provided on a side opposite from the TFTs 47 relative to the scintillator, and the like. The dose detection sensor can be provided in any surface, as long as the surface is orthogonal to an X-ray incident direction and in parallel with the TFT active matrix substrate. However, in the case of the dose detection sensor for the independent AEC, the dose detection sensor is preferably in size of an invisible level. More specifically, the dose detection sensor is on the order of ten pixels in size. In each of the above embodiments, in the AEC, the average of the output values of the plurality of dose detection sensors is calculated, and the integral value of the average is compared with the emission stop threshold value. However, a median or a sum is calculated instead of the average of the output values of the plurality of dose detection sensors, and an integral value of the median or the sum may be compared with the emission stop threshold value. According to the present invention, since the output values of the plurality of dose detection sensors are distributed, the use of the median or the sum can obtain the same effect as in the case of the average. Each of the above embodiments is described with taking the electronic cassette, being the portable type X-ray image detecting device, as an example, but the present invention may be applied to a stationary type X-ray image detection device contained in the imaging stand. The console 17 and the electronic cassette 16 are separate from each other, but the console 17 is not necessarily an independent device, and the electronic cassette 16 may has the function of the console 17. In a like manner, the source control device 14 and the console 17 may be integrated in one unit. In each of the above embodiments, the positions of the detection pixels 65 are already known in manufacturing the image detector 35, and the image detector 35 stores the position (coordinates) of every detection pixel 65 in a non-volatile memory (not shown) in advance, but this is not essential. To be more specific, every pixel 45 may be read out in a non-destructive manner, and pixels to be used as the detection pixels may be chosen from all the pixels 45 at any time to read out output values therefrom. For example, in response to choice of the body part to be imaged in an imaging menu, the pixels 45 in needed position are appropriately chosen as the detection pixels. At this time, the detection pixels may be chosen from the pixels 45 such that the arrangement period S1, S2 of the detection pixels or the detection pixel groups does not coincide with the arrangement period F. The present invention is not limited to each of the above-described embodiments, and, as a matter of course, is modified into various configurations within the scope of the present invention. The present invention is applicable to an imaging system using another type of radiation such as γ-rays, instead of the X-rays. Although the present invention has been fully described by the way of the preferred embodiment thereof with reference to the accompanying drawings, various changes and modifications will be apparent to those having skill in this field. Therefore, unless otherwise these changes and modifications depart from the scope of the present invention, they should be construed as included therein. |
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054901863 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a shipping container for a nuclear fuel assembly, and in particular, to such a container for nuclear fuel assemblies which have a plurality of fuel rods held in a hexagonal array by a plurality of grids spaced longitudinally along the fuel rods. The invention also relates to a hold-down device for securing the bottom nozzle of the nuclear fuel assembly. 2. Background of Information In the shipping and storage of nuclear reactor fuel elements and assemblies, which contain large quantities and/or enrichments of the fissile material, U.sup.235, it is necessary to assure that criticality is avoided during normal use, as well as under potential accident conditions. For example, fuel shipping containers are licensed by the Nuclear Regulatory Commission (NRC) to ship specific maximum fuel enrichments (i.e., weights and weight percent U.sup.235) for each fuel assembly design. In order for a new shipping container design to receive licensing, it must be demonstrated to the satisfaction of the NRC that the new container design will meet the requirements of the NRC Rules and Regulations, including those defined in 10 CFR 71 which is incorporated herein by reference. These requirements define the maximum credible accident (MCA) that the shipping container and its internal support structures must endure in order to maintain the subcriticality of the fuel assemblies therein. U.S. Pat. No. 4,780,268, which is assigned to the assignee of the present invention and which is incorporated herein by reference, discloses a shipping container for transporting two conventional nuclear fuel assemblies having a square top nozzle, a square array of fuel rods and a square bottom nozzle. The container includes a support frame having a vertically extending section between the two fuel assemblies which sit side by side. Each fuel assembly is clamped to the support frame by clamping frames which each have two pressure pads. This entire assembly is connected to the container by a shock mounting frame and plural shock mountings. Sealed within the vertical section are at least two neutron absorber elements. A layer of rubber-cork cushioning material separates the support frame and the vertical section from the fuel assemblies. The top nozzle of each of the conventional fuel assemblies is held, along the longitudinal axis thereof, by four longitudinally attached bolts at the four corners of the square top nozzle. The bottom nozzle of some of these conventional fuel assemblies has a chamfered end. These fuel assemblies are held, along the longitudinal axis thereof, by a bottom nozzle spacer which holds the chamfered end of the bottom nozzle. This and other shipping containers (e.g., RCC-4 for generally square cross-sectional geometry pressurized water reactor (PWR) fuel assemblies) used by the assignee of the present invention are described in certificate of compliance No. 5450, Docket 71-5450, U.S. Nuclear Regulatory Commission, Division of Fuel Cycle and Material Safety, Office of Nuclear Material Safety and Safeguards, Washington, DC 20555, which is incorporated herein by reference. In nuclear reactors of the type originally designed in the former Soviet Union, the reactor core is comprised of a large number of elongated fuel assemblies. Each of these fuel assemblies includes a plurality of fuel rods held in an organized hexagonal array by a plurality of hexagonal grids spaced longitudinally along the fuel rods and secured to stainless steel control rod guide thimble tubes. Subsequently, the Soviet-style fuel assemblies were redesigned by the assignee of the present invention in order to provide, for example, more reliable operation. The guide thimble tubes of the redesigned fuel assemblies extend above and below the ends of the fuel rods and are attached to top and bottom nozzles, respectively. Such fuel assemblies are arranged in the reactor vessel with the bottom nozzles resting on a lower core plate. An upper core plate rests on the top nozzles. These fuel assemblies may contain U.sup.235 concentrations of up to about 4.80 to 5.00 weight percent U.sup.235. Under normal manufacturing conditions, the dimensions of the fuel assemblies may vary. For example, the dimensions of the six sides of the hexagonal array may differ by about .+-.2.0 mm between individual fuel assemblies. The top nozzle of the fuel assembly includes a cylindrical outer barrel, a cylindrical inner barrel and a hub. The outer barrel forms a first end of the top nozzle at the top of the fuel assembly. The inner barrel, which has a diameter smaller than the outer barrel, is attached to the hub, which forms a second end of the top nozzle opposite from the first end. The outer barrel has a shoulder facing the second end. The inner barrel telescopes into the outer barrel. The hub interfaces the plurality of fuel rods at the second end. The relatively heavy (e.g., 70 pounds) top nozzle is susceptible to transportation induced damage to the guide thimble tubes. For example, during normal transportation, vibration in the top nozzle inner barrel may be detrimental to the guide thimble tubes. Because of the unique design of the fuel assembly, which allows movement of the outer barrel along the longitudinal axis of the fuel assembly with respect to the relatively smaller inner barrel, it is not feasible to position adjustable hardware for securing the top nozzle in order to provide the necessary supporting restraint of the fuel assembly during shipment thereof. The bottom nozzle includes a longitudinally extending recess formed by a hexagonal barrel, a spherical taper, and a cylindrical barrel which has a diameter smaller than the hexagonal barrel. The spherical taper forms a tapered bore within the longitudinally extending recess tapering toward the bottom end. The spherical taper, also, forms an internal shoulder between the hexagonal barrel and the bottom end. There is a need, therefore, for an improved shipping container for a nuclear fuel assembly having a double-barrelled top nozzle. There is also a need for an improved shipping container for a nuclear fuel assembly having a double-barrelled bottom nozzle. More particularly, there is a need for such a container for a nuclear fuel assembly having a hexagonal geometry. There is an even more particular need for such a container which accommodates for manufacturing tolerances in the hexagonal geometry. There is another more particular need for such a container for a nuclear fuel assembly including a top nozzle having an outer barrel and an inner barrel of smaller diameter which telescopes into the outer barrel. There is yet another more particular need for such a container for a nuclear fuel assembly including a bottom nozzle having a longitudinally extending recess formed by a hexagonal barrel, a spherical taper, and a cylindrical barrel having a diameter smaller than the hexagonal barrel. There is still another more particular need for such a shipping container for transporting high enrichment fuel assemblies. SUMMARY OF THE INVENTION These and other needs are satisfied by the invention which is directed to a shipping container for a nuclear fuel assembly. The fuel assembly includes an array of a plurality of fuel rods; and a top nozzle having a top end, an outer barrel, an inner barrel, and a shoulder between the barrels. The shipping container may include a support mechanism for supporting the top nozzle and the plurality of fuel rods, a housing for the support mechanism and the fuel assembly, and a top nozzle holder secured to the support mechanism for holding the top nozzle. The top nozzle holder may include a shoulder holder for holding the shoulder. The top nozzle holder may also include an end holder for enclosing and holding the top end. The end holder may further include a spacer member, a resilient spacer and a support member. The spacer member may be secured to the support mechanism. The resilient spacer may be attached to the support member which forms a surface supported by the spacer member for holding the top end of the top nozzle therein. The resilient spacer may separate the support member from the top end of the top nozzle. The top nozzle holder may further include a shoulder clamp for clamping the shoulder holder to the support mechanism. The shoulder holder may include a resilient split ring having a first gap for positioning around the inner barrel, and a resilient split support for encasing the resilient split ring. The resilient split support may have a bore running therethrough, a second gap, and a counter-bore which encases the resilient split ring therein adjacent the shoulder. The shoulder clamp may clamp the resilient split support thereby closing the first gap of the resilient split ring, closing the second gap of the resilient split support, and securing the inner barrel to the support mechanism. The nuclear fuel assembly may also include a bottom nozzle and a plurality of grids supporting the array. The shipping container may further include a support mechanism for supporting the top nozzle, the plurality of grids, and the bottom nozzle; a housing for housing the support mechanism and the nuclear fuel assembly; a top nozzle holder secured to the support mechanism for holding the top nozzle; a plurality of grid supports for supporting the array; a plurality of clamping mechanisms for clamping the array; a plurality of guide plates for guiding the nuclear fuel assembly between adjacent ones of the plurality of grid supports; and a bottom nozzle holder secured to the support mechanism for holding the bottom nozzle. The support mechanism may have a first surface for abutting the array and a second surface which is perpendicular to the first surface. Each of the plurality of clamping mechanisms may clamp a corresponding one of the plurality of grids to a corresponding one of the plurality of grid supports. Each of the plurality of grid supports may support a corresponding one of the plurality of grids on the second surface. The nuclear fuel assembly array may be a hexagonal array having six sides. The first surface of the support mechanism may abut a first side of the array. Each of the guide plates may have two surfaces for guiding a second side and a third side of the hexagonal array. Each of the grid supports may include a first support for supporting the second side of the array, a second support for supporting the third side of the array, a base plate for fixedly supporting the first and second supports thereto, a bearing pad for slidably supporting the base plate, and a limiter for limiting a sliding motion of the base plate on the bearing pad which is fixedly mounted to the second surface of the support mechanism. Alternatively, each of the guide plates may have a guide side for guiding the nuclear fuel assembly, and an absorbing side having a coating of gadolinium oxide. The bottom nozzle of the nuclear fuel assembly may include a longitudinally extending recess. The bottom nozzle holder may be secured to the support mechanism for holding the bottom nozzle and may include a recess holder for holding the bottom nozzle within the longitudinally extending recess. The recess holder may include a wedge mechanism for wedging against the bottom nozzle within the longitudinally extending recess and a moving mechanism for moving the wedge mechanism within the longitudinally extending recess. The bottom nozzle may further include a bottom end and a tapered bore or shoulder within the longitudinally extending recess tapering toward the bottom end. The recess holder may include a gripper mechanism for gripping the tapered bore or shoulder within the bottom nozzle and a moving or engaging mechanism for moving the gripper mechanism against the tapered bore or shoulder. The gripper mechanism may include a plurality of grippers for gripping the shoulder within the bottom nozzle. Each of the grippers may have a gripping end and a pivot end. The engaging mechanism may include a base for pivotally mounting the pivot end of each of the grippers and a moving mechanism for moving the gripping end of each of the grippers. The moving mechanism may include an operating mechanism for moving the moving mechanism which engages each of the gripping ends in order to move the gripping ends toward the shoulder within the bottom nozzle. The operating mechanism may also disengage the moving mechanism in order to move the gripping ends away from the shoulder within the bottom nozzle. The base may be inserted adjacent the support mechanism and within the bottom end of the bottom nozzle. The bottom nozzle may include a hexagonal barrel, a spherical taper, and a cylindrical barrel having a diameter smaller than the hexagonal barrel. The spherical taper may interconnect the hexagonal barrel and the cylindrical barrel which forms the bottom end of the nuclear fuel assembly. The bottom nozzle holder may further include a spacer having a hole for inserting the cylindrical barrel therein and a tapered surface for abutting the spherical taper in order to space the bottom end of the nuclear fuel assembly from the support mechanism. The moving mechanism may include a cam mechanism having a plurality of cam surfaces for camming a corresponding one of the gripping ends of the plurality of grippers. Adjacent ones of the plurality of grippers may include a spring mechanism for forcing each of the adjacent grippers against a corresponding one of the plurality of cam surfaces. The nuclear fuel assembly may have a central longitudinal axis. Each of the support mechanism, the base and the moving mechanism may have a hole which is positioned on the central longitudinal axis. The support mechanism may have a surface and the hole of the moving mechanism may be threaded. The operating mechanism may include a screw mechanism for rotating the moving mechanism, a collar, and a spring biased between the moving mechanism and the collar in order to provide a pre-load force for the screw. The screw may have a head and a shaft. The head may abut the surface of the support mechanism. The shaft may have a non-threaded portion and a threaded portion. The non-threaded portion may be adjacent the head and may pass through the holes of the support mechanism and the base. The threaded portion may be adjacent the non-threaded portion and may be threaded through the threads of the hole of the moving mechanism. The collar may be fixedly attached to the threaded portion and separated from the moving mechanism. The moving mechanism may further include a first blocking mechanism for blocking rotation of the moving mechanism. The first blocking mechanism may include a plurality of blocking surfaces which are between adjacent ones of the plurality of cam surfaces. Each of the blocking surfaces may abut the corresponding one of the gripping ends of the grippers whenever the moving mechanism is fully disengaged. The moving mechanism may further include a second blocking mechanism for blocking rotation of the moving mechanism. The second blocking mechanism may include a plurality of blocking tabs. Each of the blocking tabs may be attached to a corresponding one of the cam surfaces in order that each one of the blocking tabs abuts the corresponding one of the gripping ends of the grippers whenever the moving mechanism is fully engaged. Alternatively, a bottom nozzle holder may be provided for use with a shipping container for a nuclear fuel assembly. The nuclear fuel assembly may include a plurality of fuel rods; and a bottom nozzle having a longitudinally extending recess, a bottom end, and a shoulder within the longitudinally extending recess. The bottom nozzle holder may include a gripper mechanism for gripping the shoulder within the bottom nozzle, and an engaging mechanism for engaging the gripper mechanism against the shoulder. |
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description | The present invention relates to handling of fuel assemblies for nuclear reactors. More specifically, the present invention relates to a method and device for installing a top nozzle repair sleeve in a nuclear fuel assembly to assist in structural transfer of forces when the fuel assembly is lifted. Present-day nuclear power reactors use fuel in the form of fuel assemblies. The fuel assemblies usually are comprised of a uranium dioxide matrix, generally in the shape of pellets, stacked end-to-end. The stacked pellets are clad on their respective exterior by a zirconium metal alloy which prevents direct contact between the uranium dioxide pellets and reactor coolant. The fuel rods formed by the pellets cladded by the zirconium metal alloy are placed in a structural skeleton in a side by side arrangement to allow a dense packing of nuclear material. At the top and bottom of the fuel assembly, respective top and bottom nozzles are positioned to allow coolant to be channeled through the fuel assembly to remove heat generated by the nuclear fuel. Industry experience has discovered that fuel assemblies can degrade over time near the top nozzle section of the fuel assembly. The top nozzle is connected to an interior shaft, (i.e. a guide thimble), to establish a structural connection between the top nozzle and a remainder of the fuel assembly. Industry experience has identified that a connection established between the top nozzle and the guide thimble, often in the form of a guide thimble sleeve, can be prone to stress corrosion cracking through continued fuel assembly use. When the nuclear reactor is refueled and depleted fuel assemblies removed from or repositioned in the reactor core, each fuel assembly is lifted and/or removed from the reactor using a fuel handling or polar crane. Degradation of the structural support established between the top nozzle and the guide thimbles prevents movement of the degraded fuel assembly as the fuel assembly may not be structurally adequate to lift its own weight. During inspections of the fuel assembly prior to lifting, if it is determined that the fuel assembly exhibits stress corrosion cracking or some other defect in the guide thimble structural load path, either the fuel assembly is moved in a piece-wise fashion, a special lifting device is fabricated to lift the assembly, or an extensive and lengthy structural repair is performed to allow the fuel assembly to be lifted. Current repair alternatives to correct stress corrosion cracking often delay nuclear reactor restart, thereby increasing the economic costs for the reactor owner. There is a need to provide a method and device to repair damaged fuel assemblies such that a damaged fuel assembly may be lifted in a safe condition. There is a further need to provide a method and device to provide a permanent repair for fuel assemblies which have stress corrosion cracking in areas such as the top nozzle, the guide thimble sleeves and the guide thimble. There is a further need to provide an economically efficient method and device for repair of degraded fuel assemblies such that the repair may be accomplished quickly and efficiently. There is a still further need to provide a device and method to repair damaged fuel assemblies that are radioactive. There is a still further need to provide a device and method to repair damaged fuel assemblies that are radioactive. It is an objective of the present invention to provide a method and device to repair damaged nuclear fuel assemblies such that a damaged fuel assembly may be lifted safely. It is also an objective of the present invention to provide a method and device to provide a permanent repair for fuel assemblies which exhibit structural defects, such as stress corrosion cracking, between the top nozzle, the guide thimble sleeves and the guide thimble. It is furthermore an objective of the present invention to provide an economically efficient method and device for repair of degraded fuel assemblies such that the repair may be accomplished quickly and efficiently. It is also an objective of the present invention to provide a device and method to repair damaged fuel assemblies that are radioactive. These and other objectives of the present invention will be achieved as illustrated and described. The present invention provides a repair sleeve for a nuclear fuel assembly. The present invention provides a shaft with a first end, a second end and a diameter, the diameter configured to fit into a guide thimble opening of a top nozzle of the nuclear fuel assembly, wherein the diameter of the shaft is dimensioned such that an exterior of the shaft snugly fits into the guide thimble opening, wherein the shaft has at least two openings. The present invention also provides for at least two tendons extending through the openings, wherein the tendons are configured to deflect in an instance of a horizontal load on the tendon, the tendons having a dimple configured to be inserted into a dimple of a guide thimble sleeve, wherein the shaft is configured to internally accept a control component of the fuel assembly and wherein at least two tendons do not deflect under a load when a thimble insert assembly is installed. The invention furthermore provides a method to repair a fuel assembly. The method comprises providing a repair sleeve, the repair sleeve having a shaft with a first end, a second end and a diameter, the diameter configured to fit into a guide thimble opening of a top nozzle of the fuel assembly. The method also provides that the diameter of the shaft is dimensioned such that an exterior of the shaft snugly fits into the guide thimble opening, wherein the shaft has at least two openings. The method also provides that at least two tendons extend through the openings, the tendons configured to deflect in an instance of a load on the tendon, the tendons having a dimple configured to be inserted into a dimple of a guide thimble sleeve. The method furthermore provides for inserting the repair sleeve in the guide thimble opening in the top nozzle of the nuclear fuel assembly such that the dimples of the tendons project into the dimples of the guide thimble sleeve, and inserting a control component into an interior of the repair sleeve. Referring to FIG. 1, a repair sleeve 10 is illustrated which is used to aid in structural load transfer in a fuel assembly. Structural defects arising from stress corrosion cracking, for example, limit the capability of operators to move fuel assemblies as these structural defects may negatively impact safe lifting of the fuel assembly. Structural defects are often found in a swaged area 18 which connects the top nozzle 12 to the guide thimble 22. The repair sleeve 10 has a shaft 14 which is configured to allow a structural load transfer of force from the body of the fuel assembly through the top nozzle 12 of the fuel assembly. The repair sleeve 10 is configured with a first end 24 and a second end 26. The second end 26 is configured to be inserted into a nozzle opening 30 in the top nozzle 12 of the fuel assembly. The first end 24 may protrude from a top surface 38 of the top nozzle to allow for connection of lifting components as required. The first end 24 may be configured such that an installer may easily differentiate the first end 24 from the second end 26. Differentiation may be through geometric variation, such as an end which flairs outwardly, through the incorporation of a hole, or marking the exterior of the first end 24. Other configurations are possible and as such, the configuration presented in FIG. 1 is but one exemplary embodiment. The shaft 14 of the repair sleeve 10 may be made of stainless steel, for example, to allow for rigidity during a fuel assembly lift. Other materials, such as Inconel, stainless steel, and zirconium alloys, may also be used as well as materials which will eliminate or lessen potential galvanic reaction between fuel assembly structural materials and the repair sleeve 10. The material chosen for the repair sleeve 10 shall not exhibit permanent plastic deformation of the material upon insertion of the repair sleeve 10 into the top nozzle 12, associated guide thimble sleeve 32 and guide thimble 22. The shaft 14 may be configured with a plurality of sleeve openings 28. The number of sleeve openings 28 may be varied for the shaft 14. Tendons 20 may be positioned through the sleeve openings 28. The tendon width 40 may be varied such that the overall physical dimensions of the tendons 20 may be altered. The tendons 20 may be configured with a dimple 16 which corresponds to openings in a dimple area 34 in the guide thimble sleeve 32. The embodiment provided in FIG. 1 illustrates a connection between a first dimple area 34 and the projection 16 of the tendon 20. The repair sleeve 10, however, may have an overall length chosen such that the projection 16 of the tendons 20 extend to a second or third dimple area further inside the fuel assembly guide thimble. The diameter of the sleeve 10, except for the projection 16, may be a constant value. The length of the projection 16 which projects into the dimple area 34 may be configured to closely fit into the overall shape defined by the dimple area 34. The projections 16 may be configured in a trapezoidal shape, a hemispherical shape or other appropriate geometry. The number of dimples 16 in contact with dimple areas 34 may also be varied such that more or less structural support is established. The number of projections 16 per tendon 20 may also be varied. Tendon length may be chosen such that the tendon 20 may extend to and between multiple dimple areas, providing additional structural support connection. Referring to FIG. 2, a second embodiment of a repair sleeve 70 is illustrated. The repair sleeve 70 has a shaft 72. The repair sleeve 70 has a first end 56 and a second end 58. The second end 58 is configured to be inserted into a guide thimble opening of the top nozzle 52. The first end 56 of the repair sleeve 70 may be configured with a lapped edge 54 which extends beyond an external top surface 74 of the top nozzle 52. The lapped edge 54 may have an external diameter which is greater in circumference than the external diameter of the opening of the top nozzle 52 for the guide thimble 50. Although illustrated as a circular lapped edge 54, other configurations are possible, such as square, hexagonal, or octagonal for example. The lapped edge 54 may be finally configured while the repair sleeve 70 is installed in the top nozzle 52. Alternatively, the lapped edge 54 may be preformed prior to installation of the repair sleeve 70. A projection 60 may be formed on a tendon 66 of the repair sleeve 70. The length of the tendon 66 may be chosen such that the projection 60 is placed in a dimple area 76 formed from the swaged area 78 of the guide thimble 50 and the guide thimble sleeve 68. The contact established between the projection 60 and the dimple area 76 may be configured to allow a transfer of a specified amount of force. Similar to the first embodiment, the repair sleeve 70 may be modified such that the overall length of the sleeve 70 may reach multiple dimple areas in the guide thimble 50. The projection 60 may be formed in any geometric configuration such as a hemispherical, trapezoidal or other arrangement. Referring to FIG. 3, a graph of the structural capacity of the repair sleeve 10 is illustrated. The vertical axis of the graph represents load carrying capacity of the repair sleeve 10. The horizontal axis of the graph represents overall position of the repair sleeve. As illustrated, the repair sleeve provides for an increase in load with a corresponding increase in displacement. Load carrying capacity then decreases after a maximum load carrying capacity is reached. Load carrying capacity is related to the amount of penetration of the projections into the dimple area. Greater penetration of the projections into the dimple area allows greater load carrying capacity. The present invention provides a structural support that engages existing features of fuel assemblies to secure the upper or top nozzle of the fuel assembly to internal guide thimbles. Multiple repair sleeves may be used to transfer only a partial load of the fuel assembly if required. This connection is used during lifting of the fuel assembly to allow the individual fuel rods to be lifted in unison with a desired factor of safety. The shaft 14 is split into separate sections (the tendons 20) by the sleeve openings 28 in the shaft material in the location of the dimple area 34 to deflect into the dimple area (34). Operationally, a repair sleeve 10 is provided. The repair sleeve 10 is inserted into a guide thimble opening in the top nozzle 12 of the nuclear fuel assembly such that the projections 16 of the tendons 20 project into the dimple area 34 of the guide thimble sleeve. The insertion may be performed through a robotic device, or remote delivery tooling to install the sleeve in an irradiated environment or through use of a crane. The tooling can be configured to deliver singular or multiple sleeves at a time and install the sleeves to the engaged position. The insertion of the sleeve 10 in the top nozzle 12 causes the tendons 20 of the sleeve 10 to flex inward toward a centerline of the sleeve 10. The sleeve insertion is then continued until the dimples 16 of the sleeve 10 intersect a dimple area 34 of the swaged area 18. The tendons 20 of the sleeve 10 then extend away from a centerline of the sleeve 10 to allow the projections 16 to project into the dimple area 34. A thimble insert assembly 1 (control component, BPRA, WABA, plug) of the fuel assembly is then inserted into an interior of the repair sleeve 10. The installation of the thimble insert assembly into the repair sleeve 10 prevents the dimples 16 from exiting the dimple area 34 through plastic deformation of the sleeve 10. A structural connection is thereby established between the projections 16 and the dimple area 34. The first end 24 may be configured to extend from the top nozzle 12 or may be manipulated such that a desired geometric configuration is established. The repair sleeve 10 may be removed from the guide thimble 22 by removing the control component from the interior of the repair sleeve 10. The first end 14 of the repair sleeve 10 may then be pulled such that the tendons 20 of the sleeve 10 deflect allowing the dimple 16 to be removed from the dimple area 34. The present invention provides several advantages to other methods and devices for lifting fuel assemblies. The installation of the repair sleeve can be accomplished in a quick and efficient manner limiting economic expense. Moreover, the repair sleeve can be installed from a remote location, thereby limiting radiation exposure for workers installing the sleeves. The installation of the repair sleeve is performed with tools that are not complex, thereby allowing ordinary skilled craftspeople to install the sleeves with minimal special training. The present invention has a minimum of moving parts, thereby limiting potential failure over the life of the repair sleeve. The present invention also uses materials which are known to be successful in harsh environments, thereby limiting further degradation of the fuel assembly over time or affecting nuclear components in the facility. The present invention furthermore can be adapted to achieve different configurations to allow for differing attachment configurations and structural capacities. The present invention furthermore provides a configuration that will allow inspection of the sleeve through visual examination in an installed condition. The repair sleeve will not negatively affect overall fuel assembly pressure drop due to its relatively small size. The repair sleeve small size additionally limits the ultimate amount of disposed radiation waste for the fuel assembly. Additional sleeves 10 may be installed in a fuel assembly to provide further support. In the foregoing specification, the invention has been described with reference to specific exemplary embodiments, thereof. It will, however, be evident that various modifications and changes may be made thereunto without departing from the broader spirit and scope of the invention as set forth in the claims. The specification and drawings are accordingly to be regarded in an illustrative rather than a restrictive sense. |
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048851248 | abstract | In the operation of a power reactor having a pressure vessel (2) containing a core and a plurality of thimbles extending into the core to provide passages for a flux detector, a movable support member (12) located outside of the pressure vessel (2) and carrying a plurality of tubular members (10) each detachably connected to a respective thimble, and a transfer unit (20) normally coupled to the tubular members (10) for introducing a flux detector into a selected thimble via a respective tubular member (10), obstructions are removed from a selected thimble while the reactor remains in operation, by the steps of:. disconnecting at least the respective tubular member from the transfer unit (20); PA0 introducing an obstruction removal device (40,42) into the respective tubular member(10); and PA0 advancing the obstruction removal device (40,42) through the respective tubular member (10) and into the selected thimble to the location of the obstruction. |
abstract | In an X-ray diffraction apparatus, a high brightness source, such as a rotating anode generator, is combined with demagnification X-ray optics to produce a beam with small image size and high-intensity. In one embodiment, an elliptical X-ray optic is positioned relative to the source and image focal points so that the magnification of the optic is less than one. The combination can produce high-intensity beams with beam images at the sample of less than 0.1 mm. |
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061987921 | claims | 1. An apparatus for forming an optical image of a mask pattern on a wafer that comprises: a first chamber housing a wafer to be exposed to extreme ultraviolet (EUV) radiation to form a pattern image on the wafer; a second chamber that is separated from the first chamber by a partition which defines an aperture that is permeable to inert gas and that houses an image system, which is disposed between the mask having a pattern for circuit manufacturing and the wafer, for receiving EUV radiation reflected from the mask and directing it to the wafer through the aperture; and means for maintaining a flow of inert gas over the wafer surface to remove contaminants that develop from the wafer upon exposure to the EUV radiation. (a) providing a photolithographic system that comprises: (b) exposing the mask to EUV radiation which is at least partially reflected onto a surface of the wafer, to form a pattern image on the wafer surface; and (c) maintaining a flow of inert gas over the wafer surface to remove contaminants that develop from the wafer upon exposure to the EUV radiation. 2. The apparatus of claim 1 wherein the means for maintaining the flow of inert gas establishes a gaseous flow rate to achieve a mass transfer Peclet number between 20 and 35. 3. The apparatus of claim 1 wherein the means for maintaining a flow of inert gas comprises a source of inert gas that delivers a stream of inert gas to the surface of the wafer below the aperture. 4. The apparatus of claim 3 wherein the aperture defines an opening on the lower surface the partition that has an area of 1 cm.sup.2 to 5 cm.sup.2. 5. The apparatus of claim 4 wherein the opening defines a slit that has a width of 0.4 cm to 1.5 cm and a length of 2.6 cm to 3.7 cm. 6. The apparatus of claim 1 further comprising vacuum means for supplying a vacuum to the second chamber. 7. The apparatus of claim 1 wherein the first chamber includes a source of inert gas having an inlet at a first side of the first chamber and an outlet at a second side of the first chamber that is substantially opposite that of the inlet. 8. The apparatus of claim 1 wherein the inert gas comprises helium, argon, hydrogen, oxygen and mixtures thereof. 9. The apparatus of claim 1 wherein the contaminants comprise hydrocarbons. 10. The apparatus of claim 1 wherein the second chamber comprises a second outlet through which part of the inert gas and contaminants exit. 11. A method for forming an optical image of a mask pattern on a wafer that comprises the steps of: 12. The method of claim 11 wherein the flow of inert gas to achieve a mass transfer Peclet number between 20 and 35. 13. The method of claim 11 wherein the means for maintaining a flow of inert gas comprises a source of inert gas that delivers a stream of inert gas to the surface of the wafer below the aperture. 14. The method of claim 13 wherein the means for maintaining the flow of inert gas establishes a gaseous flow rate to achieve a mass transfer Peclet number between 20 and 35. 15. The method of claim 14 wherein the aperture defines an opening on the lower surface the partition that has an area of 1 cm.sup.2 to 5 cm.sup.2. 16. The method of claim 11 further comprising the step of supplying a vacuum to the second chamber. 17. The method of claim 11 wherein the first chamber includes a source of inert gas having an inlet at a first side of the first chamber and an outlet at a second side of the first chamber that is substantially opposite that of the inlet. 18. The method of claim 11 wherein the inert gas comprises helium, argon, hydrogen, oxygen and mixtures thereof. 19. The method of claim 11 wherein the contaminants comprise hydrocarbons. 20. The method of claim 11 wherein the second chamber includes a second outlet through which part of the inert gas and contaminants exit. 21. A process for fabrication of a device comprising at least one element having a dimension .ltoreq.0.25 .mu.m, such process comprising construction of a plurality of successive levels, construction of each level comprising lithographic delineation, in accordance with which a subject mask pattern is illuminated to produce a corresponding pattern image on the device being fabricated, ultimately to result in removal of or addition of material in the pattern image regions, in which illumination used in fabrication of at least one level is extreme ultra-violet radiation, characterized in that the process employs a chamber that houses a wafer that is exposed to extreme ultraviolet radiation to form a pattern image on the wafer wherein the chamber has a partition that has an aperture through which the radiation enters the chamber, wherein the partition and surface of the wafer define a path, and wherein a flow of inert gas is maintained along the path and over the wafer surface to remove contaminants that develop from the wafer upon exposure to the radiation. 22. The process of claim 21 in which lithographic delineation is by projection. 23. The process of claim 22 in which projection comprises ringfield scanning comprising illumination of a straight or arcuate region of a projection mask. 24. The process of claim 22 in which projection comprises reduction ringfield scanning in accordance with which an imaged arcuate region on the image plane is of reduced size relative to that of the subject arcuate region so that the imaged pattern is reduced in size relative to the mask region. 25. The process of claim 21 wherein the flow of inert gas has a gaseous flow rate to achieve a mass transfer Peclet number between 20 and 35. |
051981840 | abstract | A reactor pressure vessel is disposed in a reactor containment vessel and is supported by a pedestal having a cylindrical structure. The inside of the reactor containment vessel is divided into upper and lower drywells by means of a diaphragm floor. A line, a cable and a duct are disposed in and between the upper and lower drywells in the reactor containment vessel. The pedestal comprising a plurality of concrete wall sections and a plurality of connecting vent sections which are arranged alternately along a circumferential direction of the cylindrical pedestal, wherein the line, the cable and the duct are arranged in each of the connecting vent sections and a vent pipe is arranged in each of the concrete sections. A vacuum breaker is further disposed in the reactor containment vessel at a portion above the open end of the vent pipe, the vacuum breaker is connected to a fixing pipe for mounting a vacuum breaker to the pedestal and the fixing pipe has one end opened to the drywell. The vent pipe may have an end portion extended to the diaphragm floor and opened to the drywell and a vacuum breaker is mounted to the extended portion of the vent pipe and connected to a fixing pipe for mounting a vacuum breaker to the pedestal, the fixing pipe having one end opened to an inside of the vent pipe. |
041486870 | abstract | Nuclear fuels in the form of rod bundle with one or several beryllium rods inserted therein are proposed for use in power reactors positioned to obtain reactivity increase in order to save the D.sub.2 O inventory in a heavy water reactor or to relax the requirement of uranium enrichment if used in light water reactors. |
description | This is a 371 national phase application of PCT/JP2008/063840 filed 1 Aug. 2008, claiming priority to Japanese Patent Application No. JP 2007-206427 filed 8 Aug. 2007, the contents of which are incorporated herein by reference. The present invention relates to an electrostatic coating apparatus for electrostatically coating an object to be coated and more particularly to an electrostatic coating apparatus provided with a rotary atomizing head that rotates to atomize a coating material. There has heretofore been known an electrostatic coating apparatus including a rotary atomizing head that rotates to atomize a coating material and configured to electrostatically coat an object to be coated such as a vehicle body. Such apparatus is arranged to drivingly rotate the rotary atomizing head applied with electrostatic high voltage, atomizing a fluid coating material supplied to this rotary atomizing head into fine particles by centrifugal force while electrically charging the fine coating particles with the electrostatic high voltage applied to the rotary atomizing head, thus ejecting out the particles. In general, electrostatic coating is performed in such a manner of setting an object to be coated to a positive electrode and an electrostatic coating apparatus to a negative electrode, thereby forming an electrostatic field therebetween, and attracting an atomized coating material negatively charged to the object by electrostatic force. The above electrostatic coating apparatus is disclosed in for example Patent Literature 1. The electrostatic coating apparatus of Patent Literature 1 employs an electric motor as a driving source for driving the rotary atomizing head to rotate. The use of the electric motor can provide improved control response related to rise time and fall time, thus controlling the number of revolutions of the rotary atomizing head to a desired number in a short time (e.g., in about 0.5 seconds). Accordingly, coating can be performed more efficiently than the case using an air motor. The motor can attain a stable number of revolutions, leading to improved coating quality. Citation List Patent Literature Patent Literature 1: JP2007-98382 A Technical Problem The rotary atomizing head is applied with electrostatic high voltage. Thus, when this high voltage is also applied to the electric motor, the high voltage is also applied to a power supply circuit of the electric motor, imposing a burden on the power supply circuit. Therefore, it is preferable to electrically insulate the electric motor from the rotary atomizing head and a high-voltage member having the same potential as the former. However, the voltage applied to the rotary atomizing head and others is an extremely high voltage. To reliably insulate the electric motor from the rotary atomizing head and others, therefore, an insulation distance between the rotary atomizing head and others and the electric motor, in particular, a creepage insulation distance has to be sufficiently long. As a result, the electrostatic coating apparatus is apt to be increased in size just by the long insulation distance. The electrostatic coating apparatus is sometimes mounted for example in a robot for use and thus size reduction and weight reduction are demanded. The present invention has been made in view of the circumstances and has a purpose to provide an electrostatic coating apparatus capable of electrically insulating an electric motor from a member to which an electrostatic high voltage is applied and reducing the size and weight of the electrostatic coating apparatus. Solution to Problem A solution is an electrostatic coating apparatus of a rotary atomizing type for electrostatically coating an object to be coated, comprising: a rotary atomizing head that rotates to atomize a coating material and that is applied electrostatically with high voltage; an electric motor that drives the rotary atomizing head to rotate and that is electrostatically grounded; a spindle made of an electrically insulating material for electrically insulating the electric motor from the rotary atomizing head and a speed increasing device mechanically connected to the rotary atomizing head and having the same potential as the rotary atomizing head, the spindle being inserted through the electric motor and mechanically connected to the speed increasing device, and the spindle including one or more insulation distance enlarging portions configured to increase a creepage insulation distance from the rotary atomizing head or the speed increasing device to the electric motor; and one or more fixed insulating members fixedly placed between the speed increasing device and the electric motor for electrically insulating the electric motor from the rotary atomizing head and the speed increasing device, the fixed insulating members including one or more insulation distance enlarging portions configured to increase a creepage insulation distance from the rotary atomizing head or the speed increasing device to the electric motor, the spindle includes, as the insulation distance enlarging portion, a zigzag portion having a zigzag form to increase the creepage insulation distance, and the fixed insulating member including, as the insulation distance enlarging portion, a zigzag portion having a zigzag form to increase the creepage insulation distance. The electrostatic coating apparatus of the invention includes the spindle and the fixed insulating member for electrically insulating the electric motor from the rotary atomizing head and the speed increasing device. Thus, electrostatic high voltage applied to the rotary atomizing head and the speed increasing device is not applied to a power supply circuit through the electric motor and thus no burden is imposed on the power supply circuit. In addition, each of the spindle and the fixed insulating member has the insulation distance enlarging portion configured to increase the creepage insulation distance. The creepage insulation distance from the rotary atomizing head or the speed increasing device to the electric motor can be made sufficiently long. Accordingly, the rotary atomizing head or the speed increasing device and the electric motor can be placed at a short distance in the electrostatic coating apparatus. Providing the sufficient creepage insulation distance by the insulation distance enlarging portion formed in each of the spindle and the fixed insulating member can also achieve size reduction and weight reduction of the spindle and the insulating member. This makes it possible to reliably electrically insulate the electric motor from the member to which electrostatic high voltage is applied and also to reduce the size and weight of the electrostatic coating apparatus. Each term “spindle” and “fixed insulating member” includes the “insulation distance enlarging portion” configured to enlarge the creepage insulation distance. The “insulation distance enlarging portion” may include for example, as mentioned later, a zigzag portion formed in a zigzag shape to increase the creepage insulation distance, an extended portion formed in an extending shape to increase the creepage insulation distance, or the like. (Deleted) Furthermore, the electrostatic coating apparatus of the invention includes, as the insulation distance enlarging portion of the spindle, the zigzag portion having a zigzag form to increase the creepage insulation distance from the rotary atomizing head or the speed increasing device to the electric motor. In addition, the apparatus includes, as the insulation distance enlarging portion of the fixed insulating member, the zigzag portion having a zigzag form to increase the creepage insulation distance from the rotary atomizing head or the speed increasing device to the electric motor. The presence of such zigzag portion can easily provide the long creepage insulation distance. Accordingly, the electric motor can be reliably insulated from the rotary atomizing head or the speed increasing device. Furthermore, in the above electrostatic coating apparatus, preferably, the spindle includes, as the insulation distance enlarging portion, an extended portion to increase the creepage insulation distance, and the fixed insulating member includes the insulation distance enlarging portion, an extended portion to increase the creepage insulation distance. The electrostatic coating apparatus of the invention includes, as the insulation distance enlarging portion of the spindle, the extended portion having an extended form to increase the creepage insulation distance between the rotary atomizing head or the speed increasing device to the electric motor. In addition, the apparatus includes, as the insulation distance enlarging portion of the fixed insulating member, the extended portion having an extended form to increase the creepage insulation distance between the rotary atomizing head or the speed increasing device to the electric motor. The presence of such extended portion can easily provide the long creepage insulation distance. The electric motor can be reliably insulated from the rotary atomizing head or the speed increasing device. 100 Electrostatic coating apparatus 110 Housing 116c Air ejecting port 116 Air ejecting section 120 Rotary atomizing head 125 Speed increasing device (High-voltage member) 130 AC servomotor (Electric motor) 130g Outer peripheral surface 140 Spindle (First insulating member) 141 Cylindrical portion 141kuk Rear-end-side portion (of a rear-end-side thin portion) (First extended part) (Insulation distance enlarging portion) 143 First zigzag portion (Insulation distance enlarging portion) 150 Fixed insulating member (Second insulating member) 151 Main body 153 Second zigzag portion (Insulation distance enlarging portion) 155 Second extended portion (Insulation distance enlarging portion) 160 Coating cartridge 165 Coating valve 170 Coating supply pipe 180 Air path KA Cooling air SA Shaping air A detailed description of a preferred embodiment of the present invention will now be given referring to the accompanying drawings. FIG. 1 shows an electrostatic coating apparatus 100 in this embodiment. FIG. 2 is a cross sectional view of the apparatus 100 taken along a line A-A in FIG. 1. FIG. 3 shows a front-end-side part of this electrostatic coating apparatus 100 in an enlarged view. FIG. 4 shows a spindle (a first insulating member) 140 of the electrostatic coating apparatus 100. FIG. 5 shows a fixed insulating member (a second insulating member) 150. This electrostatic coating apparatus 100 is mounted on an arm AM of a robot indicated by a broken line in FIG. 1 to perform electrostatic coating on a vehicle body (not shown) which is an object to be coated. In FIGS. 1, 3 to 5, the left side in each drawing is assumed as a front end side, the right side is assumed as a rear end side, the upper side is assumed as an upper side, and the lower side is assumed as a lower side. This electrostatic coating apparatus 100 includes a housing 110, a rotary atomizing head 120 placed closer to the front end side than the housing 110, and a speed increasing device (a high-voltage member) 125 mechanically connected to the rotary atomizing head 120 as shown in FIG. 1. The electrostatic coating apparatus 100 further includes an AC servomotor (an electric motor) serving as a driving source of the rotary atomizing head 120, and the spindle 140 placed through this AC servomotor 130 and mechanically connected to the speed increasing device 125. The electrostatic coating apparatus 100 further includes a fixed insulating member 150 fixedly placed between the speed increasing device 125 and the AC servomotor 130, a coating cartridge 160 filled with a coating material, and a coating valve 165. The housing 110 is made of insulating resin and has an opening 110c on the front end side in which a front end member 115 made of metal is fixedly mounted to close the opening 110c. This front end member 115 is provided with an air ejecting section 116 formed therethrough for communication between the outside and inside of the member 115. This air ejecting portion 116 includes an air ejecting port 116c through which shaping air SA is ejected out (leftward in FIG. 1). A rear end of this air ejecting portion 116 is communicated with an air path 180 mentioned later. Accordingly, when compressed air (in this embodiment, cooling air KA mentioned later) is supplied to the air ejecting portion 116 via the air path 180, the whole amount of the compressed air (the cooling air KA) is ejected out as the whole amount of the shaping air SA through the air ejecting port 116c. This front end member 115 is electrically connected to a high voltage cascade (a high-voltage generator) 119 placed on the lower side in the housing 110 through a high-voltage cable 118 arranged in the housing 110. This high-voltage cascade 119 is operated to generate electrostatic high voltage and apply it to the front end member 115. In use, therefore, the front end member 115 has a potential of about −90 kV. The rotary atomizing head 120 made of metal is rotatably attached to the front end side of the front end member 115. On the other hand, the speed increasing device 125 is placed on the rear end side of the front end member 115 and mechanically connected to the rotary atomizing head 120. The rotary atomizing head 120 is mechanically connected to the speed increasing device 125 as mentioned above. The speed increasing device 125 is mechanically connected at its rear end to the spindle 140 inserted through the AC servomotor 130 mentioned later. The rotary atomizing head 120 is therefore driven to rotate by rotation driving force of AC servomotor 130 through the speed increasing device 125 and the spindle 140. Furthermore, the front end member 115 is applied with electrostatic high voltage by the high-voltage cascade 119 as mentioned above. Since the speed increasing device 125 fixedly attached to the front end member 115 and the rotary atomizing head 120 connected to the speed increasing device 125 are made of metal, the speed increasing device 125 and the rotary atomizing head 120 are similarly applied with electrostatic high voltage and they have a potential of about −90 kV. The rotary atomizing head 120 is further connected at its radial center to a coating supply pipe 170 made of a SUS tube (see FIG. 2 in addition to FIGS. 1 and 3). The rotary atomizing head 120 is rotated at high speed (about 30000 revolutions per minute in this embodiment) by the AC servomotor 130 and the speed increasing device 125, thereby atomizing the fluid coating material supplied to the rotary atomizing head 120 through the coating supply pipe 170, by centrifugal force into fine particles, thus ejecting out the atomized coating material. At that time, the rotary atomizing head 120 is applied with electrostatic high voltage and the coating material supplied to the rotary atomizing head 120 is negatively charged. Accordingly, the vehicle body to be coated is relatively set at positive voltage (concretely, ground voltage) and subjected to coating. An electrostatic field is thus formed between the rotary atomizing head 120 and the vehicle body, so that the negatively charged atomized coating material can be efficiently coated on the vehicle body. The speed increasing device 125 has a publicly known configuration. Specifically, this speed increasing device 125 has a two-stage speed increasing mechanism including a front-stage planetary gear mechanism and a rear-stage planetary gear mechanism both not shown. An input shaft of the front-stage planetary gear mechanism is mechanically connected to the spindle 140 mentioned later. On the other hand, an output shaft of the rear-stage planetary gear mechanism is mechanically connected to the rotary atomizing head 120. Thus, the rotation driving force of the AC servomotor 130 is increased in speed in two stages by the front-stage planetary gear mechanism and the rear-stage planetary gear mechanism of the speed increasing device 125 and then is transmitted to the rotary atomizing head 120. The speed of the speed increasing device 125 in this embodiment is multiplied six times. Therefore, the number of revolutions of the AC servomotor 130 is set to 5000 rpm, the number of revolutions of the rotary atomizing head 120 can reach 30000 rpm required for atomization of the coating material. The AC servomotor 130 is placed in a predetermined position in the housing 110 on the rear end side than the speed increasing device 125. This AC servomotor 130 includes an outer peripheral surface 130g in a zigzag form having protrusions and recesses each extending circumferentially and arranged alternately in an axial direction (see FIG. 3). This outer peripheral surface 130g therefore has a larger surface area as compared with the case having no protrusions and recesses. In FIG. 1, for convenience of illustration, the protrusions and recesses are not shown. This AC servomotor 130 is electrically connected to a power supply circuit not shown through a power supply cable 133 and others. The AC servomotor 130 is driven to rotate by the electric power supplied from the power supply circuit. The AC servomotor 130 is connected to the outside through the power supply cable 133 and others and electrostatically grounded. In the AC servomotor 130, the spindle 140 is placed through a radial center thereof. This spindle 140 is integrally made of insulating resin. This spindle 140 has a cylindrical portion 141 extending in a cylindrical form from the front end side to the rear end side as additionally shown in FIG. 4. This cylindrical portion 141 includes a rear-end-side thin portion 141ku having a thin wall located on the rear end side than the axial center of the cylindrical portion 141, a thick portion 141w having a thick wall located on the front end side than the axial center, and a front-end-side thin portion 141su having a thin wall located on the front end side than the thick portion 141w. Of the rear-end-side thin portion 141ku, a front-end-side portion 141kus located on the front end side than the center of the thin portion 141ku is placed through the AC servomotor 130. On the other hand, a rear-end-side portion (a first extended portion (an insulation distance enlarging portion)) 141kuk located on the rear end side than the center of the thin portion 141ku extends from the AC servomotor 130 toward the rear end side. Of the thick portion 141w, a rear-end-side portion 141wk located on the rear end side than the center of the thick portion 141w is placed in the AC servomotor 130. On the other hand, a front-end-side portion 141ws located on the front end side than the center of the thick portion 141w extends from the AC servomotor 130 toward the front end side. A front end portion of the thick portion 141w is mechanically connected to the speed increasing device 125. Radially inside the cylindrical portion 141, a cylindrical resin pipe 173 made of insulating resin is placed with a gap from the cylindrical portion 141 (see FIGS. 1 to 3). This resin pipe 173 covers the coating supply pipe 170 for supplying a coating material to the rotary atomizing head 120 with no gap therebetween. Together with the cylindrical portion 141 of the spindle 140, the resin pipe 173 is to electrically insulate the AC servomotor 130 from electrostatic high voltage. In other words, the front end member 115 is applied with electrostatic high voltage by the high-voltage cascade 119 and the speed increasing device 125 and the rotary atomizing head 120 are also applied with electrostatic high voltage, as mentioned above, so that the rotary atomizing head 120 is similarly applied with electrostatic high voltage. Accordingly, the coating supply pipe 170 made of metal and placed through the inside of the AC servomotor 130 is also applied with electrostatic high voltage from the coating material and hence has a potential of about −90 kV. To electrically insulate the AC servomotor 130 from the coating supply pipe 170 applied with high voltage, consequently, the resin pipe 173 and the resin spindle 140 (the cylindrical portion 141) are arranged between the coating supply pipe 170 and the AC servomotor 130. Of the cylindrical portion 141 of the spindle 140, on the radially outer side of the front-end-side portion 141ws of the thick portion 141w, a first zigzag portion (an insulation distance enlarging portion) 143 having a zigzag comb-shaped cross section is provided as shown in FIG. 4. This first zigzag portion 143 has a disk portion 143a radially outwardly extending in a disk shape from the front-end-side portion 141ws of the thick portion 141w. The first zigzag portion 143 further has a 1-1 cylindrical portion 143b extending from a predetermined position on the radially inner side of the disk portion 143a toward the front end side and externally surrounding the front-end-side portion 141ws of the thick portion 141w in concentric fashion. The first zigzag portion 143 also has a 1-2 cylindrical portion 143c extending from a predetermined position of the disk portion 143a and externally surrounding the 1-1 cylindrical portion 143b in concentric fashion. Furthermore, the first zigzag portion 143 has a 1-3 cylindrical portion 143d extending from a predetermined position on the radially outer side of the disk portion 143a toward the front end side and externally surrounding the 1-2 cylindrical portion 143c in concentric fashion. In this embodiment, as above, the spindle 140 includes the first zigzag portion 143 and thus the creepage insulation distance is sufficient long between the speed increasing device 125 to which the electrostatic high voltage is applied and the AC servomotor 130. To be more concrete, a creepage insulation distance AB between a point A located on the rear end side of the speed increasing device 125 and a point B located on the front end side of the AC servomotor 130 is considerably long because of the presence of the first zigzag portion 143. Accordingly, creeping discharge from the speed increasing device 125 to the AC servomotor 130 can be prevented reliably and thus the AC servomotor 130 can be insulated reliably from the speed increasing device 125. In this embodiment, the rotary atomizing head 120 is placed apart on the further front end side relative to the speed increasing device 125 and therefore the AC servomotor 130 is also reliably insulated from the rotary atomizing head 120. Since the spindle 140 includes the rear-end-side portion (the first extended portion) 141kuk of the rear-end-side thin portion 141ku, the creepage insulation distance from the speed increasing device 125 to which electrostatic high voltage is applied to the AC servomotor 130 is sufficiently long. To be specific, a creepage insulation distance CD from a point C located on the rear end side of the speed increasing device 125 to a point D located on the rear end side of the AC servomotor 130, passing the inside of the AC servomotor 130, is considerably long because of the presence of the rear-end-side portion 141kuk. Accordingly, creeping discharge from the speed increasing device 125 to the AC servomotor 130 can be reliably prevented and thus the AC servomotor 130 can be surely insulated from the speed increasing device 125. The fixed insulating member 150 is placed between the AC servomotor 130 and the speed increasing device 125. This fixed insulating member 150 is integrally made of insulating resin. This fixed insulating member 150 has a substantially cylindrical main body 151 most of which is located between the AC servomotor 130 and the speed increasing device 125. The main body 151 contacts with the speed increasing device 125 on the front end side and contacts with the AC servomotor 130 on the rear end side. A second zigzag portion (an insulation distance enlarging portion) 153 having a zigzag comb-shaped cross section is provided on the radially inner side of the main body 151. This second zigzag portion 153 has a 2-1 cylindrical portion 153b extending from a predetermined position of the main body 151 toward the rear end side and surrounding the front-end-side portion 141ws of the thick portion 141w of the spindle 140 in concentric fashion. The second zigzag portion 153 also has a 2-2 cylindrical portion 153c extending from a predetermined position of the main body 151 and surrounding the 2-1 cylindrical portion 153b in concentric fashion. Furthermore, the second zigzag portion 153 has a 2-3 cylindrical portion 153d extending from a predetermined position of the main body 151 and surrounding the 2-2 cylindrical portion 153c in concentric fashion. The 2-1 cylindrical portion 153b of the second zigzag portion 153 is located on the radially outer side of the thick portion 141w of the spindle 140 and on the radially inner side of the 1-1 cylindrical portion 143b of the first zigzag portion 143 of the spindle 140 (see FIG. 4 as well as FIG. 5). The 2-2 cylindrical portion 153c of the second zigzag portion 153 is located on the radially outer side of the 1-1 cylindrical portion 143b of the first zigzag portion 143 and on the radially inner side of the 1-2 cylindrical portion 143c of the first zigzag portion 143. The 2-3 cylindrical portion 153d of the second zigzag portion 153 is located on the radially outer side of the 1-2 cylindrical portion 143c of the first zigzag portion 143 and on the radially inner side of the 1-3 cylindrical portion 143d of the first zigzag portion 143. The main body 151 is formed at its rear end with a second extended portion (an insulation distance enlarging portion) 155 having a cylindrical shape extending from the main body 151 toward the rear end side. This second extended portion 155 is located on the radially outer side of the outer peripheral surface 130g of the AC servomotor 130. In this embodiment, the fixed insulating member 150 includes the second zigzag portion 153 and thus the creepage insulation distance is sufficient long between the speed increasing device 125 to which the electrostatic high voltage is applied and the AC servomotor 130. To be more concrete, a creepage insulation distance EF between a point E located on the rear end side of the speed increasing device 125 and a point F located on the front end side of the AC servomotor 130 is considerably long because of the presence of the second zigzag portion 153. Accordingly, the AC servomotor 130 can be reliably insulated from the speed increasing device 125. In this embodiment, the rotary atomizing head 120 is placed on the further front end side relative to the speed increasing device 125 and therefore the AC servomotor 130 is also reliably insulated from the rotary atomizing head 120. Since the fixed insulating member 150 includes the second extended portion 155, the creepage insulation distance between the speed increasing device 125 to which electrostatic high voltage is applied and the AC servomotor 130 is sufficiently long. To be specific, a creepage insulation distance GH from a point G of the speed increasing device 125 to a point H of the AC servomotor 130 is considerably long because of the presence of the second extended portion 155. Accordingly, the AC servomotor 130 can be reliably insulated from the speed increasing device 125. Next, the air path 180 through which the cooling air KA passes will be explained (see FIGS. 1 and 3). This air path 180 includes a first path section 181 extending from the vicinity of the rear end of the outer peripheral surface 130g of the AC servomotor 130 toward the front end side along the outer peripheral surface 130g. In the housing 110, this first path section 181 is defined by an inner peripheral surface 111f of a housing cylindrical portion 111 surrounding the outer peripheral surface 130g of the AC servomotor 130. In the first path section 181, the outer peripheral surface 130g of the AC servomotor 130 is exposed. A rear end 181k of this first path section 181 is communicated to the outside of the electrostatic coating apparatus 100 through a path section not shown and connected to a pressure air source not shown placed outside. Accordingly, when the cooling air (compressed air) KA is supplied from the pressure air source to the air path 180, the cooling air KA flows through the first path section 181 from its rear end 181k toward a front end 181s. In this first path section 181, the outer peripheral surface 130g of the AC servomotor 130 having a jagged surface, providing a large surface area, is exposed. Accordingly, the AC servomotor 130 is more efficiently cooled by the cooling air KA. The air path 180 includes a second path section 183 continuous to the front end 181s of the first path section 181 and extending along the first path section 181 on the radially outer side thereof toward the rear end side. This second path section 183 is defined by the outer peripheral surface 111g of the housing cylindrical portion 111 of the housing 110 and an inner peripheral surface 115f of the second extended portion 155 of the fixed insulating member 150. The cooling air KA flowing through the first path section 181 while cooling the AC servomotor 130 then flows through the second path section 183 from its front end 183s to rear end 183k. Furthermore, the air path 180 has a third path section 185 located on the radially outer side than the second path section 183 and having one end continuous to the rear end 183k of the second path section 183 and the other end continuous to the air ejecting section 116. This third path section 185 is defined by the inner peripheral surface 111f of the housing 110 and the outer peripheral surface 150g of the fixed insulating member 150 and also by the inner surface 115f of the front end member 115 and the outer peripheral surface 125g of the speed increasing device 125. The cooling air KA having flowing through the second path section 183 then flows through the third path section 185 from its rear end 185k to front end 185s. The cooling air KA is thus supplied to the air ejecting section 116. Subsequently, the whole amount of this cooling air KA is ejected as the whole amount of the shaping air SA to the outside through the air ejecting port 116c. The electrostatic coating apparatus 100 further includes the coating cartridge 160 made of resin as shown in FIG. 1. This coating cartridge 160 is mounted in the housing 110 on the rear end side. This coating cartridge 160 is filled with a water-based coating material to be used for coating. A front end of this coating cartridge 160 is connected to a coating valve 165 made of metal and placed on the rear end side than the AC servomotor 130 in the housing 110. This coating valve 165 draws up the coating material from the coating cartridge 160 to supply the coating material to the rotary atomizing head 120 through the coating supply pipe 170. The front end member 115, the speed increasing device 125, and the rotary atomizing head 120 are applied with electrostatic high voltage by the high-voltage cascade 119 as mentioned above. Thus, the coating material supplied to the rotary atomizing head 120 is also applied with the electrostatic high voltage. This coating material is supplied to the rotary atomizing head 120 through the coating cartridge 160, the coating valve 165, and the coating supply pipe 170 as mentioned above. Accordingly, when the electrostatic high voltage is applied to the coating material, the electrostatic high voltage is also applied to the coating valve 165 and the coating supply pipe 170 both made of metal. Thus, each of the valve 165 and the pipe 170 has a potential of about −90 kV. However, since part of the housing 110 made of insulating resin is present between the coating valve 165 and the AC servomotor 130, the AC servomotor 130 is also reliably electrically insulated from the coating valve 165 to which the electrostatic high voltage is applied. As explained above, the electrostatic coating apparatus 100 in this embodiment includes the spindle 140 and the fixed insulating member 150 whereby the AC servomotor 130 is electrically insulated from the rotary atomizing head 120 and the speed increasing device 125. Accordingly, the electrostatic high voltage applied to the rotary atomizing head 120 and the speed increasing device 125 is not applied to the power supply circuit of the AC servomotor 130 therethrough. No burden is therefore imposed on the electric circuit. In addition, the spindle 140 includes the first zigzag portion 143 and the rear-end-side portion (the first extended portion) 141kuk of the rear-end-side thin portion 141ku as the insulation distance enlarging portion. This makes it possible to provide the long creepage insulation distances AB and CD between the speed increasing device 125 and the AC servomotor 130. Accordingly, the speed increasing device 125 and the AC servomotor 130 can be placed at a short distance in the electrostatic coating apparatus 100. The spindle 140 also can have a reduced size particularly in its axial direction, achieving the weight reduction. The electrostatic coating apparatus 100 can therefore be reduced in size and weight while providing reliable electric insulation of the AC servomotor 130 from the speed increasing device 125 to which the electrostatic high voltage is applied. The fixed insulating member 150 includes the second zigzag portion 153 and the second extended portion 155 as the insulation distance enlarging portion. This makes it possible to provide the long creepage insulation distances EF and GH between the speed increasing device 125 and the AC servomotor 130. Accordingly, the speed increasing device 125 and the AC servomotor 130 can be placed at a short distance in the electrostatic coating apparatus 100. The fixed insulating member 150 also can have a reduced size particularly in its axial direction, achieving the weight reduction. The electrostatic coating apparatus 100 can therefore be reduced in size and weight while providing reliable electric insulation of the AC servomotor 130 from the speed increasing device 125 to which the electrostatic high voltage is applied. In the present embodiment, the spindle 140 and the fixed insulating member 150 have the first zigzag portion 143, the rear-end-side portion (the first extended portion) 141kuk, the second zigzag portion 153, and the second extended portion 155 as the insulation distance enlarging portion. This makes it possible to easily provide the long creepage insulation distances AB, CD, EF, and GH, thereby reliably insulating the AC servomotor 130 from the speed increasing device 125. Furthermore, the present embodiment includes the speed increasing device 125 and therefore the number of revolutions of the AC servomotor 130 can be reduced just by the speed increased by the speed increasing device 125. To be concrete, the number of revolutions of the AC servomotor 130 can be reduced to 5000 revolutions per minute corresponding to one-sixth of the number of revolutions of the rotary atomizing head 120. Therefore, even though the spindle 140 is made of insulating resin lower in rigidity than metal and others, the spindle 140 is unlikely to be broken by the centrifugal force or the like. The present invention is explained along the above embodiment but is not limited thereto. The present invention may be embodied in other specific forms without departing from the essential characteristics thereof. |
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059178752 | description | FIG. 1 shows a nuclear fuel assembly 1 of the general type used in a pressurised water reactor. The fuel assembly 1 comprises a lower support member in the form of a bottom nozzle 2. At each corner the bottom nozzle 2 is provided with a foot 3 having a lower surface 4 which, in use, rests on the lower core plate of a nuclear reactor, and a recessed surface 4a. Provided in each of two diagonally opposite feet 3 is a locating hole 5 for the purpose hereinafter described. Extending upwardly from the bottom nozzle 2 are a number of control rod guide thimbles 6, the upper ends of which are connected to an upper support in the form of a top nozzle 7. Two locating holes 8 are provided in the upper surface of the top nozzle 7. At regularly spaced locations along the fuel assembly transverse spacer grids 9 are attached to and supported by the guide thimbles 6. Each spacer grid 9 comprises a cellular structure, formed by intersecting metal strips, in which a parallel array of fuel rods 10 are supported. An instrumentation tube 6aextends along a longitudinal central axis A of the assembly between the bottom and top nozzles 2, 7. Before installation in the core region of a nuclear reactor, it is necessary for a newly constructed fuel assembly 1 to satisfy certain stringent inspection procedures. These procedures include carrying out a full dimensional check, including measuring the channel spacing between adjacent fuel rods 10, a visual surface examination of the fuel rods, and obtaining the weight of the fuel assembly 1. It is particularly important that any tendency for the fuel assembly 1 to lean or tilt with respect to the vertical axis of the assembly, that is, to incline from the perpendicular, is detected and rectified before it is installed in the reactor core. These procedures are carried out automatically in accordance with the present invention by an automatic inspection station, generally designated by the numeral 11, as seen in FIGS. 2, 3 and 4. The inspection station 11 comprises a baseplate 12 and a top plate 13 which are interconnected by four vertical rectangular-section, columns 14, 15, 16, 17. A fuel assembly 1 to be inspected is positioned in the station 11 so that the bottom nozzle 2 is supported on a base assembly 18 and the top nozzle 7 is located by a top nozzle support 19. The top nozzle support 19 comprises a housing 20 of a generally hollow rectangular form which encompasses the column 16. Incorporated in the housing 20 is a stepper motor (not shown) by means of which the top nozzle support 19 can be raised and lowered. Projecting from the housing 20 is a top nozzle support bracket 22 which is movable by drive means (not shown) in a horizontal direction. Referring to FIG. 5, a fuel assembly 1 is shown in position beneath the support bracket 22 of the top nozzle support 19. The support bracket 22 is a fabricated structure comprising a lower plate 23 in which are mounted two guide housings 24 arranged on either side of a central rib 25. Each guide housing 24 has a recessed bore 26, extending downwardly from an upper surface of the housing, and a guide passage 27, extending from a base of the recessed bore 26 to the lower surface of the housing 24. Slidably arranged within the recessed bore 26 is a plunger 28 which is connected to a pneumatic cylinder 28 by means of an adaptor assembly 29. A fuel assembly locating pin 30, secured in the plunger 28, extends through the guide passage 27. At an end remote from the plunger 28 each locating pin 30 has a tapered region 31 leading to a reduced diameter end portion 32. As hereinafter described, the cylinders 28 can be actuated so as to either partially extend or fully extend the locating pins 30 along vertical axes from the retracted position shown in FIG. 5. When the locating pins 30 are partially extended the reduced diameter portion 32 projects in the locating holes 7 and when fully extended the larger diameter portion of the pins extend into the locating holes. The larger diameter of the locating pins 30 is substantially the same as the diameter of the locating holes 7. The bottom nozzle 2 is supported by the base assembly 18, which is illustrated in more detail in FIGS. 6, 7, 8 and 9. Incorporated in the base assembly 18 is a grinding facility for removing material from the lower surfaces 4 of the bottom nozzle 2. The base assembly 18 comprises a fixed mounting member 33 which has a horizontal base in the form of a circular flange 34 and a vertical hollow cylindrical post 35. The flange 34 is fixedly attached to an annular base 36 which is supported on the baseplate 12. A recess is machined in the upper surface of the flange 34 to accommodate a gear wheel 37 which is secured to the flange 34 by screws 38. Meshing with the gear wheel 37 are gear teeth 39 formed on the end of an output shaft 40 of a stepper motor 41. The stepper motor 41 is mounted on a bracket 42 secured to a rotor 43 arranged for rotation about the post 35. Thus, on operation of the stepper motor 41, the output shaft 40 rotates to effect rotation of the rotor 43 about the post 35. Rotational location of the rotor 43 on the post 35 is provided by an upper bearing sleeve 44 and a lower bearing sleeve 45 interposed between the rotor and the post. Axial thrust of the rotor 43 is borne by three equi-spaced air bearings each comprising a thrust plate 45a having an air pressure chamber 46 formed in an upper surface thereof adjacent to the lower surface of the rotor 43. Pressurised air is supplied to the chamber 46 by means of a pipe 47 and by passages formed in the flange 34 and the thrust plate 45a. The rotor 43 is of a generally cruciform cross-section when viewed in plan and comprises four radial protrusions 48, 49, 50, 51 as seen in FIG. 8. Formed in each of the diametrically opposite protrusions 49, 51 are an upper cylindrical recess 52 and a lower cylinder recess 53 of slightly smaller diameter than the upper recess 52. Fixedly mounted in the upper end of the upper recess 52 is a guide housing 54. A vertical passage extends through the guide housing 54 and accommodates a vertically slidable locating pin 55. The lower end of the locating pin 55 seats in a plunger 56 slidably arranged within the recess 52 and to which the pin is secured by a screw 57. Movement of the plunger 56, and hence the locating pin 55, is derived from a double-acting pneumatic actuator 58. The actuator 58 is secured to a transverse mounting plate 59 so as to extend vertically through the lower recess 53. A piston rod 60 extending from the actuator 58 is attached to the plunger 55 by means of an adaptor 61. By operation of the two actuators 58 the locating pins 55 can be Inserted into or retracted from the locating holes 4 provided in the feet 3 of the bottom nozzle 2. Appropriate sensors (not shown) are provided to indicate at a remote control location whether the locating pins 55 are in a retracted or withdrawn positions. Located between the radial protrusions 48, 49 is a grinding assembly 62 radially spaced from the rotational axis of the rotor 43 and which is used for removing material from the bottom surfaces 4 of the feet 3. The grinding assembly comprises a pneumatic motor 63 to which pressurised operating air is supplied through a pipe 64. Secured to an output shaft 65, projecting upwardly from the motor 63, is a grinding wheel 66. The grinding wheel 66 is clamped between a collar 67 and a clamp member 68 which is urged against the grinding wheel by a screw 69 threaded into the end of the output shaft 65. A sheet metal shroud 70 forms a waste collection chamber 71 which communicates with a suction means (not shown) for withdrawing swarf ground from the bottom nozzle feet 3. The motor 63 is fixedly mounted in a sleeve 72 which is arranged for axial movement within a recess 73 formed in the rotor 43. An extension piece 74, welded to the bottom end of the sleeve 72, has a central hole 75 formed with an internal acme screw thread. The hole 75 receives an output shaft 76 having an external acme screw thread corresponding to that formed in the extension piece 74. Thus, on operation of the stepper motor 77, rotation of the output shaft 75 causes an incremental axial movement of the extension piece 74 and the sleeve 72, thereby resulting in a corresponding incremental upward movement of the motor 63 and the grinding wheel 66. As will be described later, rotation and incremental upward movement of the grinding wheel 66 occurs during operation of the stepper motor 41 so that the grinding operation removes material from the feet 3 as the rotor 43 rotates about the post 35. Extending centrally through the base assembly 143 is a gimbal mounting arrangement 78 which is supported on a hydraulic/pneumatic intensifier 79 and a load cell 80. The gimbal mounting arrangement 78 comprises a hollow sleeve 81 arranged for longitudinal movement within the post 35. A reduced diameter portion substantially mid-way along the sleeve 81 defines an air bearing chamber 82. Pressurised air is supplied to the chamber 82 through a supply pipe 83 and a passage 84 to provide an air cushion between the sleeve and the post. Sealed to the lower end of the sleeve 81 is an annular connecting member 85 which is secured to the sleeve by means of a plate 86 and screws 87. The upper end of the sleeve 81 supports a gimbal mount 88 having a screw-threaded spigot 89 which is received in a correspondingly screw-threaded recess formed in the end of the sleeve. The gimbal mount 88 has a transverse upper end surface 90 which joins with the cylindrical side surface of the mount as a spherical male bearing portion 91. A tube 92 extends centrally through the sleeve 81 and has a lower end sealed to the connecting member 85 and an upper end sealed within a recess extending into the gimbal mount 88 from a lower end thereof. Seated on the gimbal mount 88 is a gimbal 94. A spherical female bearing cup 95 is formed in the lower surface of the gimbal 94 so as to correspond in shape to the spherical male bearing 91 provided in the gimbal mount 88. Preferably, the radius of the spherical bearing cup 95 is slightly larger than the radius of the spherical male bearing 91. Swivelling movement of the gimbal 94 on the gimbal mount 88 is limited by a screw 96 which extends through a clearance hole 97 centrally located in the gimbal and is threaded into the gimbal mount 88. A suitable lubricant for the spherical bearing is provided in the space formed between the end surface 90 of the gimbal mount 88 and the bearing cup 95 of the gimbal 94. A cover plate 98 extends over the top of the gimbal 94. Beneath the cover plate 98 two diametrically opposite cut-outs 99 are formed in the gimbal 94. Each cut-out 99 accommodates a probe 100 which senses the presence, through a hole 101 in the cover plate 98, of a bottom nozzle 2 when a fuel assembly is placed on the base assembly 18. Referring to FIG. 9, rotation of the rotor 43 can be prevented by the insertion of a shotbolt 102 into an aperture 103 formed in the rotor. The shotbolt 102 is slidably arranged in a guide hole 104 provided in the flange 34 of the mounting member 33. Movement of the shotbolt 102 is derived from an actuator 105 mounted on the annular base 36. A setscrew 106 attaches the shotbolt 102 to the actuator 105. A probe 107, extending into the flange 34, is used to sense the position of the shotbolt 102. When a fuel assembly 1 is installed in the inspection station, the lower surfaces 4 of two of the bottom nozzle feet 3 can rest on the upper surfaces 108 of the two radical protrusions 48, 50, and the lower surfaces 4 of the other two feet 3 rest on the upper surfaces 109 of the two guide housings 54. Each of the support surfaces 108, 109 lie in a common horizontal plane. A check on the dimensional accuracy of the fuel assembly 1 and a visual inspection of the surfaces of the assembly are carried out by operation of a measuring head assembly 110. The measuring head assembly 110 comprises a carriage 111 which is driven up and down the column 14 by a triplex chain 112 operated by a chain drive and tensioner unit 113. Mounted on the carriage 111 are four cameras for effecting the height and envelope measurements and for the surface inspection, and two channel spacing probes for measuring the spacing between each fuel rod 10. To balance the weight of the measuring head assembly 110 a counterbalance 114 is connected to the drive chain 112 and arranged to move along the column 15. The measuring head carriage 111 is shown in FIGS. 10 and 11 in which the measuring equipment has been removed for clarity. The carriage 111 is a fabricated structure comprising a bearing section 115 formed from upper and lower rims 116, 117 respectively. Each of the rims 116, 117 has four straight sides formed into a square cross-section which encompasses the column 14. Four bearing support plates 118 extend between the upper and lower rims 116, 117 each support plate being parallel to an adjacent side of the column 14. Two bearing assemblies 119 are provided in the upper region of each bearing support plate 118 and two further bearing assemblies 119 are provided in the lower region of each bearing support plate. Each bearing assembly 119 includes a roller 120 rotatable about an axis perpendicular to the side of the column 14 and arranged to run along the side surface 121 of a guide rail 122 secured to and extending along the length of the column 14. As seen in FIG. 11, a guide rail 122 extends along each of the sides of the column 14. The bearing assemblies 119 are arranged in pairs so that the rollers 120 of each pair run along opposite side surfaces 121 of the guide rails 122. Extending from the lower end of the bearing section 115 at right angles to the column 14 is a tray section 123 on which the measuring equipment is mounted. The tray section 123 comprises a flat plate 124 supported on an array of hollow-section ribs 125. A circular aperture 126, having a diameter such that it will allow a fuel assembly 1 to pass through, is provided in the plate 124. Secured to the upper surface of the plate 124 around the periphery thereof are four tee-sectioned slides 127, 128, 129, 130 along which the measuring instrumentation can be moved. Associated with each of the slides is a toothed rack 131, 132, 133, 134 secured to the sides of the plate 124. Gear teeth are provided on the lower surface of each rack for engagement by means for moving the measuring instrumentation along the slides, as hereinafter described. The chain 112 for raising and lowering the carriage is attached to the upper and lower ends of a chain mounting plate 135 provided at the base of the bearing section 115. A reader head (not shown) is provided on the carriage 111 for scanning a scale fixed to one of the guide rails 122, thereby enabling the vertical position of the carriage to be determined during measurement operations, as hereinafter described. Arranged for movement along each of the tee-shaped slides 127, 130 is a camera and probe assembly 140, as illustrated in FIGS. 12 and 13. Movement of the camera and probe assembly 140 along the respective slide 127, 130 is derived from a stepper motor 141. A gear pinion 142 mounted on the output shaft of the motor 141 meshes with a respective toothed rack 131, 134 secured to the carriage 111. The motor 141 is secured to a lower bearing housing 143 and is biased by a spring assembly 144 so as to urge the pinion 142 into meshing engagement with the associated toothed rack 131, 134. The bearing housing 143 straddles the horizontal portion of the slide 127, 130 and supports four bearing assemblies 145 and two roller assemblies 146 at each side thereof. Each of the eight bearing assemblies 145 includes a roller 147 arranged to run along either an upper or lower surface of the horizontal portion of the slide 127. Each of the two roller assemblies 146 on both sides of the housing 143 includes a roller 148 arranged to run along opposite vertical surfaces of the horizontal portion of the slide 127. A reader head (not shown) is provided for reading a scale provided adjacent to the toothed rack 131 to indicate the distance travelled by the camera and probe assembly along the slide 127. Signals corresponding to the distance travelled are transmitted to a remote control system. Extending transversely to the bearing housing 143 and fixedly secured thereto is a tube 149 of square cross-section. Secured within the tube 149 is a stepper motor 150 having an output shaft coupled to a lead screw 151. The lead screw 151 is received in a correspondingly threaded hole extending centrally along a rectangular-sectioned operating bar 152. Mounted on the end of the operating bar 152 remote from the motor 150 is a camera 153. Accurate linear movement of the operating bar 152 is provided by a number of bearing cassettes 154, two of which are secured in each side of the mounting tube 149. Each of the bearing cassettes 154 has a roller 155 which contacts a surface of the operating bar 152. On operation of the motor 150 the lead screw 151 is rotated, thereby causing linear movement of the operating bar and the camera 153. The amount of linear movement of the camera 153 is measured by a reader head 156 which scans a scale fixed to the operating bar 152. Electrical signals corresponding to the distance travelled are transmitted to the remote control system. Included in the two camera and probe assemblies 140 is a probe 157 which functions to check the channel spacings between adjacent fuel rods. Preferably the probe 157 incorporates a strain gauge bridge which converts mechanical deflection of the probe into corresponding electrical signals which are transmitted to the remote control system. Advancement of the probe 157 into the fuel assembly 1 and retraction therefrom is derived from a stepper motor 158, the output shaft of which carries a pinion 159. The pinion 159 is urged into meshing engagement with a fixed gear-toothed rack 160 by a spring device 161 acting on the motor 158. The motor 158 is mounted on an upper bearing housing 162 which is arranged for movement along a pair of guide bars 163 located on either side of the tube 149. Four bearing cassettes 164 are arranged in the top surface of the upper bearing housing 162, each cassette having a roller in contact with an upper horizontal surface of a respective guide bar 163. Two further bearing cassettes 164 are provided in each of the side surfaces of the upper bearing housing 162, each cassette having a roller in contact with the outer vertical surfaces of an associated guide bar 163. A bearing assembly 165 is also provided in each of the side surfaces of the upper bearing housing 162, each assembly having a roller in contact with a lower horizontal surface of a respective guide bar 163. The probe 157 is connected to the upper bearing housing 162 by means of a bracket 168 and an electromagnetic coupling assembly 169. Supported at the end of the tube 149 is a 3-point temperature-compensated calibration device 170 through which the probe passes prior to insertion in the fuel assembly 1. The calibration device 170 includes two plates 171, 172, the latter having three projections so as to define with the plate 171 three gaps of varying dimensions. The gaps increase progressively along the path of the probe 157 towards the fuel assembly 1 and are selected to provide three calibration measurements across the specific measuring range for a particular type of fuel assembly. If the probe 157 becomes stuck in the fuel rods of the fuel assembly it will be released automatically by the electromagnetic coupling assembly 169 thereby preventing possible damage to the fuel rods 10. Arranged for movement along each of the tee-shaped slides 128, 129 on the measuring head carriage 111 is a camera assembly 180, as seen in FIGS. 14 and 15. Movement of the camera assembly 180 along the respective slide 128, 129 is derived from a stepper motor 181. A gear pinion 182 mounted on the output shaft of the motor 181 meshes with a respective gear-toothed rack 132, 133 secured to the measuring head carriage 111. The motor 181 is secured to a pivotable mounting 183 supported on a pivot pin 184 fixed to a lower bearing housing 185. A spring assembly 186 acts on the mounting 183 so as to urge the pinion 182 into meshing engagement with the associated toothed rack 132, 133. The bearing housing 185 straddles the horizontal portion of the slides 128, 129 and supports four bearing assemblies 145 and two roller assemblies 146 at each side thereof. Each of the eight bearing assemblies 145 includes a roller 147 arranged to run along either an upper or lower surface of the horizontal portion of the slide 128, 129. Each of the two roller assemblies 146 on both sides of the housing 185 includes a roller 148 arranged to run along opposite vertical surfaces of the slide. A reader head 187 is provided for reading a scale provided adjacent the toothed rack 132, 133 to indicate the distance travelled by the camera 180 along the slide. Electrical signals corresponding to the distance travelled are transmitted to the remote control system. Welded to the lower bearing housing 185 is an upper bearing housing 188 which supports a movable guide bar 189 of square cross-section. Four bearing assemblies 190 are provided in an upper plate 191 of the upper bearing housing 188, two of these bearing assemblies having rollers 192 in contact with one vertical side of the guide bar 189 and the other two of the bearing assemblies having rollers 192 in contact with the other vertical side of the guide bar 189. Four further bearing assemblies 190 are provided in a side plate of the upper bearing housing 188, two of these bearing assemblies 190 having rollers 192 in contact with the upper horizontal surface of the guide bar 189 and the other two bearing assemblies having rollers 192 in contact with the lower horizontal surface of the guide bar. Secured to a vertical surface of the guide bar 189 by means of a fixture 193 is a gear toothed rack 194. A stepper motor 195 is mounted on the upper bearing housing 188 and drives a pinion 196 which meshes with the rack 194. Supported at the forward end of the guide bar 189 is a carrier 197 in which a camera 153 is mounted. Thus, operation of the motor 195 transmits a drive through the pinion 196 and the rack 194 to the guide bar 189 and the camera 153. Linear movement of the camera 153 is measured by a reader head 199 which scans a scale arranged to move with the guide bar 189. Electrical signals corresponding to the distance travelled by the camera are transmitted to the remote control system. Each of the four cameras 153 is thus movable along a respective slide 127, 128, 129, 130 and also in a direction perpendicular to the slides. The cameras 153 are operable in two modes: as measuring instruments to check the dimensions of the fuel assembly, or as means to carry out a visual inspection of the fuel rod surfaces. When used as measuring instruments, the cameras 153 operate as optical probes to obtain the focal length of surfaces viewed by the camera and corresponding information is transmitted to the remote control system. This information together with the signals transmitted by the reader head relating to the position of the camera is processed by a computer at the remote control position to determine the dimensions required. When operating as a surface inspection system, the cameras emit video signals to a video scanning system which combines the signals from all the cameras onto a single video recorder. The pictures obtained from each camera may be viewed on a monitor which may be divided into quadrants, each quadrant representing the picture from one of the cameras, thereby allowing all four sides of the fuel assembly to be viewed simultaneously. If desired, touch probes may be incorporated in the camera assemblies 180, whereby measurements of the fuel assembly are checked by probes which are moved into contact with the fuel assembly. The signals emitted by the touch probes when they contact the fuel assembly surfaces and the signals from the reader heads are processed by the computer to determine the required dimensions. The chain drive and tensioner unit 113 for raising and lowering the measuring head assembly 110 by means of the triplex chain 112 is illustrated in FIGS. 16, 17 and 18. Secured to the baseplate 12 is a drive unit support frame 210 to which a stepper motor 211 is attached by means of a mounting plate 212. An output shaft from the motor 211 is connected to a gearbox 213 by a coupling 214. The coupling 214 is mounted on a worm shaft 215 which drives a gear wheel mounted on a gearbox output shaft 216. A chain drive sprocket 217 is keyed to the output shaft 216 which is supported in a pedestal bearing 218 mounted on the support frame 210. Engaging the drive sprocket 217 is the triplex chain 112 which is connected to the measuring head carriage 111 and to counterweight 114. Tension is maintained in the chain 112 by a tensioner unit 219 arranged on the column 15. The tensioner unit 219 comprises a bracket 220 which encompasses and is frictionally clamped to the column 15. The bracket 220 has two side plates 221, 222 interconnected by two removable tie plates 223, 224 which are attached to the side plates by dome-head nut and screw assemblies 225. Each of the side plates 221, 222 and the tie plates 223, 224 extend substantially parallel to an adjacent side of the column 15. Extending through openings provided in the two side plates 223, 224 is an idler shaft 226 which is rotatably supported in bushes 227. Each of the bushes 227 is stationarily located in a bearing housing 228 secured to a respective side plate 221, 222. A chain tension sprocket 229 is mounted on the idler shaft 226 at an end thereof extending through the side plate 222. The tension sprocket 229 engages with the run of the triplex chain 112 at a position between the drive sprocket 217 and the counterweight 114. Connected to the end of the idler shaft 226 and extending through the side plate 221 is a universal coupling 230 which forms part of a manually-operated drive mechanism (not shown). This drive mechanism enables idler shaft 226 to be rotated manually so as to raise or lower the measuring head assembly 110 along the column 14. The bracket 220 frictionally engages the column 15 by means of upper and lower friction pads 231. As seen in FIG. 18, the pads 231 extend through apertures 232 formed in the side plate 222 to contact a surface 233 of a guide rail 234 secured to and extending along the column 15. Each of the pads 231 is resiliently urged into contact with the guide rail surface 233 by four springs 235. Each spring 235 acts on a set of nuts 236 screwed on a pin 237 which is secured to the side plate 222. Lateral location of the bracket 220 is provided by several slide plates 238, a pair of which is associated with each of the upper and lower friction pads 231. The slide plates 238 are located against side surfaces 239 of the guide rail 234. Each plate 238 can be adjusted by loosening two nuts to allow axial movement of two screws 240. In a similar manner, upper and lower friction pads 241 extend through apertures formed in the side plate 221. The friction pads 241 are rigidly secured to the side plate 221 and are drawn against a further guide rail 242, secured to and extending along the column 15, by the action of the springs 235. Associated with each of the friction pads 241 is a pair of lateral slide plates 243 which are adjustable in a manner similar to that shown in FIG. 18. To facilitate precise movement of the measuring head carriage 111 along the column 14 the weight of the measuring head assembly 110 is balanced by the counterweight 114. As seen in FIGS. 19 and 20, the counterweight 114 takes the form of a bracket 250 which encompasses the column 15. The bracket 250 has two bearing support plates 251, 252, interconnected by two removable tie plates 253, 254. The two support plates 251, 252 and the two tie plates 253, 254 define a rectangular opening through which the column 15 extends. A spring-loaded bearing assembly 255 is provided in the bearing support plate 251 and comprises two bearing housings 256 each of which supports a respective end of a stationary axle 257. Rotatably supported on the axle 257 are two laterally spaced rollers 258 which run along the surface 233 of the guide rail 242. The rollers 258 are urged against the surface 233 by four spring assemblies 259, two of which are associated with each of the housings 256. Each of the spring assemblies 259 comprises a stud 260 secured to the support plate 251 and extending through the housing 256. Coil springs 261 surrounding the studs 260 are pre-loaded by means of nuts 262 so as to act on the bearing housings 256 and urge the rollers 258 against the guide rail surface 233. Four further bearing assemblies 263 are mounted in the side plate 251, two at the upper end and two at the lower end thereof. Each of the bearing assemblies 263 comprises a support 264, which is rotatable within the side plate 251, and a roller 265. The bearing assemblies 263 are so disposed that the rollers 265 run along a respective side surface 239 of the guide rail 233. Since the axes of the rollers 265 are offset with respect to the rotational axis of the support 264, rotation of the support enables the rollers 265 to be brought into contact with the guide rail side surfaces 239. Upper and lower bearing assemblies 266 are mounted in the bearing support plate 252. Each of the bearing assemblies 266 is similar to the bearing assembly 255, except that the former are not resiliently mounted. Thus, each of the bearing assemblies 266 includes a pair of rollers 258 arranged to run along the outer surface 233 of the guide rail 234. Contact of the rollers 258 with the surface 233 is maintained by the reaction produced by the springs 261 associated with the bearing assembly 255 in the bearing support plate 251. Intermediate the upper and lower bearing assemblies 266 are a further pair of bearing assemblies 263, each including a roller 265 arranged to run along a side surface 239 of the guide rail 234. These bearing assemblies 263 are of a similar construction to the bearing assemblies 263 in the bearing support plate 251 and therefore the eccentrically mounted rollers 265 can be adjusted in a similar manner to bring them into contact with the guide rail side surfaces. Two extension plates 267, 268 are attached to the bearing support plate 251. An annular boss 269 is provided in the extension plate 267 for supporting a chain connector 270 which is secured in the boss 269 by a nut 271 and a locknut 272. Attached to the connector 270 are two end links of the drive chain 112. Casings 273, 274 formed on the top and bottom, respectively, of the extension plates 267, 268 each accommodates a lead weight 275. The weights of the two lead weights are selected so that the total weight of the counterbalance 114 matches precisely the weight of the measuring head assembly 110. As seen in FIG. 2, the run of the chain 112 attached to the measuring head assembly 110 extends from the drive sprocket 217 to a sprocket 280 mounted on the baseplate 12. From the sprocket 280 the chain 112 is attached to the carriage 111 (see FIG. 11) and is then trained around two further sprockets 281, 282 supported from the top plate 13. After passing around the sprocket 282 the chain 112 is connected to the counterbalance 114 (see FIGS. 19 and 20) and then extends to the driven sprocket 229 of the chain tensioner unit 219 which maintains the desired tension in the chain. Referring again to FIG. 2, the fuel assembly 1 is initially held in place by a pneumatic clamping assembly 300 arranged for movement along the column 16. The clamping assembly 300 comprises a carrier 301 which incorporates a plurality of bearing assemblies 302. Each bearing assembly 302 includes a roller (not shown) arranged to run along the side surfaces of guide rails 303 attached to the column 16. This arrangement is similar to that employed on the measuring head carriage 111, as previously described with reference to FIGS. 10 and 11. A clamping head 304 attached to the carrier 301 has four sides 305 which define a rectangular aperture through which the fuel assembly 1 extends. Secured to each of the four sides 305 is a pneumatic cylinder 306, the operation of which moves a clamping plate (not shown) towards and away from the sides of the fuel assembly 1. The driving arrangement and the counterbalance system for the clamping assembly 300 are similar to those employed for the measuring head carriage 111. Movement of the carrier 301 is effected by a triplex drive chain 307 which is driven by a drive sprocket 308. Rotation of the drive sprocket 308 is derived from a stepper motor and gearbox assembly 309. From the drive sprocket 308 the chain 307 extends upwardly around a tension sprocket 310 and then connects to a counterbalance 311 which is arranged to move along the column 17 (see FIG. 4). The counterbalance 311 is of a similar design to the counterbalance 114 (see FIGS. 19 and 20) employed for the measuring head assembly 110 and serves to counteract the weight of the clamping assembly 300. After connection to the counterbalance the chain 307 extends upwardly and is trained successively around two further sprockets 312, 313 supported by the top plate 13. The chain 307 then extends downwardly for connection to the carrier 301 and then finally around a sprocket 314 supported on the baseplate 12. Before the inspection station 11 can be used to carry out the inspection procedures on a fuel assembly it must be calibrated accurately in each of its moving axes for straightness, linear measurement and squareness between each of the axes. Calibration in the Z axis of each column, corresponding to the longitudinal direction of a fuel assembly, may be carried out using a laser interferometry technique so as to check straightness in two planes and linearity in the third plane. Calibration in the two Y axes, corresponding to the directions along two opposite sides of the fuel assembly, and the two X axes, corresponding in directions to the other two opposite sides of the fuel assembly, may be carried out with the aid of an accurately constructed granite artefact to check for squareness of each axis to the other. Before receipt of a fuel assembly 1 the various components of the inspection station 11 are positioned at their required initial locations. In particular, the measuring head carriage 111 is lowered down the column 14 by means of the chain 112, moved by the chain drive and tensioner unit 113, so that the camera and probe assemblies 140 are below the top of the base assembly 18. Each of the four cameras 153 and the two channel spacing probes 157 are in their fully retracted positions. With the clamping plates associated with the pneumatic cylinders 306 in the retracted positions, the carrier 301 of the clamping assembly 300 is lowered down the column 16 by the chain 307 which is driven by the stepper motor and gearbox assembly 309. The carrier 301 is located so that the clamping head 304 is positioned above the camera and probe assemblies 140 but below the top of the base assembly 18. The two top nozzle locating pins 30 in the top nozzle support 19 are fully retracted into the guide passages 27, as shown in FIG. 5. In the base assembly 18, see FIG. 6, the two locating pins 55 are extended above their respective guide housings 54 by operation of the pneumatic actuators 58. The gimbal 94 is elevated to its raised position by actuation of the hydraulic/pneumatic intensifier 79. The intensifier 79 acts on the plate 86 and the connecting member 85 to move the sleeve 81 through the post 35. This causes elevation of the gimbal mount 88, which, in turn, raised the gimbal 94. A fuel assembly handling system (not shown) brings a fuel assembly 1 from a fuel assembly construction apparatus and places the fuel assembly so that the bottom nozzle 2 rests on the raised gimbal 94. The gimbal 94 is raised sufficiently so that lower surfaces 4 of two diagonally opposite feet 3 of the bottom nozzle 2 are above the upper surfaces 108 of the two radial protrusions 48, 50 and the lower surfaces 4 of the other two diagonally opposite feet 3 are above the upper surfaces 109 of the two guide housings 54. Each of the locating pins 55 extends into a respective locating hole 5 provided in the feet 3. By operation of the stepper motor and gearbox assembly 309 the carrier 301 is then moved up the column 16 to a position substantially as shown in FIG. 2. Activation or the pneumatic cylinders 306 extends the associated clamping plates so as to engage and securely hold the fuel assembly 1. Sensors incorporated in the clamping assembly 300 confirm that the fuel assembly 1 has been secured, enabling the fuel assembly handling system to be withdrawn. By actuation of the pneumatic cylinders 28 the locating pins 30 are partially extended so that the reduced diameter portions 32 are located within the locating holes 7 provided in the top nozzle 6. The pneumatic cylinders 306 are then actuated to release the clamping plates from the fuel assembly 1. The weight of the fuel assembly 1 is transmitted to the load cell 80 through the gimbal mounting arrangement 78. Signals corresponding to the weight of the fuel assembly are transmitted by the load cell 80 to a computer system which records and stores the transmitted data. After recording and storing the fuel assembly weight data, the intensifier 79 is activated to lower the gimbal 94 thereby allowing the feet 3 to rest on the base assembly 18. Actuation of the pneumatic cylinders 28 is then effected so as to fully extend the locating pins 30 and to insert the maximum diameter of the pins into the top nozzle locating holes 7. The stepper motor and gearbox assembly 309 is operated so that the chain 307 moves the carrier 301 up the column 16 to a park position at which the sides 305 are located above the upper end of the fuel assembly 1. Operation of the chain drive and tensioner unit 113 is then initiated to lower the measuring head carriage 111 so that it is returned to its initial position. The inspection station 11 is now in a condition to carry out an envelope check on the fuel assembly 1 which involves obtaining external measurements of the assembly. To facilitate this procedure, datum points on the fuel assembly are designated. Thus, for measurements in the Z (vertical direction) of the fuel assembly the lower surfaces 4 of the bottom nozzle 2 serve as the Z datum points. For horizontal measurements, a corner of the bottom nozzle defines the XY datum point. Thus, measurements along the X axis represent horizontal measurements along two opposite faces of the assembly, and measurements along the Y axis represent measurements along the other two faces at right angles to the measurements along the X axis. To effect the envelope measurement check, the carriage 111 is raised so that the optical probe cameras 153 of the two camera and probe assemblies 140 (FIGS. 12, 13) and the two camera assemblies 180 (FIGS. 14, 15) are positioned so as to view the surfaces of the bottom nozzle 2. By operation of the stepper motors 141, the gear pinions 142 rotate and, by virtue of their meshing engagement with the respective toothed racks 131, 134, the camera and probe assemblies 140 are moved along their respective slides 131, 134. At specified measuring points, each of the motors 141 is stopped and the stepper motors 150 are operated so that rotation of the leadscrews 151 causes forward movement of the operating bars 152. Thus, cameras 153 are moved towards the bottom nozzle 2 until the bottom nozzle surface is in focus. The positions of the cameras 153 are known from signals emitted by a reader head (not shown) scanning a scale adjacent the toothed racks 131, 134 and by the reader heads 156 which scan the scales fixed to the operating bars 152. The signals from the reader heads are transmitted for storage at the remote control system where they are processed according to a dimensional check routine. Similarly, by operation of the stepper motors 181 the rotating gear pinions 182 move along the respective toothed racks 132, 133 so that the camera assemblies 180 are moved along their respective slides 128, 129. Each of the motors 181 is stopped at the required measuring point and then the stepper motors 195 are operated to move the guide bars 189 in a forward direction. This causes the cameras 153 to move towards the bottom nozzle 2 until the surface thereof is in focus. The positions of the cameras are known from signals emitted by the reader heads 187 which scan the scales adjacent to the toothed racks 132, 133 and by the reader heads 199 which scan the scales associated with the guide bars 189. The signals from the reader heads are transmitted for storage at the remote control system where they are processed according to a dimensional check routine. Each of the four cameras 153 can be operated to carry out a dimensional check at several spaced positions, typically at three positions, along the bottom nozzle surface. The chain drive and tensioner unit 113 is then operated so that the chain 112 raises the measuring head carriage 111 up the column 14. The carriage 111 is positioned so that the cameras 153 are able to view the surfaces of the fuel assembly top nozzle 1. A check on the dimensions of the top nozzle is then carried out in accordance with the procedure described above for the bottom nozzle. Dimensional checks for each of the spacer grids 9 are then carried out as described above for the top and bottom nozzles. After checking the XY (horizontal) dimensions of the fuel assembly 1, the measuring head carriage 111 is lowered to a position below the fuel assembly. The four cameras 153 are each moved to a position at one end of their respective slides 127, 128, 129, 130 and retracted away from the fuel assembly. Various measurements of the fuel assembly 1 in the Z (vertical) direction are then checked. This is effected by moving the carriage 111 along the column 14 to the desired position and then moving the cameras 153 horizontally along their respective slides 127, 128, 129, 130 to the required measurement positions. Movement of the carriage 111 along the column 14 is then carried out while using the cameras 153 to detect the upper and lower limits of the component surfaces. The reader head provided on the carriage 111 scans the scale fixed to a guide rail 122 on the column 14 to enable the vertical position of the carriage above the Z datum to be established. This information, in conjunction with the detection signals provided by the cameras 153, is processed by the remote control system to check the required Z-heights. The Z-heights obtained by the foregoing procedure can include the height of the upper surface of the bottom nozzle and the lower end of the fuel rods 10, the heights of the top edges of the spacer grids 9 above the Z-datum, the distance between the upper ends of the fuel rods 10 and the lower surface of the top nozzle 7, and the height of the upper surface of the top nozzle above the Z-datum. The data resulting from the above-mentioned dimensional checks are stored at the remote control system and compared with the data representing the desired measurements for the fuel assembly. After completion of the Z-height dimensional checks, the cameras are retracted to the initial position below the fuel assembly. The four cameras 153 can also be used to provide a visual check of the fuel rod surfaces, enabling surface defects, such as scratches or other marks, to be detected. This involves traversing the measuring head carriage 111 along the column 14 while focusing the cameras 153 on a pair of fuel rods 10. The carriage 111 is moved along the fuel assembly several times, the cameras being moved laterally each time so as to view a different pair of fuel rods. An output signal is taken from all of the cameras and combined onto a single video recorder for display on a colour monitor screen. The screen may be divided into quadrants, each quadrant representing the image viewed by one camera. From the screened image, the operator is able to detect surface imperfections on the fuel rods. Following the surface examination procedure the cameras 153 are retracted and the measuring head carriage 111 is returned to below the fuel assembly 1. The channel spacings between adjacent fuel rods 10 can be checked using the two probes 157 associated with the camera and probes assemblies 140, see FIGS. 12 and 13. This procedure entails moving the measuring head carriage 111 along the column 14 to the required vertical position and then operating the stepper motor 141 to move the camera and probe assemblies 140 along the slides 127, 130 to the first channel spacing to be checked. The stepper motors 158 are operated so as to advance the probes 157 through a respective calibration device 170. The probes 157 are then advanced through the fuel rod spacings to a selected depth into the fuel assembly. Electrical signals corresponding to the mechanical deflections of the probes 157 are transmitted to the remote control system where they are processed to enable the gap between adjacent fuel rods to be determined. This procedure is then repeated at all of the selected measuring positions until all of the required channel spacing measurements have been obtained. After all of the channel spacing checks have been completed, the cameras 153 and the channel spacing probes 157 are retracted and the measuring head carriage 111 is moved to below the bottom of the fuel assembly. To correct the tendency of a fuel assembly 1 to incline from the perpendicular, the following procedure is carried out. The two lower locating pins 55 are inserted into the bottom nozzle locating holes 5 and the two upper locating pins 30 are extended so that the larger diameter portion of each pin extends into the top nozzle locating holes 8. The fuel assembly 1 is now supported in a true vertical position. If the data obtained from the foregoing dimensional checks indicates that the fuel assembly tilts, ie that it is inclined to the perpendicular this requires rectification by removing the required amount of material from the lower surfaces 4 of the bottom nozzle. The tilt rectification procedure is initiated by the operator selecting a `grinding` option at a control computer. This procedure entails switching on two viewing cameras (not shown) arranged on either side of the base assembly 18. With the four cameras 153 and the two probes 157 in their retracted positions, the measuring head carriage 111 is raised to a position substantially mid-way along the fuel assembly 1. Operation of the stepper motor and gearbox assembly 309 lowers the clamping assembly 300 to a position substantially as shown in FIG. 2. Suction means (not shown) are then switched on to create a vacuum within the shroud 70 (see FIG. 7). Both of the two locating pins 55 are then retracted from the bottom nozzle locating holes 5 so as to be withdrawn into the guide housings 54. The gimbal 94 is then moved to a raised position by actuation of the hydraulic/pneumatic intensifier 79. As the gimbal 94 moves to the raised position it engages a recessed surface 4a of the lower support member 3. The fuel assembly 1 is therefore raised so that the lower surfaces 4 are elevated above the upper support surfaces 108, 109 of the rotor 43. Since the gimbal 94 is supported on the gimbal mount 88 by a spherical bearing, formed by the spherical female bearing cup 95 and the spherical male portion 91, compensation is made for any possible inclination of the recessed surface 4a. As movement of the gimbal 94 occurs along the longitudinal central axis A of the fuel assembly 1 and the top nozzle 7 is restrained by the locating pins 30, the fuel assembly 1 is retained truly vertical when in the raised position. Pressurised air is supplied through the supply pipe 64 to the pneumatic motor 63 to cause rotation of the grinding wheel 66. Referring to FIG. 9, the actuator 105 is operated so as to withdraw the shotbolt 102 from the aperture 103 in the rotor 43. The stepper motor 41 is then operated to rotate the output shaft 40 and the gear teeth 39 formed thereon which is in meshing engagement with the gear wheel 37. Since the gear wheel 37 is connected to the fixed mounting member 33, the gear teeth 39 move around the periphery of the gear wheel 37. The bracket 42 on which the stepper motor 41 is mounted is secured to the rotor 43 so that the latter rotates around the post 35 forming part of the mounting member 33. As a result, the grinding assembly 62 rotates about the central axis of the base assembly 18. The rotating grinding wheel 66 is raised in incremental steps by operation of the stepper motor 77 until contact is made with a lower surface 4 of a bottom nozzle foot 3. Such contact can be determined audibly or visually by viewing a monitor screen displaying images transmitted by the two viewing cameras. If the first pass does not achieve the removal of material from each of the lower surfaces 4, the grinding wheel 66 is raised by a further increment. This is achieved by rotating the output shaft 75 of the stepper motor 77 to cause an upward axial movement of the extension piece 74 and the sleeve 72 and, in consequence thereof, the motor 63 and the grinding wheel 66. Incremental movement of the grinding wheel 66 is continued until material is being removed by the grinding wheel 66 from the lower surfaces 4 of each of the bottom nozzle feet 3. This can be observed by the operator viewing a monitor screen which displays images transmitted by the two viewing cameras. Each incremental movement of the grinding wheel 66 may be by an amount set between 5-20 micron. Waste material removed from the bottom nozzle is removed from within the shroud 70 by the suction means. As a result of the grinding operation, the lower surfaces 4 of the bottom nozzle 2 will lie in a horizontal plane which is parallel to a horizontal plane containing the support surfaces 108 and 109. Thus, when the fuel assembly 1 stands on the bottom nozzle feet 3, the assembly should be truly perpendicular. The grinding operation is terminated by turning off the motor 63 and then retracting the grinding wheel 66. After stopping the stepper motor 41, the rotor 43 is secured by engaging the shotbolt 102 therewith and then the suction source is turned off. The locating pins 55 are raised so as to be inserted in the bottom nozzle locating holes 5. Lowering of the gimbal 94 allows the bottom nozzle lower surfaces 4 to rest on the support surfaces 108 and 109. With the four cameras 153 in their retracted positions, the measuring head carriage 111 is lowered to a position below the fuel assembly 1. After releasing the clamping plates on the clamping assembly 300, the latter is moved to a position above the fuel assembly 1. The two viewing cameras can now be turned off. If required, the procedure described above for checking whether the fuel assembly is inclined with respect to the perpendicular can be repeated and, if necessary, rectification can be carried out by removing additional material from the bottom nozzle 2. After completion of the inspection procedures, the fuel assembly 1 is held by the clamping assembly 300, and the bottom and top nozzle locating pins 55, 30, respectively are retracted. With the measuring head carriage 111 positioned below the fuel assembly, the fuel assembly handling system is operated to engage the assembly. The clamping assembly 300 is then de-activated and moved to a position below the fuel assembly 1 which can now be removed by the handling system. |
050948008 | description | SUMMARY OF THE INVENTION The press of the present invention is characterized primarily in that: the press to a large extent comprises components that can be remotely handled, with the press ram being divided into several press ram sections and the cover being divided into several cover portions; a displacement drive mechanism is provided for the longitudinal displacement of the press shaft between parallel tie rods that interconnect two crosspieces, one for supporting the hydraulic cylinder, which is a short press cylinder, and the other, oppositely disposed crosspiece supporting whichever of the counterpunch and transfer shaft that is present; the press shaft, at an end thereof that is disposed in the pressing direction, is provided with: a horizontal support surface for whichever of the counterpunch and transfer shaft that is present, vertical abutment surfaces for bracing a pressing force at the counterpunch, vertical abutment surfaces of a stop of the press shaft to brace one of the cover portions in the pressing direction, and vertical abutment surfaces for end face engagement against the transfer shaft. The advantages achieved with the present invention consist in the short overall length of the press, the good ability to remotely handle and conveniently remove the components, the compression process from which no particles can escape and which is achieved due to the longitudinal movability of the press shaft or line and the configuration of the counterpunch, and the non-restrained insertion and removal procedures for the counterpunch and for the transfer shaft. To increase the service life of the press, it is proposed pursuant to one specific embodiment of the present invention that the press shaft be provided with a replaceable U-shaped, upwardly open wear insert, and that to improve the remote handling capability, the U-shaped wear insert of the press shaft be comprised of several wear insert sections that are successively disposed in the pressing direction, with that wear insert section that is the last or trailing one in the pressing direction be positively connected with the press shaft via holding strips, and that wear insert section that is first or leading in the pressing direction is disposed ahead of a remotely operable displacement drive means with which all of the wear insert sections are brought into engagement against on another in such a way that no gap exists therebetween and the last wear insert section is brought to rest against the holding strips at the base and side walls of the press shaft. Also to increase the service life, in another specific embodiment of the present invention those surfaces of the cover sections that are directed toward the press shaft are provided with wear plates. In so doing, the cover section that is last or trailing in the pressing direction is positively connected with the wear plate via holding strips, and the first or leading cover section in the pressing direction is disposed ahead of a remotely operable displacement drive means with which prior to a pivoting and tightening of the cover bolts, all of the cover sections are brought to rest against one another in such a way that no gap exists between them, and the last section is brought to rest against the abutment surface of the stop of the press shaft. This prevents particles from becoming jammed between the sections, which could lead to increased wear of the press ram. In order with little expenditure of time to be able to undertake a closing and opening of the cover sections, and in order for insertion of shorter already preshortened fuel cell skeletons or other structural elements not to have to open the entire cover, it is provided pursuant to a further specific embodiment of the present invention that the cover sections, independently of one another, be secured to the upper side of the press shaft via hook-like cover bolts that are pivotably secured in pairs in crossbars and positively engage in recesses on the side walls of the press shaft, and that the cover bolts in a remotely operated manner are pivoted in pairs into the engagement position and are subsequently tightened via nuts. In order to keep the overall length of the press as short as possible, and in order to be able to reliably retract the press ram, which is divided into sections, out of the forward position, it is proposed pursuant to a further specific embodiment of the present invention that one or more extension rams be insertable between the cylinder ram, which is detachably secured to the piston rod of the press cylinder, and the guide ram, which rests against the material that is to be compressed, with these extension rams being positively interconnectable via connector pieces that can be placed in appropriate recesses. In order to achieve an automatic coupling together of the various press rams, it is finally proposed pursuant to the present invention that that end of the connector pieces that faces the press cylinder be provided on both sides with pivot pins that are disposed transverse to the pressing direction, with the underside of the opposite end that is directed in the pressing direction being provided with a hook-like projection that is provided with an inclined leading surface, with the connector pieces positively engaging in appropriate recesses on the upper side of the extension rams and the guide ram. Dividing the cover and the press ram into sections has the additional advantage that in order to compress short structural elements that do not have the length of the fuel cell skeleton, it is necessary to only partially open the cover, so that that cover section in the high compression region that is secured with a greater number of bolts can remain closed. The method of operating the inventive press comprises the steps of: during a loading process and during the compression process, disposing the press shaft in a forward end position, with the abutment surfaces of the side walls of the press shaft for bracing a pressing force at the counterpunch resting against facing abutment surfaces of the counterpunch, which rests on the horizontal support surface of the press shaft and has a back side braced against the other crosspiece, whereby in this position of the counterpunch an extension thereof that is directed toward the press shaft extends into the interior of the press shaft by a given distance to close off the cross-sectional area of the press in a substantially gap-free manner; after conclusion of the compression process, moving the press shaft to a rear end position, whereby a first gap results between an end face of the extension of the counterpunch and the vertical abutment surfaces of the press shaft for end face engagement against the transfer shaft, and a second gap results between a abutment surface of the counterpunch and the abutment surfaces of the press shaft for bracing a pressing force at the counterpunch, with these gaps providing free spaces for a non-restrained removal of the counterpunch; after removal of the counterpunch, placing the transfer shaft on the horizontal support surface and advancing the press shaft to such an extent in the pressing direction until those vertical abutment surfaces for end face engagement against the transfer shaft and of the one cover portion that are disposed in a plane rest against the immediately facing abutment surface of the transfer shaft and the transfer shaft is subsequently braced against the other crosspiece; further advancing that press ram section that is most to the front in the pressing direction to introduce the highly compressed pressed object into the transfer shaft, whereby this press ram section extends into the transfer shaft by a prescribed distance; and withdrawing this press ram section to again retract the press shaft to the rear end position, whereby an adequate spacing results for a non-restrained removal of the transfer shaft. Further specific features of the present invention will be described in detail subsequently. DESCRIPTION OF PREFERRED EMBODIMENTS Referring now to the drawings in detail, in the vertical cross-sectional view of the inventive power press illustrated in FIG. 1, the press shaft 3 is horizontally disposed on slide strips 13 of the transverse supports 12 of the mounting base 11 in such a way that the press shaft 3 can be longitudinally shifted in the direction of pressing. Secured to one end of the mounting base 11 is a crosspiece 6 against which the rear end of the press cylinder 5 is supported. At the opposite end of the mounting base 11, the crosspiece 8 is movable in the pressing direction. The crosspieces 6 and 8, as viewed in the pressing direction, are positively interconnected (see FIGS. 2 and 6) by two parallel tie rods 4 that are spaced from both sides of the press shaft 3; both ends of each tie rod 4 are provided with hammer heads 9. At that end 3b disposed in the pressing direction, the press shaft 3 is provided with a horizontal support or surface 3c on which, in the operating positions of the press illustrated in FIGS. 1, 2, and 3, rests the counterpunch 7, the back end of which is supported on the crosspiece 8. In the illustrated embodiment, the interior of the press shaft 3 is provided with an exchangeable or replaceable U-shaped wear insert 16 that is divided into sections 16a to 16f to facilitate removal thereof. The wear insert section that is disposed last in the pressing direction is positively connected with the press shaft 3 by holding strips 17 that are placed in grooves or notches. Via a displacement drive 18 (see FIG. 7) that acts in the pressing direction and that is supported on the press shaft 3, the wear insert sections 16a to 16f are brought to rest against one another with no gaps therebetween, and the wear insert section 16f is pushed against the holding strips 17. The press shaft 3 is closed off by the cover portions 2a to 2c, the surfaces 19 of which that face the upper side 27 of the press shaft 3 each being provided in the illustrated embodiment with wear plates 20a to 20c. The cover portions 2a to 2c are secured with hook-like cover bolts 24 that are pivotably received in pairs in crossbars 38, and that engage or mesh with recesses 25 of the side walls 22 of the press shaft 3. Via a central control 26 (see FIG. 8), and a linkage disposed within the crossbar 38, the cover bolts 24 are remotely operated in pairs so as to pivot into or out of engagement with the recesses 25, with the cover bolts 24 subsequently being tightened via nuts 39. With regard to the cover portion 2c , in the region of high compression the crossbars 38 are spaced close together, whereas in the remaining compression region of the cover portions 2a and 2b, the crossbars 38 are spaced further apart. The wear plate 20c is held fixedly in position on the cover portion 2c by the holding strip 36 to prevent shifting in the pressing direction. Prior to final tightening of the hook-shaped cover bolts 24, the cover portions 2a to 2c are brought to rest against one another without a gap therebetween via the displacement drives 23 that act upon the free end face of the cover portion 2a and are supported on the press shaft 3, whereby the free end face of the cover portion 2c is supported at the vertical abutment surface 3f against the stop 21 of the press shaft 3. The press ram, which is controlled by the press cylinder 5 via the piston rod 37, is divided into several press ram sections 1a to 1g, with the press ram section 1a (cylinder press ram) being connected to the free end of the piston rod 37 of the press cylinder 5 in such a wa that it is detachable via a plug connection and can be removed from above. The press ram section 1g (guide ram) serves for direct engagement against the structural material that is to be compressed. In order to reach the full stroke of the press ram up to production of the highly compressed pressed object 15, extension rams 1b to 1f of equal length are disposed between the cylinder ram 1a and the guide ram 1g. The maximum stroke of the piston rod 37 of the press cylinder 5 is slightly greater than the length of one of the extension rams 1b to 1f. This assures adequate free space for the insertion of a respective extension ram. The production of a highly compressed pressed object 15 is effected via the following steps: a) The empty press shaft 3 is moved to its rear end position, i.e. in the direction of the crosspiece 6, via the displacement drive 35, which is secured to the mounting base 11; PA1 b) After the counterpunch 7 is placed upon the horizontal support surface 3c, the press shaft 3 is moved by the displacement drive 35 in the direction toward the crosspiece 8 until the end faces 3d of the side walls 22 of the press shaft 3 abut against the abutment surfaces 7a that are disposed to the side on the counterpunch 7, and the counterpunch itself comes to rest against the crosspiece 8. In this position, the extension 10 of the counterpunch 7 extends by a specified distance T into the compression chamber; PA1 c) By means of the displacement drive means 18, the wear insert sections 16a to 16f are brought to rest against one another in such a way that no gap exists between them; PA1 d) The cylinder ram 1a and the directly abutting or engaging guide ram 1g are in their starting positions; PA1 e) After insertion of the fuel cell or element skeleton, the cover portions 2c, 2b, and 2a are placed on and are abutted against one another via the displacement drive 23, with the cover section 2c being brought to rest at the abutment surface 3f against the stop 21 of the press shaft 3. The hook-like cover bolts 24 that are pivotably mounted in the crossbars 38 are subsequently pivoted into the engagement position, and are tightened by activation of the nuts 39. The displacement drive means 23 is thereafter again moved into its starting position; PA1 f) After the first stroke of the press cylinder 5 has been effected to compress the fuel element skeleton, the piston rod 37, together with the cylinder ram 1a that is secured thereto, are retracted into the starting position. The cover 2a (loading cover) is removed after the tightening nut 39 have been loosened and the hook-like cover bolts 24 have been swung away; PA1 g) The extension ram 1f is disposed between the cylinder ram 1a and the guide ram 1g, which remains against the pressed object. As best shown in FIG. 9, a connector piece 28 is disposed in the recess 34a that is disposed on the upper side of the extension ram 1f. In particular, the connector piece 28 is pivotably mounted in the recess 34a via pivot pin means 29, 30 that extend to both sides transverse to the pressing direction. As the cylinder ram 1a is advanced, the bottom, inclined leading surface 33a of the connector piece 28 runs up on the bottom edge of the opening in the end of the recess 34b (in the guide ram 1g) until the hook-like projection 32 falls into the depression of the recess 34b, either due to the force of gravity or due to abutment by the inclined leading surface 33b that is on the upper side. PA1 h) The second stroke is subsequently effected and the cylinder ram 1a is again retracted into the starting position; the cover portion 2a (loading cover) can remain removed after the first stroke; PA1 i) In a similar manner, the extension rams 1e, 1d, and 1c are successively inserted, and the respective compression strokes are carried out; PA1 k) The last extension ram 1b that is to be inserted is additionally also connected to the cylinder ram 1a via a connector piece 28 in order after conclusion of the pressing process, and the still to be described ejection of the highly compressed pressed object 15, to be able to again retract the press ram sections 1b to 1f that are interconnected via connector pieces 28; PA1 l) After achieving the greatest operating pressure of the press cylinder 5, i.e. after a prescribed pressing stroke has been achieved, the pressing process is considered completed; the press cylinder 5 is hydraulically relieved, and the press shaft 3 is moved to its end position in the direction of the crosspiece 6 via its displacement drive 35, whereby the counterpunch 7, with its extension 10, receives a free space E (see FIG. 3) for removal thereof. In so doing, the counterpunch 7 is held securely in place against the crosspiece 8 via non-illustrated stops; PA1 m) After the counterpunch 7 has been removed, in place thereof a transfer or receiving shaft 14 is placed upon the support surface 3c. By means of the displacement drive 35, the press shaft 3 is advanced to such an extent in the direction toward the crosspiece 8 until those vertical abutment faces 3e, 3f of the press shaft 3 and cover portion 2c that are disposed in a plane rest against the immediate facing abutment surface 14a of the transfer shaft 14 and the latter is subsequently supported against the crosspiece 8; PA1 n) By further advancing the press ram 1g (guide ram), the highly compressed pressed object 15 is shoved into the transfer shaft 14, whereby the guide ram 1g projects by a prescribed distance U into the transfer shaft 14 (see FIG. 4); PA1 o) After retraction of the guide ram 1g, and after the press shaft 3 has been moved back in the direction toward the crosspiece 6, a sufficient distance V (see FIG. 5) results between the abutment faces 14a of the transfer shaft 14 and the abutment surfaces 3e, 3f of the press shaft for the non-restrained removal of the loaded transfer shaft; PA1 p) After the transfer shaft 14 that is loaded with the highly compressed pressed object 15 has been removed, the extension rams 1b to 1f are removed in a sequence opposite to that in which they were inserted, whereby the cylinder ram 1a is always respectively coupled via the connector piece 28 to the immediately adjacent extension ram. The counterpunch 7, the transfer shaft 14, the press ram sections 1a to 1g, the cover portions 2a-2c, the crossbars 38 with the cover bolts 24 that are pivotably mounted therein, as well as the U-shaped wear inserts 16a-16f, are provided with receiving openings to assure that they can be remotely handled via appropriate lifting mechanisms. The present invention is, of course, in no way restricted to the specific disclosure of the specification and drawings, but also encompasses any modifications within the scope of the appended claims. |
description | This application claims priority under 35 U.S.C. §119 of provisional application Ser. No. 61/022,174 filed Jan. 18, 2008 entitled: Suspended Radiation Protection for Protection of Worker. This invention relates in general to radiation protection and, more particularly, to a suspended personal radiation protection system. Radiation is used to perform many medical diagnostic and therapeutic tests and procedures. Medical, veterinary, or research personnel may be involved in the performance of these procedures. These professionals are being exposed to scattered radiation as they perform their work. The long-term effects of this exposure are poorly understood at the present time, but are considered serious enough to warrant mandatory protection for operators, who are required to wear garments or barriers that contain materials, which absorb a significant proportion of the radiation. In order to properly treat patients, operators require freedom of motion. Providing a personal radiation protection garment that properly protects operators, while allowing operators to move freely and comfortably, presents a significant challenge for medical operators in radiation environments. In accordance with the present invention, a method, a system, and an apparatus for implementing a suspended personal radiation protection solution are provided that substantially eliminate or reduce the disadvantages and problems associated with previous approaches. In accordance with one embodiment of the present invention, a system for offering radiation protection includes a garment that contours to an operator's body. The garment protects the operator from radiation. The garment is supported by a suspension component that reduces a portion of weight of the garment for the operator, the garment including a belt, which includes a release mechanism that offers an entry into the garment. In more specific embodiments, the release mechanism is a quick release that allows the operator to disengage from the garment using a single hand movement. The belt can include at least one flexible joint. The belt opens to allow the operator to enter the garment, and the operator, in entering and exiting the garment, is able to limit his contact to components on or near a front of the garment such that the operator can operate the release mechanism for the garment without losing sterility. In still other embodiments, the release mechanism includes a spring mechanism that exerts a force on the belt. The garment allows the operator, who is wearing the garment, to move freely in X, Y, and Z spatial planes, and the garment can be substantially weightless to the operator. Further, the garment may include a sleeve on at least one side of the garment. The garment can include a rapid, easy closure of the belt around the operator, potentially accomplished by manually squeezing the belt itself with the operator's arms and, thereby, closing it at its hinges and mobile joints. Alternatively, closure may occur when a cable arrangement or mechanical linkage is activated which draws the hinges closed. The easy opening of the belt enables an ideal exit for the operator: allowing the operator to only touch components on (or near) the front of the device in order to preserve sterility (i.e., the sterile condition of the environment). Typically, operator contact towards the rear of the device would result in a loss of sterility of the hands. The simple mechanisms of the garment require minimal gross hand and finger movements so that these operations can be accomplished through the sterile cover without unnecessary fumbling. Such a device offers automation for the opening of the belt such that the operator need only activate a given quick release to exit the garment. The other motions may be automated and triggered by this initial activation. This feature could be accomplished by a spring mechanism (wire or gas springs) or similar tensioning mechanism that exerts a force on the belt, which results in opening. The suspended personal radiation protection device includes a suspension component. The suspension component empowers the operator to move freely in the X, Y, and Z spatial planes simultaneously, while the protective radiation garment is substantially weightless to the operator. The suspension component is further operable to support a partial weight of the operator such that the operator can move around in substantially zero gravity, or (at a minimum) the operator bears just a portion of his total weight. The radiation protection device further includes an optional face shield, which is transparent to visible light: allowing unhindered vision while protecting an operator from radiation. Other unique components of the garment are detailed below. Important technical advantages of certain embodiments of the present invention include optimally supporting the weight of a radiation protection garment worn by operators. Ironically, the suspension component allows radiation protection garments to be heavier and more comprehensive. As a result, radiation protection garments can protect larger areas of operator's body and be constructed thicker to increase X-ray attenuation. Increased radiation protection reduces an operator's risk of cancers, cataracts, skin damage, etc. Other important technical advantages of certain embodiments of the present invention include reducing the risk and incidence of musculoskeletal injuries from wearing heavy radiation protection garments. Operators using the present invention have normal freedom of motion, as if the operator is not wearing heavy material. Furthermore, the suspension device supports a majority of the operator's weight such that operators can work for longer periods without fatigue. Other technical advantages of the present invention will be readily apparent to one skilled in the art from the following figures, descriptions, and claims. Moreover, while specific advantages have been enumerated above, various embodiments may include all, some, or none of the enumerated advantages. For purposes of teaching and discussion, it is useful to provide some overview as to the way in which the following invention operates. The following foundational information may be viewed as a basis from which the present invention may be properly explained. Such information is offered earnestly for purposes of explanation only and, accordingly, should not be construed in any way to limit the broad scope of the present invention and its potential applications. Radiation is used to perform many medical diagnostic and therapeutic tests and procedures. The human patient or animal is subjected to radiation using minimal doses to enable completion of the medical task. Exposures to radiation are monitored to prevent or reduce risks of significant damage. Medical, veterinary, or research personnel may be involved in the performance of such procedures in great numbers. Over many years, these professionals are being exposed to scattered radiation as they perform their work. Although their daily exposure is generally less than that for the patient, there are adverse cumulative effects to the operators. These long-term effects are poorly understood but are considered serious enough to warrant mandatory protection to workers in the form of garments or barriers that absorb a significant proportion of the radiation. There is a wide variety of such barriers commercially available, but these solutions have significant limitations for the operators who must come in close contact with the subject. These operators may be physicians and their assistants, or technically skilled medical personnel, who perform simple or complex medical procedures using their bodies and hands in proximity of the patient. In many cases, scatter radiation from the subject or physical elements in the direct radiation beam will pose significant health risks and unacceptably high exposure. Risks of radiation exposure at the levels of medical personnel include cancers, cataracts, skin damage, etc. A review of current protective systems outlines their limitations. Radiation-absorbing walls are useful to contain the radiation to a room, but do not prevent exposures within their confines. Barriers within the room [such as floor or ceiling supported shields] are effective at blocking radiation for personnel who are not in close contact with the radiation field [such as some nurses and technologists] but must be positioned or repositioned frequently when personnel move around the room. They also provide cumbersome interference for operators performing the actual medical procedure. They may also be difficult to keep sterile when attempting to use them within the sterile field. The most commonly used protection for operators involves the use of garments containing radiation-absorbing materials, generally lead or other metals, which are worn in the fashion of a coat, smock, skirt, vest, etc. and do not contaminate the sterile field because they are worn underneath the sterile covering gown. These garments are heavy and uncomfortable, and their long-term usage is known to be associated with diseases of the spine [in the neck and back], knee disorders, and other musculoskeletal problems, which can result in disability, medical expenses, and decreased quality of life for the operator. The trade-off between protection and garment weight results in the frequent use of garments that do not cover the legs, head, torso, and eyes optimally, and may provide sub-optimal radiation protection due to the thickness of the metallic material being limited by the tolerability of the operator. To protect other radiation-sensitive tissues [such as the corneas of the eye and the thyroid], special heavy glasses containing metallic compounds and a collar around the neck are often worn. Even when the operator is encumbered with these items, the base of the skull [which may contain sensitive bone marrow] and the face are still unprotected. Personal face and neck shields address this problem, and are commercially available, but are rarely worn due to their cumbersome nature and heavy weight. Such problems have been present for many years and there are current solutions that attempt to address them. Modifications to floor-supported mobile shields appear to attempt to provide improved dexterity for the operator relative to the standard bulky mobile barrier, and a floor support system with a modified garment design also attempts the same. However, they are still obstacles to free movement of the operator. Another system of barriers (such as those referred to as radio-protective cabins) around the patient has been proposed, but that appears cumbersome, confining, and inhibitory to operator movement both gross and fine, patient/subject contact, and sterile field operation. Ceiling mounted barriers around the patient also appear to limit contact between patient and operator, and may make control of a sterile field difficult. One configuration includes a ceiling mounted device, which supports the weight of a lead garment, involving a dolly movable in one linear axis, with or without an extension arm that rotates around a central point on the dolly. Such mechanical configurations are in place for other types of suspended barriers and their motion mechanics may not be well suited for use with something attached to the operator's body since the operator must frequently move rapidly and freely in all three spatial axes. Typically, the operator will walk in unpredictable and rapid patterns over an operating area. One configuration includes the garment being suspended by a simple expansion spring, which will provide uneven forces depending on its degree of expansion occurring with operator motion [due to the nature of its simple spring mechanics]. It may also result in harmonic motions that affect operator dexterity. In addition, failure of the spring due to cycle stresses could lead to operator injury. In addition, location of the spring in a vertical direction above the operator could result in limitations due to ceiling height. Integration of the system with the heavy image intensifier monitor screen could further encumber the operator from normal motion. A discussion of the types of motion performed by operators during their work is relevant. Operators generally stand next to an operating table where the patient is positioned. They often reach over the patient to various parts of the body, and they may lean forward while reaching for items, surfaces, etc. This puts stress on the spine when heavy garments are worn. They may bend or stoop, but rarely is this possible because the workspace containing the patient limits vertical motion. In addition, most procedures involve a sterile field where the operator's hands, arms, and torso [from neck to waist] must remain confined, so excessive vertical motion is prohibited. The operator may move considerably in the X and Y plane, which is horizontal and parallel to the floor, by walking or turning their body. The operator requires freedom of motion in these directions. Overhead cranes have been available for many years and are commonly employed in the materials-handling industry. The following is a description of a bridge crane. A bridge crane includes at least one bridge, and a trolley moving on the bridge, end trucks arranged at the ends of the main bridge to support the main bridge, wheels arranged to the end carriages intended to move along substantially parallel rails substantially parallel to the end trucks. Smaller cranes [such as those to be used to support a load up to 250 pounds] are often operated by workers without the aid of motorized assistance because the crane's movable parts are light enough to be manipulated by hand. Different systems are employed to suspend the load from the cranes, including hoists, balancers, and intelligent assist devices. Tool balancers are also currently available and help to suspend tools in the workspace in a manner that provides ergonomic benefits for workers using them. The tool balancer is generally attached over the workspace, and reels out cable from which the tool is suspended. Adjustments may be made to provide a “zero gravity” balancing of the tool at the desired height such that the worker may move the tool up or down within a working range without having to bear a significant portion of the tool's weight. Adjustments may cause the tool balancer to exert a stronger upward force such that the operator must apply a downward force on the tool to pull it down to the workspace and the balancer will cause the tool to rise when the operator releases it. Tool balancers may be of the spring or pneumatic variety, referring to the mechanism, which provides the force for its operation. A spring tool balancer, such as in the preferred embodiment of this invention, generally contains a coiled flat spring [similar to a clock spring], which is attached to a reel with a conical shape and which serves as the platform for the winding of the cable. The conical shape provides a variable mechanical advantage, which offsets the variance of the force provided by the spring as it winds or unwinds. The result is a relatively constant force on the cable within a definable working range. Safety concerns mainly involve falling objects, strength of the suspension device, strength of the cable, and operator falls. The balancer can be attached to the trolley by its own hook and a safety chain. The suspension device is commercially available at specified maximum loads, which include a wide safety margin. The mounting of the suspension device will be done according to architectural standards. Detachment of the garment from the suspension system will require certain care. A cable stop will prevent the hanger from going higher than the set level. Some balancers are equipped with a locking mechanism that prevents motion of the cable during load change or removal. This permits simple removal or exchange while standing at ground level. Alternatively, without activating a locking mechanism, The worker could stand on a step stool and lift the load upwards until it contacts the balancer stop, and then remove the garment without concern for sudden upwards, uncontrolled motion of the balancer cable and hanger. Alternatively, a weight, which is approximately equivalent to the weight of the garment, could be attached to the hanger prior to disengaging the garment. This will drop the garment and require it to be supported by the worker, who may then disengage it from the hanger. The weight will prevent any upward motion of the hanger in an uncontrolled manner. The next time the garment is attached, the weight could be removed after secure attachment of the garment is confirmed. For most operations, the garment need not be detached from the cable. It could be left suspended and moved out of the way of other activities. Another alternative method would involve setting the force on the balancer to be slightly greater than the weight of the garment. Once removed from the body, the garment would then slowly and safely rise up until stopped by the cable stop. Upon next use, it could easily be pulled back down into position. Annual inspections of the system may be performed for cable frays, hook lock malfunctions, and rail component flaws. In the event of an operator fall, it is unlikely that the system will contribute to operator harm since the balancer cable is long enough to allow the operator to reach the floor. Any harm to the operator should be the same as if not attached to the cable, except perhaps for some beneficial effect of the upward force of the suspension system. In the event of spring breakage, most balancers are equipped with automatic cable locking mechanisms to prevent dropping of the load. In the event of cable or fastener breakage, the frame of the garment/hanger may be designed such that there are pads over the shoulders of the operator which would gently engage the operator's shoulders to support the weight of the device, which is approximately equivalent to a moderately heavy backpack. This latter malfunction should be avoidable with adequate cable and fastener strength and annual inspections. In the event that rapid detachment of the operator from the system is necessary due to emergency, this can be achieved by a simple removal of the garment from the body without detachment from the system. The garment can be left hanging, and the suspended garment can be moved to the end of the runway, clear of the moving patient or stretcher. Turning back now to the general problem of radiation, it is evident that people are often exposed to radiation in the course of their work. The proposed concept, outlined herein, describes a device and technique intended to address many of the aforementioned problems. It provides extensive shielding for the operator: covering a large part of the body. The shielding capacity can be increased with thicker, heavy metal layering, thus reducing a dose to the operator because the device is weightless [or nearly so] to the operator. The device is close to the body of the operator, much like a conventional apron, however it is not supported by the operator. It moves with the operator as he/she moves around within the working field and sterile field, and allows movement of arms and body parts to accomplish the procedure at hand. The overall effects of the device are: improved comfort for the operator who is no longer supporting heavy-shielding clothing, improved radiation protection to an operator through a much greater portion of body shielding [compared to a conventional apron], as well as more effective shielding of much of the covered parts due to greater use of the shielding material. This approach also offers a musculoskeletal benefit due to the absence of a significant weight burden on the operator. FIG. 1 is a simplified block diagram of a suspended personal radiation protection system 10. System 10 includes an operator, a patient, a radiation source 16, radiation rays, a component (detailed in subsequent FIGURES), and a release 20 for the personal radiation protection garment. The garment includes a face shield 12, an outer apron 14, and a sleeve element 18. Each of these components is outlined in greater detail below with respective FIGURES that further highlight some of their potential intricacies and capabilities. In general, the garment and shield 12 are suspended from a hanger, which is supported by a given suspension component. [Other example suspension components are outlined below with reference to FIGS. 5A-5L.] FIGS. 2-4 further illustrate this garment architecture from side and front views. An operator can position himself in the garment such that the operator is not supporting the weight of the garment. In this sense, he is liberated from the typical and problematic weight constraint. While using radiation to treat a given patient, the operator can move freely in the X, Y, and Z spatial planes such that the garment and shield 12 are substantially weightless. (Note that U.S. patent application Ser. No. 11/611,627 entitled “System and Method for Implementing a Suspended Personal Radiation Protection System” is hereby incorporated by reference herein.) In accordance with the teachings of the present invention, suspended personal radiation protection system 10 achieves an effective way for operators to protect themselves properly and comfortably from harmful radiation. System 10 consists of a framework of rigid components (such as steel for example) with some components allowing motion in various types of joints. Such design choices permit the support of a pliable component such as fabric containing heavy metals to absorb radiation. The absorption materials are positioned close to the operator and are in the pathway of scattered radiation. Also provided to the garment is an optimal face shield 12 that offers an optically transparent (or nearly transparent) component (such as leaded glass or acrylic). The shield is proximate to the operator's face, neck, and head, but distant enough to reduce potential fogging. FIGS. 5A-5L illustrate how the garment can be supported by an overhead structure such as the ceiling, a floor-supported frame, a telescopic arm, or a table supported frame. Specifically, FIGS. 5A-5B illustrate an overhead support structure 32, while FIGS. 5C-5D depict trolleys 34 and 36 that can slide on a given set of rails. FIG. 5E illustrates a support 37 that offers a trolley that may slide along rails affixed to the ceiling. The trolley contains an articulating arm that may extend beyond the confines of the ceiling mounted structures. FIG. 5F illustrates a jib crane 38 where the trolley slides on an arm, which rotates on the floor mounted pivot stand. This allows positioning under a wide arc. FIG. 5G illustrates a support 39 that offers an articulating bridge crane. The bridge moves along stationary rails in the ceiling and the trolley moves along the bridge. In addition, the articulating arm permits extension outside of the ceiling mounted rails. FIG. 5H illustrates a support 40 that offers a jib crane. The trolley slides on a swinging arm in this embodiment, which is wall mounted. Such an arrangement can readily be ceiling mounted and it too allows positioning under a wide arc. FIG. 5I illustrates a support 41 that is a reaction arm. This could be power-assisted and absorb shock and torsion forces, if necessary. FIG. 5J illustrates a support 43 that offers a hanger and swivel pivot for adjustment of the center of gravity. In this embodiment, there is a screw provided to tighten the rod. Note that the wire rope is slidable and lockable. This telescopic configuration allows placement of the suspension point of the wire cable in any position under a large arc. This further allows balance of the whole device in any way desired. For purposes of clarification, a few terms are outlined here to assist the readers in understanding some of the following descriptions. Suspension means or ‘suspension component’ can include a crane, ceiling mounted mechanisms, wire ropes, spring balancers, wire ropes, etc. This all leads to a wire rope that connects to a hanger. The hanger hangs from the suspension component by the wire rope of the spring balancer, in those embodiments employing a spring balancer, and is integrated into the device frame that sits on the shoulders and chest and that holds up the shield and apron, contains the belt, etc. In some of the tendered FIGURES, the hanger arcs from a position over the top of the head, down to the rest of the frame. The device frame is the skeleton that contours around the shoulders and chest and torso and contains the belt mechanism. The frame supports the apron and shield and is integrated rigidly or with an articulation with the hanger. In an alternative embodiment, the hanger is not a rigid rod, but instead two wire ropes that connect to the frame and support it. These two wire ropes are suspended from a horizontal bar, which may have an adjustor on it that is attached to the suspension component (wire rope). The term ‘balancer’ refers to [typically] a spring balancer. This is the zero-gravity support device that is integral to the suspension component and is what gives freedom of motion in the Z axis [while supporting the weight of the hanger/frame/rest of the device]. The hanger adjustor is one of many types of devices that permit balancing of the device in a different way than the “balancer.” The device hangs from the wire rope like a mobile, so moving the point of attachment around in a plane horizontal to the floor within a small area over the operator's head will change how the device is oriented in space [i.e., how it tilts right to left or front to back]. The “hanger” is the portion of the frame of the device that arcs up over the operator's head, and holds up the apron and shield. In one example outlined in FIG. 5J, the object is on the top of the hanger, right over the head. The cable that comes down from the balancer attaches to it. FIGS. 5A-I represent suspension components (for example “cranes” for the device) and related apparatus. These architectures shown (i.e., FIGS. 5E-5I) are alternatives to simple “bridge crane” scenarios, representing either non-bridge crane configurations with or without articulating arms, or in one case (FIG. 5G), a bridge crane with addition of an articulating arm. Another embodiment uses a bridge crane with a telescoping bridge that may reach further outside of the confines of the rails to allow more freedom of motion. Another example includes a framework similar to these cranes that is mounted on a floor-mounted framework instead of the ceiling or walls. A spring balancer may hang from the tolley on the bridge. Attached to the end of the wire rope of the spring balancer may be the “hanger” portion of the device. The hanger may be a rigid structure that arcs over the operator's head, down to the frame that is approximated near his shoulders and torso and supports the face shield and body apron. At the top of this hanger, the attachment to the wire rope from the spring balancer may occasionally require adjustment in the X-Y plane in order to keep the device properly oriented in space, without tilting left to right or front to back. This adjustment may be made using several possible devices. Turning now to FIGS. 5J-5L specifically, the device in FIG. 5J shows an attachment that allows pivoting of the horizontal bar of the hanger with slidable motion of the collar that attaches to the wire rope of the suspension system. This allows positioning of the suspension point within an arc above the hanging device, allowing positioning of the device in its neutral hanging position. FIG. 5K shows a system 49 with components slidable in orthogonal directions (X and Y axes) allowing positioning of the point of attachment of the wire rope of the spring balancer or other suspension system in any position within a defined rectangular plane, resulting in the ability to tilt the suspended hanger and frame/apron/shield as needed for fit and comfort. In actual use, the square and round rods could be oriented in a plane parallel to the floor. [Note that, as explained herein, the orientation of FIGS. 5K and 5L are illustrated slightly different than some of the included descriptions, in that the round rods are oriented vertically. Typically, these rods would be oriented horizontally, although the components are adequate. A person could just rotate them 90-degrees to put them in the same orientation as all the other FIGURES. This is just another example implementation of the possible alternatives encompassed within the present invention.] FIG. 5L shows a similar system 51 with the addition of ball screw or similar threaded rod and nut arrangement that permits easy and precise movement of the components for easy positioning of the device in its neutral hanging position. In this configuration, the round rod may be rotated, causing the nut with the suspension cable attachment to move linearly as desired. Many other adjustor mechanisms are possible including a flat rectangular casing containing two ball screws oriented orthogonally to each other. One of them moves the other one slidable along the casing such that it can be positioned in many paths, all parallel to each other [i.e., both ends of it are moved equally up and down]. On this second ball screw is a threaded nut that moves along the screw as it is rotated. The suspension cable is attached to this nut. By rotating the two ball screws, the cable suspension point (on the nut) is, therefore, moved to any point in the X-Y plane inside the casing. In another mechanism, the second ball screw is substituted with a non-threaded rod which has a slidable collar with the cable attachment point. This collar can be slide manually, and a set screw locks it in place. In another embodiment, there is a box-like casing with a cut-out in the plate that covers the top. This cut-out can be a grid pattern of many parallel rectangles that are connected in the middle by an orthogonally oriented cutout. A disc is slidable under the cutout, and is too large to be pulled through the cutout. Attached to the disc is the suspension cable, which passes through the cut-out. There is a locking mechanism, such as a cam-lock, which will lock the disc in place to prevent sliding. Adjustment is made by unlocking the lock, and sliding the disc to a location desired, then re-locking it. Alternatively, the cut-out pattern could be a spiral shape, and the disc could be slid anywhere along it. Instead of a rigid arced rod, the hanger may consist of two cables that are suspended from a square or hexagonal or other non-round horizontal bar that hangs from the spring balancer wire rope. These two cables attach to the frame approximating the torso and shoulders and holding the shield and apron. Positioning of this device could be accomplished by attachment of the suspension wire rope to a horizontal rod (top cross bar) via a collar that can slide along the rod. This rod is oriented orthogonally to the square or other non-round shaped rod that is also horizontal, and gives rise to the two cables that suspend the frame of the device approximating the shoulders and torso and attached to the shield and apron. The two cables attach to the frame in an adjustable manner, allowing front to back sliding and locking of the attachment points to the frame. Thus, between this adjustment, and the adjustment available in the hanger as described above, adjustment in the X-Y plane is possible. In this embodiment, stiff sleeves or tubes may be placed around the two lower cables in order to provide columnar support of the top cross bar, to prevent it from falling on the operators head in the event of suspension failure. Instead, it would fall forward or backward, away from the head, with diminished risk of injury. The bar could be expected to weigh less than one pound, and could be covered in a soft foam wrapping. All of these implementation examples of the FIG. 5—set offer different mechanisms for attaching the device. In some cases, the device is hanging on the rigid hanger system in a manner that permits balancing of it: much like one would balance an artistic “mobile.” By moving the attachment point of the cable to the rigid hanger, one can change/alter the balance somewhat. This balance adjustment may be helpful because changes to the lead apron or shield may throw the device slightly off-balance and cause it to tilt if the attachment point of the cable is not able to move from its “factory setting” of optimal balance. There are countless balance adjustments that could be used in conjunction with the present invention, which contemplates any such possible modifications or additions in offering an optimally balanced configuration. An alternate form of suspension employs two cables that run on either side of the operator's head and that attach at approximately shoulder level to the rigid framework. Again, this is yet another example of a possible configuration that may alleviate particular concerns or that may overcome certain obstacles presented by specific circumstances/environments. Generally, the aforementioned suspension components are used to allow motion in various directions while the garment's weight is supported in a “zero gravity” manner. These mechanisms can include a spring balancer (tool balancer), hydraulic balancer, counter balance system with weights, or constant force spring. Movement in the X and Y directions (in a plane parallel to the floor) is easily facilitated. The supported device that shields the operator may be configured to allow rapid exit and re-entry of the operator so that the operator may attend to other duties outside of the workspace area of the support system. Another embodiment outlined above includes a rigid frame attached to the support cable. The part attached to the cable may be capable of complex motion relative to the lower part of the frame [at the level of the operator's head] to allow balancing of the device by placing the cable support location in many possible locations over the head. This may be accomplished in one form by an arm that can rotate around the vertical frame while remaining at a 90-degree angle, or other fixed angle, relative to the other portion of the frame to which it can move. This can be locked or secured in a fixed position when the desired location is found. The attachment site of the cable can slide along this arm. This allows for fixation at any point [within a large plane over the operator] to account for the desired balance and angulations of orientation of the device. In regards to the actual device, the garment may contain radiation-absorbing materials, such as lead or other metals. The garment can be thicker and heavier than traditional radiation protection garments because the operator does not support the weight of the garment. Additionally, the garment can cover more of operator's body such as the operator's arms and legs, or less depending on particular needs. The garment can substantially contour to the operator's body or be loose based on specific parameters. Thus, materials and/or components may be included in the garment in order to achieve the teachings of the protective, free moving, and weightlessness features of the present invention. However, due to its flexibility, the garment may alternatively be equipped with (or include) any suitable component or material, or any other suitable element or object that is operable to facilitate the operations thereof. Considerable flexibility is provided by the structure of the garment in the context of providing an adequate suspended personal radiation protection system. Face shield 12 may contain radiation-absorbing materials such that it attenuates X-rays, but is transparent to visible light. Face shield 12 can be made heavier and it can curve or bend around the operator's face to cover more of the operator's face, as compared to traditional radiation protection face shields. Face shield 12 protects the operator from radiation approaching from the sides of operator's face and its design can be biased to one side of the face. A fastening element of the garment can be positioned in the front, side, or rear of the garment. A belt design is described herein, but alternatively the garment can be opened and closed by Velcro, buckles, or any suitable fastening means for attaching pieces of a heavy material together. An assistant can fasten the Velcro or buckles such that the operator can quickly and effortlessly receive radiation protection from the garment that is substantially contoured to operator's body. The operator can wear a sterile gown and sterile gloves in the normal manner. The belt configuration on the garment (as described in greater detail below) can include Velcro, buckles, or other fastening means such that the configuration helps secure the garment to the operator's body. The belt can be fastened on the front, side, or rear of the garment. The belt also helps the garment substantially contour to an operator's body such that operator's body is properly protected. Velcro buckles or fastening means for adjusting the garment allow an operator to adjust the length of the garment. For example, a shorter person can fold up the excess garment material and fasten the garment such that the bottom part of the garment is double-layered. Similarly, a tall person can unfasten the double-layered area of the garment to receive more radiation protection on his legs such that the garment hangs to the operator's feet. Sleeve 18 can be provided on the left or right arm [or neither arm] and sleeve 18 may contain radiation-absorbing materials such as lead or other metals. Sleeve 18 allows more protection coverage of the operator's body because the operator does not support the weight of the suspended sleeve. Sleeve 18 can also provide additional coverage for the side of the body that is most exposed to the radiation. Sleeve 18 may include a shoulder plate or harness underlying/reinforcing the sleeve configuration. FIGS. 6A-6C illustrate some of the functions of an example belt 30 that includes a quick release 20, a number of flexibility points 42 and 44, and a spring 46, as further detailed by FIG. 6B. The springs are used to keep the hinges in an open position in one example embodiment. FIG. 6C highlights a belt hinge joint mechanism allowing locking or unlocking of many joints simultaneously. A set of cables 45 attach to a tensioner and a quick release mechanism. Tiny rounded ridges 47 are at each joint, whereby a view of the ridges in the joint surface is depicted. In this embodiment, there are many joints with small male ridges and corresponding female depressions on an opposite joint face. These joints may move freely when no tension is on the cable running their length. Upon tensioning the cable, the joints come into forceful contact and the locking ridges prevent hinge motion, and the joints are thereby “locked.” When tension is released, the hinges become loose again. The present invention may be used in conjunction with a corresponding array of hinge joints that provide longitudinal support but no locking mechanism. The underlying “belt” of non-locking joints will support this locking “belt,” providing a means of locking and unlocking the joined system of two belts. These two systems may be closely related such that the supporting array contains the ridged-hinge array. Other alternatives could include a single flexible point that allows the belt to open as a set of pinchers. Other variations are certainly within the broad scope of the present invention. In operation of another example implementation, the lower end of the upper frame component may form an arc around the front of the operator (preferably at, or slightly under, the level of the chin) and pass from one shoulder to the other. This will be in a position to allow a full range of motion of the head. This frame portion can serve as a support system for the lower frame components and everything else below this level, as well as the optically transparent face and head shield, which may rest on this arc or be fixed to it. The face shield may form an arc resembling the frame arc and sweep around the operator's face, possibly with a predominance to the left since that is the side most often exposed to radiation in most settings. Shielding would also be frontal, and the right side (in one example) may be incompletely shielded, although this configuration could easily be altered. Shielding material may also be draped over the shoulder, or supported over the shoulder in a manner like an umbrella in a flexible or rigid manner. Shielding may also be draped over the front of sides of the operator from this top frame component. The lower portion of the frame may be fixed to the upper portion by three or more attachment pieces (such as rods). This may be rigid, adjustable, or have some flexibility or motion in some directions to allow for different body shapes. The joints may allow expansion or motion within a defined range (possibly only in certain directions) while maintaining rigid support of the desired structures such as the lower frame and radio-protective fabric or shielding material. As depicted in FIG. 4, a sterile covering 15 of fabric, paper, or plastic may be shaped to cover the device portions, which will be located at (or below) the chest level of the operator. This can be applied to the device prior to or after engagement of the operator with the device. A possible design of the sterile cover 15 is similar to a large plastic bag, which is passed from the bottom of the shield to the shoulder level by an assistant who does not need to touch the sterile portions, or who wears sterile gloves when applying the bag. This bag will completely contain the device components and allow operation of the belt by allowing the operator to grab the necessary components underneath the sterile layer, much as is common practice currently with sterile ultrasound covers and table-side control panel covers in the catheterization suites. The sterile cover 15 may be contained within an outer bag that is sterile on the inside (contact with sterile cover) but not necessarily sterile on its outside. This permits it to touch the floor or nearby non-sterile areas as the cover is pulled up over the shield. The outer cover is then stripped off. In another embodiment, the sterile cover 15 is applied by wrapping it around the shield in a horizontal manner [much as one might wrap a towel around their body after swimming] and then securing it with tape, a clamp, or other means. The sterile cover 15 serves to maintain sterility of the operator. The operator is sterily gowned upon entry into the device. The front of the operator comes into contact with the rear of the shield, which is sterily wrapped in the covering described above. In one operational example, the arms of the operator are held out laterally upon entry, thus not touching anything. The upper part of the shoulder (near the neck) may come into contact with non-sterile components, but this is out of the sterile field and this does not disrupt the sterile field. When the operator approaches the operating table while engaged in the device, the shield that is in front of him is sterile in the same manner that the operator's front would normally be sterile if not using the device. When the operator exits the device, she steps backward and portions of her body that are considered sterile do not brush against the non-sterile portions so the operator remains sterile much as she would be conventionally without the device. Upon re-entry, the rear of the device and the front of the operator have both remained sterile so re-entry does not result in a loss of sterility relative to the initial entry. The lower portion of the device may contain a structure for supporting the shielding material below the level of the operator's neck. It may be integrated with (or separate from) another frame structure which partially or completely surrounds the operator in a manner similar to a belt [in order to cause the device to remain in close proximity to the operator while he/she moves or walks]. The shielding material may be suspended from the arc-shaped component, which resists downward bending or sagging, thus supporting the shielding material. It may be capable of motion in a plane parallel to the floor to allow wrapping of itself, as well as the shielding material around the operator (either fully or partially). Either integrated with the frame, or separate from the frame, is a belt, which functions similarly to a large clamp that may partially (or completely) surround the operator and which may be operable by the operator without supplemental assistance. Hinges or other flexible components allow the belt to wrap around the operator and to be secured in position by tightening of the hinges and/or closing of the joints with a cable mechanism. This can include a ratchet or non-ratcheted handle system that pulls the cable with a mechanical advantage so as to close the arms of the belt and secure it around the operator. [Note that many of these possible configurations are outlined in subsequent FIGURES, but a brief description is provided here.] Instead of a cable, a simple hydraulic system may be used for the system. This allows close proximity of the device to the operator for good function upon operator motion. Another mechanism would utilize a partially flexible piece, which when pushed outward or expanded, applies pressure to the belt (thus tightening it around the operator). Another embodiment could utilize a magnetic attachment in the back to allow the rear components to lock together, thus permitting closure of a full circuit around the operator, which could then be tightened by any other manner (possibly as could occur with a conventional belt). Upon sterile exit, the operator could activate a mechanical linkage to the locked magnetic components in the rear using his sterile hands in the front in such a manner so as to unlink the rear components and allow a sterile exit from the device. In another example, the operator could wear a simple fabric or plastic belt around his chest or waist, with a conventional buckle or hook and loop closure mechanism, which contains a magnet within it. The operator could apply the belt prior to donning the sterile gown or going near the device. It would wrap around her non-sterile shirt. After applying the sterile gown and entering the device, a second magnet in the device would come in close proximity, separated only by the sterile gown and sterile wrappings, with the magnet affixed to the operator. This would result in a magnetic attachment of the device to the operator, such that the device will move with the operator. This could eliminate the need for an articulated or hinged clamp-like belt that wraps around the operator to keep the device in proximity to the operator. There could be several magnets in the belt and the device to provide a good bond. Or there could be magnets in the device and ferromagnetic material such as steel in the belt, or vice versa. Disengagement could involve a sliding or peeling motion of the operator relative to the device, such as with stooping. Or the operator could grip the frame in the front to help pull it away from the body and other magnet on the operator, until the distance between them is enough to result in sufficient weakening of the attraction so as to result in simple separation. This distance should not be great since the force is inversely proportional to the square of the distance between the magnets, and the distance between them is minute when they are engaged, so powerful magnets need not be used. Alternatively, electromagnets may be employed in the device, and the belt need only contain ferromagnetic material such as steel. These magnets could be turned on and off as needed to engage/disengage the device. Another embodiment could utilize slidable components that can be slid backwards to complete a fuller arc around the operator to provide an adequate grip for good function. These would slide rearward on a frame that is otherwise incomplete, which would allow entry of the operator into the system before the slidable components wrapped around their back. In another embodiment, a flexible belt would be suspended in a manner that allowed the operator to enter the device from the rear without becoming contaminated by the belt (either because it was sterile, or because it was suspended over the head for entry, and then dropped down to the chest or abdominal level for securing in the front). Upon sterile exit, the operator would activate a component, which raised the belt high enough for the operator to exit without contamination of the sterile field. Since the belt had been wrapped around the operator, the rear component would be considered non-sterile. Upon re-entry, the operator would enter below the suspended belt and then activate the mechanism, which lowered the belt into proper position for tightening. During this process, an operator could pass his/her arms over the sterile component of the belt at the sides towards the front. In another embodiment, the belt may be passed through a sterile covering such as a sleeve, which permits sliding of the belt around the operator such that the belt remains sterile along its entire length, even though the outer portion of the sleeve is not considered sterile in the rear of the operator. The frame may permit greater freedom of motion than other configurations, which are shoulder-supported or shoulder-based with regard to main attachment sites of load below the level of the head. Most aprons cross over the shoulders to be supported, which prevents rapid entry and exit while maintaining sterility. If the apron is supported by an overhead support system that has the material draped over the shoulder, there can be some limitation of arm motion, as well as difficulties with rapid exit and re-entry while remaining in proper sterile form, as is considered standard for operative procedures. This new design overcomes those issues, as it permits rapid entry and exit while remaining properly sterile. FIGS. 7A-7D illustrate quick releases without swivels, as depicted by items 50, 52, and 54. This mechanism can be incorporated into the vertical component of the rigid hanger, which supports the device frame and other components, and is supported by the wire rope or other suspension apparatus that attaches to the top of the arced portion of the hanger. This may be used in particular in combination with a hanger configuration that includes two rigid vertical rod-like components, on both sides of the head, instead of just two uni-lateral component. FIGS. 7E-7I illustrate a number of clutch plate joints and assemblies which may be incorporated into the belt mechanism to facilitate opening and closing including desirable functions of ease of operation, economy of hand motion and maintenance of sterility of hands, and locking function of the belt in the closed configuration so as to keep the frame in proximity to the body during operations. FIG. 7E shows a clutch plate joint 53 that includes a hydraulic extender, a clutch plate, an extension spring, and a pivot pin. In this example, ridges are provided on the plate surfaces. The end-on view of a clutch plate 55 similarly shows a hydraulic extender and an extension spring. A clutch plate 57 is also provided and it offers an angled and a straight configuration, where there are ridges and depressions present. The clutch plate mechanism consists of a joint allowing rotation around a pivot pin. The joint surfaces have rounded male ridges in a radial pattern with corresponding female depressions on the opposite joint surface. When the two clutch plate surfaces are pressed together, the ridges and depressions, in addition to normal friction, result in locking of the two surfaces, thus, locking the joint in its configuration or angle. This is accomplished by the hydraulic extender positioned over one of the surfaces, pushing against it and the outer casing. The casing is fixed, so the clutch plate is pushed against the opposing clutch plate, causing them to lock. When quick release of the lock is desired, the hydraulic pressure is released and the clutch plates may move again, allowing rotation and angulations of the joint. This is facilitated by extension springs on the opposite side, which push against the casing, pushing the plates apart and allowing free rotation and angulations of the joints. The casing contains the clutch joints, provides longitudinal rigidity to the entire apparatus, and has joints sharing the same pivot pins with the clutch joints (permitting free angulation with the clutch plate joints). FIG. 7F illustrates a pin lock mechanism 59 for the belt of the garment. FIG. 7G illustrates a ratchet mechanism 61 for the belt of the garment. FIG. 7H illustrates a friction cam mechanism 63 for the belt hinge locking and quick-releasing component. FIG. 7I illustrates a ratchet mechanism 65 to be used with the garment. For purposes of teaching, each of these FIGURES include some directions or guidance for how each component functions. Other ratchet mechanisms of different designs, as in common ratchet wrenches, may also be incorporated. FIGS. 8A-8C illustrate an example design 58 of the garment being described herein. A fastening or securing element 80 is provided with the garment and the frame is provided with release 20 and some flexible belt members 42 and 44. FIG. 8B illustrates another example design 60 and FIG. 8C illustrates yet another example design 62. FIG. 9 illustrates an example of a frame 68 to be used in conjunction with the garment. FIGS. 10A-10E illustrate a number of quick release mechanism examples with accompanying swivels in some cases. These embodiments depict an example configuration 81 that includes a ball component 74 with two casings 72 and 78. FIG. 10E depicts a separate side view 76 that offers another perspective of casing 72. To better inform the audience, some discussion of the techniques to be used with some of these components is offered. In operation of a typical scenario, an operator or assistant may pre-adjust settings on the garment to make it fit to his/her body shape optimally. This may involve setting the adjustable rigid hinge in the rear, or the connections between the upper and lower frame portions. The balance of the device may also be adjusted by adjusting the cable attachment position on the top of the hanger using the hanger adjustor mechanism, which may then be locked into a preferred position. An operator or assistant may apply a sterile covering to the device. This may cover device components extending from the bottom of the shield to the level of approximately the mid chest or higher. An operator may perform a surgical scrub of their hands and arms in the usual manner, and then don the surgical sterile gown. The operator may then hold his/her arms up partially and step into the device from the rear and apply the front of his/her body to the rear of the device: lowering the arms in a manner that causes the jointed belt component to close around the operator's body. In some embodiments, especially in those employing a cable mechanism to retract the belt components, the operator may then activate the tensioner component, which provides sufficient tension to enable a secure closure of the jointed belt around the operator. In other embodiments, such as those employing a ratchet or friction-lock mechanism in the belt, a tensioner is not employed. Tension, or locking, is automatically maintained until the operator chooses to release it for exit. An operator may then perform the normal procedures for treatment of a patient or an animal, or perform other functions that can result in radiation exposure to operators. When an operator desires to become dissociated from the device, he/she may activate the release mechanism, releasing tension on the belt and allowing it to open sufficiently for the purpose of operator exit. For an exit, the operator lifts their arms partially and steps backwards, dissociating from the device. In one embodiment, the operator will attach a small hook or other attachment component located on the device to a stationary component in the work area, prior to operator dissociation form the device. This may be within the sterile covering and attached to the sterile area (such as the instrument table or the operating table), or it may be outside of the sterile portion of the device (above the shoulder level), and attached to a non-sterile object. This attachment will prevent unwanted motion of the device along the support system, such as crane rails or a trolley, while unattended by the operator, and this will help insure the sterility of sterile components. In other embodiments, a magnet may be used for this purpose instead of a hook. Re-entry is possible in same manner as the initial entry described above. It is noteworthy and important that the operator can fully operate this device without outside or additional assistance, and while preparing for surgery, and without moving his/her hands anywhere outside the sterile field (such as towards their back, above their upper chest, or below their waist). Note, as outlined above, there are many configurations disclosed for opening and closing the belt around the operator, locking it in proper position, providing quick release mechanisms for opening the belt, and allowing operator exit. The main goals of these mechanisms are: rapid, easy closure of the belt around the operator, which is usually accomplished by manually squeezing the belt itself with the operator's arms (or alternatively by activating a linkage mechanism utilizing cables, hydraulics, or rigid rod) and closing it at its hinges and mobile joints. Similarly, the device offers a rapid, easy opening of the belt to allow an exit of the operator such that the operator only touches components on (or near) the front of the device without losing sterility. Any required motions of the hands toward the rear area would result in a loss of sterility of the hands. The release mechanisms afford simple gross hand and finger movements so that they can be accomplished through the sterile cover without fumbling. One important objective in the mechanisms outlined herein is to simply reduce the number of steps to accomplish the sterile entry/exit operations identified above. Yet another objective is automation for the opening of the belt such that the operator only activates a quick release and the other motions are set in motion. This is most simply accomplished by a spring mechanism (wire or gas springs) or a similar tensioning mechanism: exerting forces on the belt or connected apparatus, which results in opening the device. In other embodiments, automation is not incorporated and the belt is opened manually by the operator who either pulls the belt open with her hands, or activates a linkage mechanism that opens the belt. There are other mechanisms that do not necessarily satisfy all points outlined above, but are still feasible. They may accomplish all necessary actions, but not necessarily with the same expedience, economy of operator involvement, or automation. These mechanisms are also depicted in the FIGURES and discussed herein because they offer viable alternatives and promote some advantages in terms of simplicity or economy of design. In other scenarios, these alternatives are less bulky and/or provide better durability characteristics for the accompanying mechanisms. FIGS. 11A-11C illustrate grip locks shown on the belt of the garment. For an example design 86 of FIG. 11A, all of the above points identified above are satisfied, where there are no cables. Instead, example design 86 uses metallic rods attached to the belt via articulating (hinge) joints (in some ways similar to the apparatus that is sometimes present on bus doors). Opening the belt to exit the garment requires activation of the quick release only and the closing/locking operation requires manual closing of the belt only. The grip lock mechanism may be in a separate “box” as shown, or integrated into the belt frame as depicted. A front view example design 88 is also provided. In this instance, the belt is pulled by springs (e.g., flexion springs) at the hinge joints. This is allowed by a quick release, which allows rods to move freely outwards. This allows the hinge to open fully. To close the belt, the operator swings the sides of the belt inwards, pulling the rods outwards. The grip plates lock the rods in place: maintaining the belt in a closed position. Different embodiments allow the spring to attach to the metal rod. This provides more force from the same spring and results in a greater spring excursion. A gripper cage can be swivel-mounted to accommodate the change of angle of the metal rod with motion. The grip lock mechanism may be attached to the front or the back of the belt at the attachment site beyond the belt hinge joint. This determines if pulling the rods results in opening or closing of the belt. In the depicted version, the springs in front of the belt hinge joints provide the force to open the belt automatically upon release of the grip locks. This spring force could be attached in other locations in other embodiments. For example, it could be attached to the free ends of each metal rod, pulling it outward, or in other ways such as a hinge spring, as seen in FIG. 6B. In the grip lock example 92 of FIG. 11B, the belt is pulled open by the springs at the hinge joints. This is allowed by a quick release mechanism, which allows rods to move freely inwards. This allows the hinge to open fully. To close the belt, the operator swings the sides of the belt inwards: pulling the rods outwards. The grip plates lock the rods in place: maintaining the belt in the closed position. In another example design 96 of FIG. 11C, long metal rods pass through a casing and through slots in gripping plates secured to the top of the casing. The slot in the gripping plates is just large enough to accommodate the rods when passing perpendicularly through the slot. In this configuration, the rod can pass freely through the plate and the casing. In the neutral position, the extension spring between the gripping plates pushes them apart, resulting in an oblique orientation relative to the casing and the rods. This, combined with the inward tension on the rods due to the springs in the belt (not shown), results in friction between the gripping plates and the rods, thereby locking the rod in position. In one example, this is the “locked” configuration and is the configuration during operation. As with all the gripping plate mechanisms in the FIG. 11 series, the rod may freely slide in one direction even when the quick release mechanism is not activated. This occurs because in the neutral position, the gripping plates are pushed into an oblique orientation by the extension spring. Force applied to the rod in one direction will cause further obliquity of the gripping plate, resulting in friction lock, whereas force in the opposite direction will move the gripping plate slightly into a less oblique orientation, resulting in effective enlargement of the slot length and permitting movement of the rod through the gripping plate. This free motion in one direction is used to allow the operator to easily close the device around himself when desired. Closing the belt moves the rods in the direction that permits free motion without activation of the quick release mechanism. In some embodiments, the gripping mechanism may be configured to grip in both directions, requiring activation of the release mechanism to permit both opening and closing of the belt. When the operator wishes to open the belt to exit the device, the gripping plates are squeezed together by the fingers, resulting in a quick release of the rods. The spring tension in the belt (not shown) will push the rods apart, allowing the belt to open. The advantage of this embodiment over the one using cables is that the rods are automatically positioned in the neutral position (belt open) because the rods are pushed. This eliminates the need for the operator to slide the rods manually, as in the cable version. In another embodiment, a rack and pinion arrangement may be incorporated into the linkage mechanism instead of friction gripper. The rack links with the belt in an articulated manner similar to described embodiments, but is controlled by a pinion which is controlled by the operator. The pinion may include a ratchet mechanism that allows its rotation in one direction only, until quick release is activated when freedom of motion in both directions is then allowed. Alternatively, the pinion must be released to rotate in either direction. In another embodiment, a toothed rod is employed instead of the smooth rods depicted in the FIG. 11 series. Instead of a friction gripper plate, there can be a catch mechanism or pawl that sets in between the teeth in the resting or neutral position due to a spring or gravity. The teeth may be angled to as to allow easy direction of the rod in one direction, as the pawl slides over teeth like a ratchet mechanism, whereas it locks with motion in the other direction. A quick release mechanism may allow free motion in either direction. Alternatively, the teeth may be angled such that the device may be locked to motion in both directions until released. As can now be fully appreciated, such a radiation protection garment offers obvious advantages to operators who work with radiation. This is due, at least in part, to the suspended nature of the garment and shield, which together protect the operator from harmful radiation. System 10 allows an operator to have a great degree of freedom of motion commonly used during medical and research procedures. Furthermore, an operator can remain sterile while using the garment due to its intelligent design and quick release abilities. It is important to note that the stages and steps described above illustrate only some of the possible operations that may be executed by, or within, the present system. Some of these stages and/or steps may be deleted or removed where appropriate, or these stages and/or steps may be modified, enhanced, or changed considerably without departing from the scope of the present invention. In addition, a number of these operations have been described as being executed concurrently with, or in parallel to, one or more additional operations. However, the timing of these operations may be altered. The preceding example flows have been offered for purposes of teaching and discussion. Substantial flexibility is provided by the tendered system in that any suitable arrangements, chronologies, configurations, and timing mechanisms may be provided without departing from the broad scope of the present invention. Accordingly, any appropriate structure, component, or device may be included within suspended personal radiation protection system 10 to effectuate the tasks and operations of the elements and activities associated with providing optimal radiation protection. Although the present invention has been described in detail with reference to particular embodiments, it should be understood that various other changes, substitutions, and alterations may be made hereto without departing from the spirit and scope of the present invention. The illustrated device and operations have only been offered for purposes of example and teaching. Suitable alternatives and substitutions are envisioned and contemplated by the present invention: such alternatives and substitutions being clearly within the broad scope of the proposed solutions. Using analogous reasoning, suitable devices that are conducive to properly supporting the weight of the operator, the garment, and the face shield could readily be used or adopted by system 10. In addition, while the foregoing discussion has focused on medical procedures, any other suitable environment requiring radiation protection may benefit from the compatibility teachings provided herein. Similarly, the term ‘operator’ should be reasonably construed to not only include a living organism but inanimate objects (e.g., tools or robotics) where radiation exposure presents a problem as well. Although the present invention has been described with several embodiments, a myriad of changes, variations, alterations, transformations, and modifications may be suggested to one skilled in the art, and it is intended that the present invention encompass such changes, variations, alterations, transformations, and modifications as fall within the scope of the appended claims. |
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claims | 1. An apparatus suited to transport radiopharmaceuticals, comprising:a radiopharmaceutical pig with two halves each defining a respective cavity and being secured together so that the respective cavities together form a closed chamber, each of the two halves including a respective lead shield each with a surface; andencapsulating material arranged to encapsulate and seal the surface of at least one of the two lead shields to define a portion of the closed chamber by bounding same, the closed chamber being configured to accommodate a radiopharmaceutical syringe; andat least one housing within the closed chamber, the housing being formed of a non-puncture resistant material that is configured to give way to puncturing in response to driving of a tip of an attached needle of the radiopharmaceutical syringe under manual forces into the housing, the housing being elongated and configured to be open at one end and closed at an opposite end to define a space to accommodate insertion of the attached needle of the radiopharmaceutical syringe through the one end that is open so that a tip of the needle remains clear of the opposite end that is closed when flanges of the radiopharmaceutical syringe are held in position between the lead shields. 2. An apparatus of claim 1, wherein the closed chamber is configured to accommodate the radiopharmaceutical syringe with an attached needle, further comprising a radiopharmaceutical syringe with the attached needle being within confines of the closed chamber. 3. An apparatus of claim 1, further comprising at least one housing within the closed chamber, the housing being configured to accommodate insertion of a lower portion of the radiopharmaceutical syringe with attached needle through a mouth of the housing and into confines of the housing. 4. An apparatus of claim 1, wherein the encapsulating material encapsulates with securing material selected from a group consisting of ultrasonic weld material, heat seal material, adhering material, and laminating material. 5. An apparatus of claim 1, wherein the encapsulating material is secured in position to encapsulate the lead shields with mechanical fit components selected from a group consisting of screw locks, clamps, snap rings, pressure fit components and mechanical fasteners. 6. An apparatus of claim 1, wherein the radiopharmaceutical pig includes respective casings, each of the casings being secured to an outer facing surface of a respective one of the lead shields. 7. An apparatus of claim 6, wherein at least one of the casings and the encapsulating material are secured to each other with securing material selected from a group consisting of ultrasonic weld material, heat seal material, adhering material, and laminating material. 8. An apparatus of claim 6, wherein at least one of the casings and the encapsulating material are secured to each other with at least one mechanical fit component selected from a group consisting of screw locks, clamps, snap rings, pressure fit components and mechanical fasteners. 9. An apparatus of claim 1, wherein one of the two lead shields has an outwardly extending flange with a further surface that is clear of the closed chamber, the encapsulating material being arranged to encapsulate the further surface. 10. An apparatus of claim 1, wherein the two lead shields are elongated and each have edges that face each other, the edges being configured to overlap and engage each other to prevent radiation leakage. 11. An apparatus of claim 1, wherein the at least one lead shield has an outer facing surface that is likewise encapsulated by the encapsulating material and contiguous with the encapsulating material that encapsulates the surface that defines and bounds the portion of the closed chamber. 12. A method of assembly of an apparatus suited to transport pharmaceuticals, comprising:encapsulating and sealing a surface of at least one lead shield of two halves of a radiopharmaceutical pig with encapsulating material, each of the halves defining a respective cavity; andbringing together the two halves of the radiopharmaceutical pig so that the respective cavities together form a closed chamber, the encapsulating material defining a portion of the closed chamber by bounding same, the closed chamber being configured to accommodate a radiopharmaceutical syringe; andarranging a housing within one of the two halves of the radiopharmaceutical pig and then carrying out the bringing of the two halves together so that the housing is within the closed chamber, the housing being formed of a non-puncture resistant material that is configured to give way to puncturing in response to a tip of an attached needle of the radiopharmaceutical syringe under manual forces being pressed against the housing, the housing being elongated and configured to be open at one end and closed at an opposite end to define a space to accommodate insertion of at least the attached needle of the radiopharmaceutical syringe through the one end that is open so that a tip of the needle remains clear of the opposite end that is closed when flanges of the radiopharmaceutical syringe are held in position between the lead shields. 13. A method of claim 12, wherein the closed chamber is configured to accommodate the radiopharmaceutical syringe with an attached needle, further comprising inserting the radiopharmaceutical syringe with attached needle into the cavity of one of the halves of the radiopharmaceutical pig before bringing the two halves together to form the closed chamber, the radiopharmaceutical syringe with attached needle being within the closed chamber after the two halves of the radiopharmaceutical pig are brought together. 14. A method of claim 12, comprising:encapsulating an outer facing surface of the at least one lead shield with a casing; andsecurely attaching the encapsulating material and the casing to each other with a material selected from a group consisting of an ultrasonic weld seal, a heat seal, an adhering material, and a laminating material. 15. A method of claim 12, further comprising encapsulating an outer facing surface of the at least one lead shield with a casing; andsecurely attaching the encapsulating material and the respective casings to each other with at least one mechanical fit component selected from a group consisting of screw locks, clamps, snap rings, pressure fit components and mechanical fasteners. 16. A method of claim 12, wherein one of the lead shields has an outwardly extending flange with a further surface that is clear of the closed chamber, further comprising encapsulating the further surface with the encapsulating material. 17. A method of claim 12, further comprising carrying out the step of bringing the two halves of the radiopharmaceutical pig together while a housing is present in one of the two halves so that the housing is within the closed chamber after the two halves are brought together. 18. A method of claim 17, further comprising inserting the radiopharmaceutical syringe with an attached needle within confines of the housing prior to bringing the two halves of the radiopharmaceutical pig together. 19. An method of claim 12, further comprising encapsulating an outer facing surface of the at least one lead shield by the encapsulating material to be contiguous with the encapsulating material that encapsulates the surface that defines and bounds the portion of the closed chamber. |
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claims | 1. A neutron moderator comprising a fluoride sintered body, wherein the fluoride sintered body has a thickness of 60 mm or more and 64 mm or less and comprises MgF2 of a compact polycrystalline structure having a bulk density of 2.90 g/cm3 or more and 3.07 g/cm3 or less and a bending strength of 10 MPa or more and 25 MPa or less as regards mechanical strengths. 2. The neutron moderator according to claim 1, wherein the sintered body has a Vickers hardness of 71 or more and 120 or less as regards mechanical strengths. 3. The neutron moderator according to claim 1, wherein the sintered body has volume of 2,183 cm3 or more and 2,421 cm3 or less. 4. A neutron moderator comprising a fluoride sintered body, wherein the fluoride sintered body has a thickness of 60 mm or more and 64 mm or less and comprises MgF2 of a compact polycrystalline structure having a bulk density of 2.90 g/cm3 or more and 3.07 g/cm3 or less and a bending strength of 10 MPa or more and 25 MPa or less as regards mechanical strengths,wherein the sintered body is produced by a process comprising the steps of:pulverizing a high-purity MgF2 raw material and mixing it with 0.03-0.5% by weight of a sintering aid:molding the resulting mixture obtained after the mixing step at a molding pressure of 5 MPa or more using a uniaxial press device and subsequentlyat a molding pressure of 5 MPa or more using a cold isostatic pressing (CIP) device to obtain a molded article;conducting preliminary sintering of the molded article in the temperature range of 550° C.-600° C. for 4-10 hours in an air atmosphere;heating in the temperature range of 750° C.-840° C. for 5-12 hours in an inert gas atmosphere; andconducting main sintering by heating in the temperature range of 900° C.-1100° C. for 0.5-3 hours in the same atmosphere as the preceding step so as to form MgF2 sintered body having a compact structure. 5. The neutron moderator according to claim 4, wherein the sintered body has a Vickers hardness of 71 or more and 120 or less as regards mechanical strengths. 6. The neutron moderator according to claim 4, wherein the sintering aid is selected from the group consisting of carboxymethyl cellulose and calcium stearate. 7. The neutron moderator according to claim 4, wherein the sintered body has volume of 2,183 cm3 and 2,421 cm3 or less. |
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summary | ||
description | This application is a continuation of U.S. Ser. No. 11/154,085 filed Jun. 16, 2005, which application is a continuation of U.S. Ser. No. 10/619,702 filed Jul. 15, 2003, which claims priority to U.S. provisional patent application Ser. No. 60/396,322 filed Jul. 17, 2002, the disclosures of which are incorporated herein by reference. The disclosed methods and apparatus relate generally to the construction and use of magnetic focusing and correction elements for modifying the intensity distribution of ions within ribbon beams and more particularly to precision correction of the angle of incidence of ions used for implanting and doping in semiconductor devices. The process of ion implantation is useful in semiconductor manufacturing as it makes possible the modification of the electrical properties of well-defined regions of a silicon wafer by introducing selected impurity atoms, one by one, at a velocity such that they penetrate the surface layers and come to rest at a specified depth below the surface. It makes possible the creation of three-dimensional electrical circuits and switches with great precision and reproducibility. The characteristics that make implantation such a useful processing procedure are threefold: First, the concentration of introduced dopant atoms can be accurately measured by straight-forward determination of the incoming electrical charge that has been delivered by charged ions striking the wafer. Secondly, the regions where the above dopant atoms are inserted can be precisely defined by photo resist masks that make possible precision dopant patterning at ambient temperatures. Finally, the depth at which the dopant atoms come to rest can be adjusted by varying the ion energy, making possible the fabrication of layered structures. Systems and methods are desired for enhancing the ion implantation process. The ion species presently used for silicon implantation include arsenic, phosphorus, germanium, boron and hydrogen having energies that range from below 1 keV to above 80 keV. Ion currents ranging from microamperes to multi-milliamperes are needed. Tools providing implant currents greater than about0.5 mA are commonly referred to as ‘high-current’ implanters. Trends within the semiconductor industry are moving towards implantation energies below 1 keV and control of angle of incidence below 1.degree. Typically, an ion implanter for introducing such dopant materials into silicon wafers and other work pieces may be modeled into four major systems: First, an ion source where the charged ions to be implanted are produced. Secondly, an acceleration region where the energy of the ions is increased to that needed for a specified implant procedure. Thirdly, an optical ion transport system where the ion ensemble leaving the source is shaped to produce the desired implant density pattern and where unwanted particles are eliminated. Finally, an implant station where individual wafers are mounted on the surface of an electrostatic chuck or a rotating disc that is scanned through the incoming ion beam and where a robot loads and unloads wafers. One aspect of the present invention aims towards enhancing or improving ion beam transport systems. A recent improvement for ion implanter design has been the introduction of ribbon beam technology. Here, ions arriving at a work piece are organized into a stripe that coats the work piece uniformly as it is passed under the ion beam. The cost advantages of using such ribbon beam technology are significant: For disc-type implanters, ribbon-beam technology eliminates the necessity for scanning motion of the disc across the ion beam. For single-wafer implanters the wafer need only be moved under the incoming ribbon beam along a single dimension, greatly simplifying the mechanical design of end-stations and eliminating the need for transverse electromagnetic scanning. Using a correctly shaped ribbon beam, uniform dosing density is possible across a work piece with a single one-dimensional pass. The technical challenges of generating and handling ribbon beams are non trivial because the ribbon beam/end station arrangement must produce dose uniformities better than 1%, angular accuracies better than 1 degree and operate with ion energies below 1 keV. U.S. Pat. No. 5,350,926 entitled “High current ribbon beam ion implanter” and U.S. Pat. No. 5,834,786, entitled “Compact high current broad beam ion implanter”, both issued to White et al., present some features of ribbon beam technology. White et al. have also reviewed some of the problems of generating ribbon beams in an article entitled “The Control of Uniformity in Parallel Ribbon Ion Beams up to 24 Inches in Size” presented on page 830 of the 1999 Conference Proceedings of Applications of Accelerators in Research and Industry”, edited by J. L. Dugan and L Morgan and published by the American Institute of Physics (1-56396-825-August 1999). By its very nature, a ribbon beam has a large width/height aspect ratio. Thus, to efficiently encompass such a beam traveling along the Z-axis, a focusing lens for such a beam must have a slot-like characteristic with its slot extending along the X-axis and its short dimension across the height of the ribbon (the Y-direction). The importance of this is that, while the focal lengths of a magnetic quadrupole lens in each dimension are equal but of opposite sign, the angular deflections of the ribbon's boundary rays in the width and height dimensions can be very different. In addition, the magnetic field boundaries of the lens can be close to the ion beam permitting local perturbations introduced along these boundaries to have deflection consequences that are effectively limited to a small region of the ribbon beam. While in principle it is feasible to generate a wanted shape of ribbon beam directly from an ion source, in a practical situation full-length ribbon extraction may not be feasible. Often it is desirable to generate a modest-length ribbon at the source and expand it to the width required for implantation, using ion-optical expansion. Another aspect of the present invention is directed towards extracting ions from an ion source in the form of a multiplicity of individual beamlets whose central trajectories are parallel and arranged in a linear manner. Such geometry provides a precise definition of the origin and angular properties for each beamlet. Those skilled in the art will recognize that these principles remain valid even if multiple parallel rows of beamlets are used or if the central trajectories of the beamlets are not parallel when they leave the source region or if a slit-geometry is chosen for ion extraction. Furthermore, those skilled in the art will recognize that focusing and deflection elements will be needed to transport the ions between an ion source and a work piece where the particles are to be implanted. For focusing lenses to operate as ideal focusing elements it is desirable that, to first order, the angular deflection introduced to the trajectory of individual beamlets be proportional to the beamlets distance from the lens symmetry axis; namely, the magnitude of the deflecting fields should increase linearly with distance from the central trajectory of the ion beam. Quadrupole lenses satisfying the linearity requirement described above and having high length to height aspect ratio have been described by W. K. Panofsky et al. in the journal Review of Scientific Instruments volume 30, 927, (1959), for instance. Basically, their design consists of a rectangular high permeability steel frame with each of the long sides of the frame supporting a single uniformly wound coil. To generate a quadrupole field the top and bottom coils are wound equally spaced along each of the long sides of the steel frame members with the currents through the coils being excited in opposite directions when viewed from one end of the rectangular array. A north pole at the end of one bar sees a south pole facing it. On the short sides of the rectangular frame, additional coils are used to buck the magnetostatic potential at both ends of each long side preventing magnetic short circuits through the end-bars. For quadrupole field generation the opposing ampere-turns along each vertical bar are equal to the ampere-turns along each of the long bars. The currents passing through these two bucking coils will be equal but generate fields in opposing directions. For many focusing applications the correction of aberrations and the compensation of non-linear spreading of a low energy beam is critical so that the possibility for producing deviations from a linear growth of magnetic field away from the center is desirable. A method for introducing the necessary multipole components to the field has been described by Enge '328 in U.S. Pat. No. 3,541,328, particularly, the method described in this document for producing multipole focusing fields in the space between two iron cores between which ions are passed. A series of independently excitable windings, each having a coil distribution appropriate for generating a specific multipole, are wound along each of the iron cores. In the journal Nuclear Instruments and Methods, volume 136, 1976, p 213-224 H. J. Scheerer describes the focusing characteristics of such a dual rod design in accordance with the description in U.S. Pat. No. 3,541,328. Specifically, in FIG. 6 of this patent it can be seen the coils for each multipole are connected in series and powered as a single unit. The Panofsky quadrupoles and Enge multipole generators were both conceived for transmitting ions through a beam transport system where the parameters of the ion transport elements are fixed for a single experiment or measurement. They suffer disadvantages when active control of the deflecting fields is needed to correct beam parameters. First, neither design generates a dipole field contribution where the B-field is along the long axis of the rectangle. Secondly, the symmetry point (x=0) is usually established from the geometry of the coils and of the steel yokes so there is no easy way to introduce steering about the y-axis by moving the center of the lens-field distribution. In an embodiment of the present invention, a rectangular steel window frame construction provides the magnetic supporting structure needed for producing the wanted deflection fields. A feature of the present embodiment is that the windings along the long-axis bars consist of a large number of independently excited short sections. This concept allows high-order multipoles to be generated without dedicated windings and the central point of any multipole contribution can be translated along the transverse x-axis. Additional coils around the end bars are essential for eliminating magnetic short circuits when multipole components are being generated. However, these end-bar coils can also be excited independently in a manner that allows the production of a pure dipole field between the long-axis bars at right angles to the long dimension of the rectangle. Finally, when the coils on the end bars are switched off, dipole fields can be generated along the long axis of the window frame. In another embodiment of the present invention, local variations in ion density or the shape of the ribbon beam at the exit from the source are corrected by locally modifying the deflecting fields. These corrections can be made under computer control and on a time scale that is only limited by the decay rate of eddy currents in the steel. The input beam parameters needed for control involves position-sensitive faraday cups for measuring the intensity and angle distribution of ions within said ribbon beam allowing discrepancies from the wanted distribution to be corrected by modifying the deflection fields. While each of the applications of such lens variations will be discussed further, it should be appreciated that, because of linear superposition of fields in free space, the currents necessary to produce a particular type of correction can be calculated individually. This process can be repeated for each type of correction needed with the complete solution being produced by superposition. Such concurrent introduction of a selected group of multipole fields into a single beam transport element has been described by White et al. in the journal Nuclear Instruments and Methods volume A 258, (1987) pp. 437-442 entitled “The design of magnets with non-dipole field components”. The fundamental concept underlying the present invention is the creation of a region filled with magnetic fields that encompasses all trajectories comprising a ribbon beam. The d.c. magnetic fields having a magnitude and direction throughout the region that is appropriate to introduce the wanted deflections of all beamlets constituting the ribbon beam. Within the constraints implied by Maxwell's equations, magnetic field configurations can be chosen that provide controlled changes in the angular coordinates of beamlets and produce superposed corrections for: (1) angular errors, (2) differential intensity errors, (3) uniform steering about axes normal to both (y.sub.0, z.sub.0) and (x.sub.0, z.sub.0) planes, (4) the introduction of linear positive and negative focusing, (5) specialized deflection fields for aberration correction. Other objects and advantages will become apparent herinafter in view of the specifications and drawings. The unique properties of the system according to the present invention will be better elucidated by reference to a practical example. In this example, a pair of quadrupole lenses are used to expand an initially parallel set of beamlets to a broader set of parallel beamlet trajectories. FIG. 1 illustrates the beam coordinate system used in the following discussions. Three representative sections, 120, across a ribbon beam are shown. The X-axis is always aligned with the surfaces, 120, at right angles to the beamlets, 130, comprising the ribbon beam and along the surface's long axis. The Z-axis, 110, is tangential to the central trajectory, of the ribbon beam and remains coincident with the central trajectory throughout the length of the ion optical transport system, causing it to change direction as the central trajectory, 110, changes direction. At each point along the beam path the Cartesian Y-axis lies also in the surface, 120, and along the ribbon beam's cross-sectional narrow dimension. FIG. 2 shows the essential structure of an ion beam expander, 200, that optically couples an ion source, 201, having narrow width, to produce a ribbon height at a work piece or wafer, 220, that allows simultaneous ribbon beam implantation across the whole wafer width in a single traverse of the wafer 220, using linear reciprocating motion, 221. A short ribbon beam generated by the ion source 201, in the form a group of beamlets arranged in a linear array, 210, is expanded so that its width at a converging lens, 250, matches that needed at a work piece, 220, being implanted. The beam expander, 200, further comprises a diverging lens, 230, followed by a free-space drift region, 240, where the individual ion beamlets drift apart before they are collimated back to parallelism by the larger width converging lens, 250. In the preferred embodiment the work piece, 220, passes under an expanded ribbon beam pattern, 260, at constant velocity with the angle of incidence being adjustable by rotating the wafer about an axis, 270, to modify the ion impact angle, .theta. When the wafer is rotated about the axis, 270, to large angles, the beam width can be adjusted by modifying the expansion ratio to minimize beam wastage. For the geometry of FIG. 2 the ion density should be constant across the width of the ribbon beam. However, for geometries such as those of a rotating disc type implanter, the ion density within the ribbon beam must vary with implant radius. In this case, it will be clear that to produce doping uniformity at the work piece the ribbon beam ion density will generally require active correction across the ribbon beam. FIG. 3 shows the basic features of lens correctors according to the embodiment of the present invention. A high-permeability rectangular steel structure, 310, aligned with its long axis parallel to the width of a ribbon beam, 320, (X-coordinate) and with its geometric center coincident with the geometric center of the ribbon beam, supports coils, 330, 340, that are used to generate the wanted magnetic fields within a gap, 312, through which the ions forming the ribbon beam, 320, are directed. Individual coils, 330, 340, shown schematically, are distributed along both long-axis bars, 314, 316, of the rectangular steel structure, 310, with individual controllable power supplies establishing the current through each of the coils via the circuits, 350 and 351. While, for clarity, the individual coils, listed as 330 and 340, are shown with considerable separation, in practice the coils should be as close together as is practical to allow the magnetic field on the axis of beam region, 322, to vary smoothly. For some applications where the coils, 330, and 340, must have large cross section to minimize power dissipation, thin ferromagnetic plates (not shown) can be used to separate individual coils and relay the scalar potentials nearer to the ion beam boundaries. Alternatively, the coils 330 and 340 may be connected together as a continuous coil. End coils, 332 and 342, shown in FIG. 3, are not necessarily divided into multiple elements. Their primary function is to establish appropriate magnetostatic potentials that prevent magnetic short circuits between the upper and lower steel bars, 314, and 316. During quadrupole operation equal and opposite ampere-turns must be generated by coils, 332 and 342, to the ampere turns applied along the long axes of the rectangular structure. To make possible the production of several deflection modes the current directed through the end coils, 332 and 342, should be reversible and adjustable with precision. During the generation of dipole magnetic fields along the X-axis, coils 332 and 342, may be turned off. FIG. 4, illustrates a cross-section as viewed along the line A-A′, in the x-direction, shown in FIG. 3 with the addition of a surrounding vacuum enclosure. It can be seen that small high permeability steel tabs, 420 and 422, mentioned earlier, transfer the magnetostatic potential generated along each bar, 314 and 316, to the boundaries of the ion beam region, 322. The straight section of the steel tabs, 420 and 422, should be located as close as possible to the ion beam to localize the position resolution of correcting field components. Without reservations, the projections shown in FIG. 5 show the preferred embodiment of a lens-corrector enclosure. The design goal for the enclosure is to avoid exposure of the vacuum environment to the coils and their insulation. Also, to avoid vacuum to air feed-throughs for power feed and water-cooling channels. Basically, a magnetic lens/corrector can operate at ambient atmospheric pressure inside such an enclosure, 510. It has vacuum on the outside, 500, and ambient atmospheric pressure or liquid cooling on the inside, 510. The enclosure must have a depth along the Z-axis adequate to contain a coil structure as described in FIGS. 3 and 4 and sufficient magnetic path length along the ion beam that the ions can be deflected through the wanted correction angle. While those skilled in the art will recognize that there are many methods of fabricating the enclosure, 510, in the present embodiment the enclosure is machined from a suitable block of aluminum jig-plate. During operation the enclosure, 510, is bolted to a housing that is part of an implantation system's vacuum envelope, 530. Such a construction serves to define the position of the corrector element with respect to other optical elements that are part of the beam transport components used in an implanter. The corrector lens shown in FIG. 3 or 4 may be connected to the ambient atmosphere via connecting holes, 540 and 542. Through these holes, 540 and 542, pass electric power leads for each of the coils plus air or liquid cooling for the coils. The enclosure, 510, is made vacuum tight by attaching a simple plate, 550, to the flat surface, 560, sealed with O-rings, 552. The cross section view of FIG. 6 illustrates an assembled structure of a typical lens-corrector, 600, where like elements are described in previous embodiments. The rectangular high permeability bar structure, 314 and 316, is the basis of the rectangular window frame. It will be seen that for ease of wiring and cooling the steel bars may be fabricated from appropriate steel tubing that will allow easy access for the wiring and cooling lines. The Z-axis of the ribbon beam plane passes through the open center, 322, of the corrector. Power and cooling are introduced through the penetrations, 542. The electrical connections are arranged using the distribution panel, 610. FIG. 7 illustrates the background to the generation of a quadrupole field in the region between rectangular bars, 314 and 316, and how such a distribution can be modified to correct for aberrations. Assuming that a uniform current sheet, j.sub.z(x), 701, 702, is produced as illustrated by the modules around the surface of each bar, these current sheets will generates a magnetic field, B.sub.x(x), in the immediate surface of the winding given byB.sub.x(x)=.mu..sub.0.multidot.j.sub.z(x) (1) To generate a pure quadrupole field, j.sub.z(x) is constant for all values of x. Applying Ampere's theoremB.sub.y(x)=(.mu..sub.0/d).multidot.j(x).multidot.x (2) Where d is the distance from each bar to the center line, 710. Thus, for uniform currents flowing in the manner shown by the arrows in FIG. 7 north pole generated at the end of one bar sees a south pole immediately opposite on the adjacent steel bar with the magnetic field B.sub.y(x) being zero at the center of the x-dimension, measured between the vertical steel connecting bars, 721, 722, and increasing linearly from the center to each end changing sign at the center Those skilled in the art will recognize, because of superposition, that within the resolution limit of the geometry and assuming no saturation of the steel, whatever multipole is required can be excited by choosing the appropriate distribution of the current density, j(x). Clearly, individual windings having constant current and variable pitch can provide the needed variations in j(x) as has been disclosed in U.S. Pat. No. 3,541,328. However, it is realized that whatever multipole is needed can be excited by using a single group of windings provided the single winding layer is divided into a large number of short individually excited coils, 330 and 340, as illustrated in FIG. 3. Some Specific Geometries FIG. 8 is a graphical representation for understanding the generation of multipole fields that can be introduced by a lens corrector according to the embodiment of the present invention. Because excitation currents are d.c., or do not change rapidly with time, it is unnecessary to include vector potentials in the field description. Such a simplification allows the use of magnetostatic potentials, alone, for calculating the magnetic B-fields (the magnetic induction). The usefulness of this approach is that under these conditions the same equations are satisfied for magnetostatic fields as are satisfied for electrostatic fields with the driving potential for magnetostatic fields being ampere-turns rather than volts. However, it should be emphasized that such an analysis must not include the regions of current excitation which surrounds individual steel bars. Referring to equation (2) it can be seen that for quadrupole generation the difference between the magnetic potentials generated along each bar increases linearly from one end of the lens to the distant end. Thus, assuming uniformly spaced windings and equal currents through each winding, the loci of the associated magnetostatic equipotentials along each bar are straight lines that pass through zero at the center of each bar, because of symmetry. The B.sub.y(x) fields, which are produced between the bars, 314 and 316, described in FIG. 3, are excited by the negative gradient of the magnetostatic potential difference. As the distance between the high permeability steel tabs, 420 and 422, described in FIG. 4, is constant along the width of the lens/corrector, the difference between the magnetostatic potentials of each bar allows B.sub.y(x) to be calculated directly. Using this same presentation, FIGS. 9a and 9b show schematically the manner in which expansion (or contraction) of a ribbon beam ensemble can be accomplished. In FIG. 9a the magnetostatic equipotentials, 910 and 912, associated with a diverging lens, 930, in FIG. 9b produce a reduced-size ribbon beam, 950, starting from a fully expanded beam, 960, produced by equipotentials, 920 and 922. A simple linear change of all of the currents through all of the elementary coils, 330 and 340, allows expansion of the width of the ribbon beam to appropriate size before the ribbon beam impacts the wafer, 970. In FIGS. 10a and 10b, an individual beamlet, 980, is assumed to leave an ion source, 901, with intensity lower than anticipated for the remainder of the beamlets. To compensate for the reduced local ion density in the ribbon beam the fan-out pattern produced by the diverging lens, 930, is locally compressed around the attenuated beamlet, 980, by reducing the angular spacing between trajectories, 982, and 984. When satisfactory uniformity has been achieved at the entrance to lens 940, the overall spread of the fan is modified, as shown in FIGS. 9a and 9b, to allow uniform implantation of the whole work piece. It can be seen from the magnetostatic potential plot that for both bars forming the diverging lens, 930, the magnetostatic potentials, 924 and 926, no longer increase linearly from the center of each bar but rather has been reduced locally, at 925 and 927, to introduce a non-linearity in deflection angles for trajectories 984 and beyond that restores uniformity of implant intensity along the width of the ribbon beam. If necessary, angle corrections to compensate for this non-linear deflection can be introduced in lens, 940. There is a one-to-one correspondence between position along the final ribbon beam and the coil location along the first quadrupole bar allowing the computer correction algorithm to be simple and straight forward. FIGS. 11a and 11b show a method for introducing ribbon beam shifts along the x-direction or a rotation around the y-axis normal to the X-Z plane of FIG. 11b. Basically, to introduce a parallel shift all of the individual coils along both bars of the lens/corrector, 930, are electrically energized to produce a zero, 990, that is offset from the nominal center of the lens, 930. A compensating correction needed for the lens 990. To produce rotation about the y-axis the collimating currents through the lens 940, are adjusted appropriately to not return the output trajectories to being parallel to the ions leaving the source, 901. The principles used for producing the above offset in an alternate embodiment of the present invention are illustrated in FIG. 12. The coils, 330 and 340, illustrated in FIG. 3 and distributed along the bars, 314, and 316, are not energized and are left from the drawing to minimize confusion. The bucking coils, 332, 342, produce a uniform strip of magnetic B.sub.y-field, 328, that in the median plane is wholly parallel to the direction of the y-axis. Thus, there is no B.sub.x-field component along the x-direction so that it is not possible to induce motion out of the X-Z plane. Steering about the Y direction is fully decoupled from lens action and steering about the X-direction. FIGS. 13a and 13b, show a method for generating uniform B-fields along the x-direction. In FIG. 13a a pair of magnetostatic potentials, 1310 and 1316 are generated each having equal magnitude and direction along the individual bars with respect to one end. This can be achieved by energizing the coil collection, 330 and 340, shown in FIG. 3, uniformly and with the same hand. While the contribution to the magnetostatic potential from both bars would ideally be equal, it is possible for them to be unequal, as is shown in FIG. 13a. In practice, without exceptions, superposition allows all of these previously described field arrangements to be added together to produce a combination deflection structure that produces focusing, corrections of aberrations, corrections for differential variations in source output, and local steering across the ribbon ion beam around both X and Y axes. The constraint is that saturation should be minimal in the ferromagnetic members. A Useful Lens/Corrector Geometry FIG. 14 illustrates the design of a lens/corrector assembly consisting of two independent elements, 1430 and 1431, between which a ribbon beam can be directed through the slot, 322. Such a lens/corrector assembly, which is topologically identical to the rectangular steel bar structure illustrated in FIG. 3, has useful characteristics for insertion into the vacuum region of a beam-transport pipe and into the fringe field regions of a magnetic deflector where the vertical steel parts of the rectangular bar structure, 310, in FIG. 3, would short circuit the poles producing the magnetic deflection field. In principle, the vertical bars, 312, illustrated in FIG. 3, together with their associated windings, 332 and 334, have been severed at the central symmetry-point of each of the bucking windings. Referring again to FIG. 14, the bucking windings associated with the cut-away upper bar are labeled 1400, 1401. The windings that produce the focusing field are labeled 1410. After severance it should be arranged that the same current continues to pass through the resulting ‘half-windings’, 1400, 1401, so that when a lens/corrector is used in lens mode each resultant half winding will produce half the ampere turns as the original windings 332 and 334, illustrated in FIG. 3. Each element has three independently wound excitation coils that, if necessary, can themselves be wound as a collection of independent coils, 330, such as those shown in FIG. 3, to allow the introduction of multipole correction fields. Just as in the structure presented in FIG. 3 where the ampere-turns around the whole bar structure must integrate to zero, the symmetry of the independent element array, 1430 and 1431, requires that along the length of each element the total magnetostatic potential must integrate to zero. FIG. 14 illustrates the cross section of a quadrupole designed according to the above prescription. A ferromagnetic bar is located at the center of each element. This bar need not have a cylindrical cross section, but those skilled in the art will recognize that the cross-sectional area must be adequate to avoid saturation. Three independent winding sections, 1400, 1401 and 1410, are wrapped around each bar. To allow multipole generation and aberration correction the individual winding sections can themselves consist of a group of individually excited coils as was illustrated in FIG. 3, item 330. Ferromagnetic extension tabs, 420, introduced in the manner shown in FIG. 4, transfer the magnetostatic potential, generated along the length of the central steel bar, close to the boundary of the ribbon ion beam. The effect is to minimize the volume of magnetic field that must be produced and the needed ampere turns. Also, to improve the spatial resolution of the lens/corrector fields at an ion beam boundary in the lens aperture. Without reservations the bars and associated coil structures are enclosed within closed tubes, 1430, 1435, manufactured from a suitable non-magnetic material having rectangular cross section. This enclosing tube structure permits the outside walls of the tube to be in vacuum while power leads to the coils and air or water cooling is readily accessible through the ends, 1460 and 1461. A useful feature of the lens/corrector presented in FIG. 14 is while the total magnetostatic potential generated along each element must integrate to zero, it is not essential to pass equal currents through the windings within the elements 1430 and 1431. An unbalance in current ratio between the two elements changes the position of the neutral axis of the lens causing it to move in the Y-direction an introduce steering of an ion beam about the X-axis. Hydrogen Implanter FIG. 15, a further embodiment of the present invention, shows the principles of a high current H.sup.+ implanter for implanting ions into large-diameter semiconductor wafers using the ion transport elements described earlier. A suitable ion source, 10, produces a ribbon array of beamlets, 12, with all beamlets having the same energy, between 10 keV and 100 keV. A multipole corrected diverging lens, 20, introduces diverging angles into the array, 22, of beamlets to produce the necessary ribbon width. A momentum-dispersing magnetic field, 30, with its B-field vector in the plane of the diverging beamlets and approximately at right angles to the central beamlet of the array, deflects the ions at right angles to the plane of said ribbon beam allowing ions heavier than H.sup.+ to be collected into a cup, 40; this arrangement eliminates deuterium and other molecular contributions. A second multipole-corrected lens, 50, collimates the array of the diverging beamlets and returns the beamlets to parallelism. A platen supports a wafer, 60, and uniformly scans it, across the beam. This novel yet simple system employs no electromagnetic beam scanning. The advantages are short length, low cost, a simple optical path and small footprint. FIG. 16 shows the manner in which multiple-use coils can be mounted along a short section of one of the high permeability bars, 1617, to provide the high magnitude ampere-turns that are needed for exciting some deflection modes. It can be seen that continuous high-current capacity water-cooled coils, 1616, are wrapped as an under layer directly around a cylindrical magnetic core, 1617. Individually excitable coils, 1618, as shown in FIG. 3 as items 330 and 340, also surround the high permeability steel bar, 1615, to provide focusing and aberration corrections. Individual steel tabs 420, transfer the magnetostatic potentials to the region near to the beam. Any additional changes in the details, materials, and arrangement of parts, herein described and illustrated, can be made by those skilled in the art. Accordingly, it will be understood that the following claims are not to be limited to the embodiment disclosed here, and can include practices otherwise than those described, and are to be interpreted as broadly as allowed under the law. |
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051924919 | abstract | A boiling water reactor has a control element disposed in a water pond. An apparatus for the neutron-radiography testing of the control element includes a film cassette having a recording area. A water-free hood is fitted over the control element. The hood has an open bottom and a wall facing the film cassette. The wall has an opening formed therein with a cross section corresponding to the recording area. A holding device fits over the opening for receiving the film cassette. The holding device has a downwardly open side extended below the opening through which the film cassette is to be introduced into the water-free space in the holding device. A neutron source, a collimator and the film cassette lie in one measurement plane. The opening in the hood wall reduces the proportion of scattered neutrons and the impairment of image quality caused by a film of water remaining between the hood wall and the film cassette and by the hood wall itself, in prior art devices in which the film cassette is disposed outside the hood. |
claims | 1. A device containment apparatus for storing an explosive device and minimizing dispersal of radioactive material, the device containment apparatus comprising:a substantially spherical containment vessel for storing an explosive device, the vessel defining an interior area and including a door allowing selective access to the interior area;a first frame supporting the vessel and further supporting the door to be pivotally coupled to the frame and movable between a closed position preventing access to the interior area and an open position allowing access to the interior area, wherein the frame includes a base and an upper portion spaced a distance above the base;a first radiation shield including a plurality of radiation shielding panels supported on the first frame and extending between the base and the upper portion of the first frame in spaced relationship with the vessel; anda second frame and a second radiation shield supported by the second frame, the second radiation shield being positioned adjacent an outer wall of the vessel;wherein the second frame includes a first frame ring and a second frame ring, the first and second frame rings positioned at generally opposite ends of the vessel, and the second radiation shield includes a plurality of overlapping panels, each of the plurality of panels having a first end coupled to the first frame ring, a second end coupled to the second frame ring, and a shape complementary to the outer wall of the vessel. 2. The device containment apparatus of claim 1, wherein the first frame is substantially rectangular. 3. The device containment apparatus of claim 1, wherein the radiation shielding panels of the first radiation shield are arranged on the first frame along substantially planar sides that laterally surround the vessel. 4. The device containment apparatus of claim 1, wherein each of the plurality of panels of the second radiation shield includes a lead wool core encased in a nylon reinforced PVC covering. 5. The device containment apparatus of claim 1, wherein each of the plurality of panels of the second radiation shield includes a radiation shielding core encased within stainless steel plating. 6. The device containment apparatus of claim 1, wherein the plurality of panels of the second radiation shield include integrated seam plates to cover seams where adjacent ones of the plurality of panels meet. 7. The device containment apparatus of claim 1, wherein the first radiation shield includes a door shield covering an exterior surface of the door. 8. The device containment apparatus of claim 7, wherein the door shield is pivotable relative to the first frame. 9. The device containment apparatus of claim 1, wherein each of the plurality of radiation shielding panels of the first radiation shield includes a radiation shielding core encased within stainless steel plating. 10. The device containment apparatus of claim 7, wherein the door shield is coupled to the door to move with the door between the open and closed positions. |
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050193244 | abstract | The invention is directed to a centering arrangement for aligning two mutly adjacent openings of a lock and a container for receiving radioactive waste which is passed into the container through the lock. The container is provided with additional openings having a cross-sectional center point which is displaced from the cross-sectional point of the container per se. The centering arrangement includes a carrier which is pivotally journalled on a vehicle so as to be movable from a vertical plane into a horizontal plane and the vehicle is movable along rails in a horizontal direction. The longitudinal axis of the container is in a position aligned to a fixed point of the lock. The carrier includes a first container guide including an annular member having an inner conical surface and a second conical guide for positioning the container openings with respect to the rotational position of the container. The base of the carrier and the base of the container conjointly define an interface at which centering pins and corresponding counter openings are provided. The carrier is equipped with holders for holding the container in its centered end position. |
summary | ||
052271218 | claims | 1. A control room complex for a nuclear power plant, the plant having a multiplicity of components and sensors outside of the control room, the complex comprising: a main control room having at least one console which includes parameter indicators for displaying values of plant operating parameters, alarms for warning of an abnormal condition in a parameter or component, controllers for operating components and indicating the status of the controlled component, a screen for generating visual images of fluidly connected components, values of associated operating parameters, and component status, and means for manually tripping the reactor; a first type of digital processor means associated with the parameter indicators, alarms, and controllers; a second type of digital processor means associated with the screen; a plant protection system and associated third type of digital processing means, responsive to at least some of the plant sensors, for automatically tripping the reactor upon the occurrence of an unsafe event; a safeguards system for controlling at least some of the plant components upon the occurrence of an unsafe event; a component control system for controlling the plant components during normal operation; a power control system for controlling reactor power level; means for transmitting data from the protection system, the safeguards system, the component control system, and the power control system to each of the first and second digital processor means; means in each of the first and second types of digital processing means for independently computing representative values of plant parameters; means for transmitting data between the first and second types of digital processor means; means associated with the second type of processor, for providing said screen with display values of operating parameters that are based on a comparison of the representative values from the first and second types of processors. 2. The control complex of claim 1, wherein the first type of digital processor means is distributed at each console and the second type of data processing means is centralized remotely from each console. |
abstract | An X-ray radiography system for differential phase contrast imaging of an object under investigation by phase-stepping is provided. The X-ray radiography has an X-ray emitter for generating a beam path of quasi-coherent X-ray radiation, an X-ray image detector with pixels arranged in a matrix, and a diffraction or phase grating, in which the X-ray emitter has an X-ray tube with a cathode and an anode. The X-ray tube is constructed in such a way that an electron ray beam originating from the cathode is associated with focusing electronics which produce, from electrons which are incident on an anode, at least one linear-shaped electron fan beam. |
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description | Referring now to the figures of the drawings in detail and first, particularly, to FIGS. 1 and 2 thereof, there is seen a boiling water reactor and a conventional instrumentation system for monitoring and controlling in accordance with the prior art, which are considered below as a preferred example. FIG. 1 shows a typical configuration of detectors of various nuclear instrumentation systems. Fuel elements distributed over a cross section of a reactor core 1 are combined in each case to form cells 2 of four fuel elements, which are disposed around a common, cruciform control rod 3. The instrumentation includes three systems, specifically for a counter tube range or startup range AD (for a neutron flux P which reaches approximately up to 10xe2x88x925 of a normal flux for which the reactor is constructed), an intermediate range UD (where the neutron flux P is approximately 10xe2x88x926 to 10xe2x88x921) and a power range LD (where the neutron flux P is approximately from 10xe2x88x922), to which a measuring sensitivity of the assigned detectors and evaluation devices are tuned. Usually, a plurality of detectors are respectively disposed in instrument lances in the interior of the reactor core between the fuel elements. The evaluation devices of the systems are respectively combined to form redundantly operating channels in such a way that in the event of failure of a detector or the evaluation device thereof, or of the power supply, at most one channel fails. In that case, the assignment of the detectors and instrument lances to the channels can be linear (for example, FIG. 1 respectively provides three instrument lances 5 and 7 for the system of the counter tube range or startup range AD and the intermediate range UD, having detectors and evaluation units which respectively form a channel). However, it would also be possible to provide a plurality of detectors in each instrument lance and to network the evaluation devices in such a way that each channel is assigned detectors, evaluation devices and power supplies of different instrument lances. That networking is provided for the system of the power range LD. In order to form three monitoring channels, instrument lances 9, which are distributed over the entire cross section of the core, each carry at least three power distributor detectors that are disposed at a different level in each case, and provide measuring signals which can be used for the purpose of measuring the local power in the reactor, and of redundantly determining a three-dimensional power distribution through the use of three independent power distribution channels. In order to obtain an actual value (measuring signal) of a total power for the control or regulation of the reactor, the LD system includes three power range channels in which measuring signals of the power distribution detectors 9 are summed. It is possible in this case to exclude defective detectors or evaluation devices from the formation of the power distribution signals and/or a power range signal, and to compensate a distortion of the signals thereby produced by connecting or disconnecting further detector signals in the corresponding channel. Thus, in FIG. 1 each system is equipped for redundancy reasons with a plurality of mutually independent measuring channels which are similar to one another. In the case of the example shown in FIG. 1, there are three measuring channels per system. FIG. 2 illustrates a typical measuring range for each of those systems of the nuclear instrumentation. The counter tube range or startup range detectors (AD) measure the neutron flux P from the neutron source level (neutron flux of the completely shut down reactor core) up to a reactor power on the order of magnitude of 10xe2x88x925 of the rated power. In the case of irradiated cores, the source level 10 is typically at around 10xe2x88x929 of the rated power. However, in special cases, when many or all fuel elements of the core are fresh or only slightly irradiated, it can also be substantially lower. In the case of startup of a shutdown (subcritical) reactor core, because of repeated partial withdrawal of control rods, the source level continuously rises slightly until finally, a critical state (a self-maintaining chain reaction) is reached within the measuring range of the counter tube range detectors AD. Subsequently, the desired rate of rise of the neutron flux density (positive reactor period) is set by further withdrawal of control rods, in such a way that the measuring signals of the counter tube range detectors AD then indicate an exponential rise in the neutron flux density. The reactor operator can track the signal development over many decades and use it as a basis for his or her control tasks on the basis of a logarithmic signal display selected for the counter tube range detectors AD. If, while rising, the neutron flux density exceeds the magnitude of approximately 10xe2x88x926 of the rated power, the measuring range of the intermediate range detectors (UD) is reached and, in addition to the AD measuring channels, the measuring channels of those detectors indicate the respective magnitude and tendency of the neutron flux density. The UD measuring range also extends over many decades up to the power range. In order to raise the measuring accuracy, that large range is subdivided into a sequence of subranges which follow one another in the manner shown in FIG. 2 with pronounced mutual overlapping of measuring range, and are constructed, for example, with a sensitivity grading of 101/2:1 between successive subranges in each case. The switching over of the subranges is separately performed manually by the staff in the event of rising neutron flux density for each UD measuring channel when the signal approaches the upper measuring range limit of the subrange that is currently set. A trigger mark is set up for each subrange of a UD channel near the upper measuring range limit. If that mark is exceeded by the measuring signal, an alarm signal of the reactor protection system is automatically activated by the relevant UD channel as long as the overshooting lasts. If further such alarm systems are added from other UD channels, an emergency reactor shutdown (RESA) is initiated in accordance with an evaluation circuit provided for that purpose in the reactor protection system. In other words, all control rods which are partially or completely withdrawn are quickly inserted into the reactor core. In the process, the core changes over into the subcritical state, and the neutron flux density drops back to the source level. In the case of the example illustrated in FIG. 1, with three redundancies, the initiation of the RESA is generally undertaken whenever the trigger mark is exceeded in any two of the total of three UD channels (xe2x80x9c2 from 3xe2x80x9d evaluation circuit). If, in the case of the planned startup operation being considered, the reactor power approaches the upper measuring range limits of the AD channels of around 10xe2x88x925 of the rated power, the AD detectors are withdrawn downwards from the core for the purpose of extending the measuring range, and positioned below the reactor core in reflector positions with a neutron flux density which is reduced in comparison with the core interior. It may be seen from FIG. 2 that, because of the mutual overlapping of the AD and UD measuring ranges by more than a decade, the withdrawal of the AD detectors must not be instituted until the UD system has reliably taken over the monitoring of the neutron flux density. A suitable interlock device ensures that the AD detectors can in no way be withdrawn earlier from the core. It may be noted that in recent structures the functions of an AD channel and a UD channel are also combined in a so-called xe2x80x9cwide-range channelxe2x80x9d and can be fed by a single wide-range detector (WD). The measuring range of such a WD channel then includes the entire range, which is still divided into the ranges AD and UD in FIG. 2. If the reactor power exceeds the magnitude of approximately 10xe2x88x922 of the rated power, it is detected in addition by the power range channels LD seen in FIG. 2 and indicated for each channel on a linear scale having a measuring range which extends typically from 0% to 125% of the rated power. It is usual for there to be at least three LD channels present which are mutually independent and similar to one another. Each LD channel uses a multiplicity of the neutron-sensitive instrument lances (power distribution detectors xe2x80x9cLVDxe2x80x9d 9) as detectors. They are distributed virtually uniformly over the core volume and have signal contributions in the LD channel, which are calibrated in accordance with the local power density and are summed. The channels are calibrated to the reactor power at a suitable output load point with the aid of a thermal balance. The result thereof is that the accuracy of indication, the so-called xe2x80x9ctrack fidelityxe2x80x9d, of the LD channels for the reactor power is very high even in the case of arbitrary changes in power level and power distribution in the reactor core. The UD measuring channels, which are not able to measure the reactor power with a comparable, high track fidelity as in the LD channels, because of the use of individual detectors disposed locally in the core, are no longer required after and as long as the reactor power exceeds a minimum value which ensures the proper detection of the reactor power by the LD system. This minimum value is frequently fixed at 5% of the rated power. If it is exceeded, the UD channels can be taken out of engagement, for example by withdrawing their detectors from the reactor core and/or by bridging their starting functions. A series of trigger marks is included in the LD system. Of those trigger marks, the undelayed RESA triggering upon the attainment of a fixed overload limit mark typically disposed at approximately 120% of the neutron flux rated value is of importance with regard to countermeasures in the event of excursions. As in the case of the UD system, the triggering of that limit value by the various mutually redundant LD channels leads to the initiation of the countermeasures in accordance with the prescribed evaluation circuit. In the case of a 3-channel LD system, for example, that is usually likewise a xe2x80x9c2 from 3xe2x80x9d evaluation circuit (or the one already mentioned). The following sequence of an excursion results with the nuclear instrumentation of the nuclear reactor according to the prior art, that was described above by way of example. After attainment of the prompt critical reactor state in the counter tube range or in the intermediate range, for example due to defective withdrawal of control rods, the neutron flux density in the core increases so quickly that the changing over of the UD subranges can no longer keep up therewith (to the extent that the staff attempts that at all in such a case). In rapid sequence, the UD channels will exceed the RESA limit value of their respectively set subrange, and actuate the emergency reactor shutdown in accordance with the prescribed evaluation circuit. Due to the unavoidable delays in the signal processing and in electrical, mechanical and hydraulic components of the emergency shutdown system, the control rods are only inserted into the reactor core after a short delay. The detection of the rapid power rise by the UD channels is performed redundantly, with the result that the failure of a channel does not prevent the initiation of the countermeasures. An emergency reactor shutdown (RESA), or any sort of intervention in the planned startup operation which can be controlled, for example, through the use of a startup program, is intended to be performed in the startup phase (that is to say before the reactor power enters the typical measuring range of the LD detectors, in accordance with the provided startup operation) only if, because of a fault, for example an excursion, the reactor power reaches a limit value Mg which is situated in the measuring range of the detectors of the LD system (for example, 30% of the rated power P). This can be detected, according to FIG. 3, by feeding a reactor power S(t), for example the signal in at least one of the LD channels, to a limit value monitor Gg which activates a corresponding signal A to initiate the countermeasure, upon overshooting of the set limit value Mg. However, an improper overshooting of the limit value Mg during startup must be distinguished from proper operating states in the case of which the startup proceeds according to plan, or normal operation is already present. Consequently, according to FIG. 3, an AND gate AND1 is provided which acts as a filter that prevents such countermeasures in the case of proper operating states. A power band having a lower limit which is defined by a limit mark Mu is consequently placed below the trigger limit (limit value Mg). An output signal Sg in this case leads to the trigger signal A for the countermeasure only when the power band situated therebelow is traversed more quickly than in a prescribed minimum time xcex94tB (filter action of the power band). Although these operations are implemented as a rule as software for the monitoring, regulation and control present in the reactor, this filter action of the power band is represented in FIG. 3. In this case, upon overshooting of the lower limit mark Mu, which is set in a first limit value monitor Gu, the latter outputs a signal that triggers a flip flop K having a time constant that is equal to the minimum time xcex94tB and sending a corresponding pulse to an input of an AND gate AND2 in this time. Another input of this gate is connected to a second limit value monitor Go, which outputs a signal when the measuring signal S(t) for the reactor power exceeds a set upper limiting mark Mo. Thus, if the signal S(t) does not exceed the upper limit mark Mo until after the duration xcex94tB (xe2x80x9cstandby periodxe2x80x9d) has elapsed, the flip flop K (timing element) is already once again in idle state and both gates AND1 and AND2 block, with the result that activation of the trigger signal A by the signal Sg is prevented. If, however, because of an excursion, the power band is traversed quickly, that is to say the upper limit mark Mo is already exceeded inside the standby period xcex94tB, the signal Sg intervenes on the trigger signal A. The circuit layout having two AND gates and separate values for Mg and Mo permits the trigger signal A to be used, for example, to initiate a RESA triggering. However, a different triggering Axe2x80x2 of an output signal of the gate AND2 permits a less dramatic countermeasure to be instituted. Specifically, if only the signal Axe2x80x2 responds, but not the signal A, there can be faults which restabilize themselves even without a reactor shutdown, to the extent, for example, that the programmed startup operation is only slowed down or stopped. However, a simplification is achieved when the same upper limit mark is used for Mo and Mg. It is then possible in the case of a hardware circuit to dispense with the limit value monitor Gg and the gate AND1. In other words, no distinction is then made between strong excursions in the case of which both the power band (Moxe2x88x92Mu) is traversed more quickly than the time xcex94tB and the trigger limit Mg is exceeded, and weak excursions in which only the rise criterion is fulfilled but the trigger limit is not reached. The limit marks can be prescribed as a function of the respective control program (that is to say, for example, the selected control rods) or the power signal S(t) or can be prescribed in some other way as a function of operation. In the limit value monitors Gu and Go, the respective limit value Mu and/or Mo for limiting the power band is advantageously independent of operation. In this case, the power band is advantageously situated in the lower third, preferably in the lower quarter, of the rated power of the reactor. The width, that is to say the difference Moxe2x88x92Mu, and/or the minimum time xcex94tB is preferably prescribed in an operationally independent manner. It generally amounts to less than one third, preferably less than one fifth, of the rated power. The above-described diversitary excursion monitoring is preferably constructed separately for each LD channel. In accordance with an electronic evaluation provided for this purpose, the excitations possibly arriving from the various LD channels, which are mutually redundant, lead to initiation of the excursion countermeasure. The emergency reactor shutdown preferably comes into consideration as such a countermeasure. The following points of view are important in establishing the position and width of the power band and of the standby period xcex94tB, for the purpose of uniquely distinguishing between excursions and startup operations which are according to plan and possibly accelerated. The power band is expediently to be fixed in such a way that power rises of planned startup operations in any case have either already stabilized themselves below the band due to the reactivity feedback effects which are relevant in this case, or stabilize themselves at the latest within the power band, while it is traversed largely undamped in the case of all excursions. This condition is best fulfilled by the lowermost part of the measuring range of the LD channels which is useful for setting limit marks. The width of the power band (that is to say the power difference Moxe2x88x92Mu) is to be so large that the power band with the highest load change rate which can be realized in this power range in the case of normal operation is traversed on the order of magnitude of approximately one minute. It is then possible to select the length of the standby time period xcex94tB which is suitable for unambiguously distinguishing between excursions and planned startup procedures, from a relatively large time range 1 s less than xcex94t less than 1 min. Upon the observance thereof in terms of control engineering, it is then unnecessary to place any further high accuracy requirements. A parameter combination with Mu=5% of the rated power Mo=20% of the rated power xe2x80x83xcex94tB=20 s may be specified as an example which in general fulfills the optimization criteria described above. It would not be damaging if primary excursions occur which are already restabilized in a few fractions of a second and are therefore virtually not detected, for example, due to unavoidable inertias and dead times of hardware (detectors) and software (electronic evaluation). This is also not necessary, since the countermeasure must be initiated only when a measurable rise in power occurs over relatively long times (for example above one second). This is illustrated symbolically in FIG. 3 by a filter for the signal S(t), for example a smoothing filter F at the input of the limit value monitor Gu. FIG. 4 initially indicates a limit value M(LD) which is already prescribed in the prior art for the corresponding signal S(t) of the power range channels, in order to undertake a reactor shutdown in normal operation, for example in the case of power fluctuations which are caused by hydraulic instabilities, or in the case of similar malfunctions. A typical limit value M(LD) is, for example, 120% of the rated power P of the reactor. FIG. 4 also illustrates the power band by showing the upper limit mark Mo and the lower limit mark Mu. It is also assumed that at an instant t1 a primary excursion and any consequent excursions triggered thereby lead to a rise in the power in the reactor indicated by the signal S(t) even before the reactor should actually reach a power of approximately 5% of the rated power, on the basis of the planned startup procedure. The reactor is therefore in a state in which the UD systems according to the prior art take over the monitoring of the reactor. According to the invention, however, the signal S(t) of the LD channels is used for monitoring. For this purpose, upon overshooting of the lower limit mark Mu for the standby time period xcex94tB, a monitoring logic circuit is firstly activated. Upon overshooting of the upper limit mark Mo which simultaneously represents the triggering limit for the countermeasure, this triggers the corresponding monitoring signal with the aid of which the countermeasure is initiated. As a result of this countermeasure (for example a shutdown in the case of which all control rods are once again inserted completely into the core), the reactor power decays again without having reached a dangerous value. The operating staff can then once again cancel the countermeasure and institute a new startup procedure according to a better planned program, or resume the startup procedure interrupted by the countermeasure, in a planned manner. In this planned startup procedure, the reactor power then grows according to the plan and reaches the values at which the LD systems supply a display of the reactor power which has track fidelity, and the signal S(t) is therefore monitored by the LD systems according to the prior art. In this case, the LD signal S(t) once again overshoots the lower limit mark Mu at an instant t2. However, this limit mark is in a power range (for examplexe2x89xa75%) in which a reactor that has been started in a planned manner and appropriately heated up exhibits no more excursions. During the planned power increase, the signal S(t) now traverses the power band at a correspondingly low rate of rise which does not lead to overshooting of the upper limit mark Mo until an instant t3, that is to say long after expiration of the standby time period xcex94tB. FIG. 5A and FIG. 5B initially illustrates the AD channel detectors and instrument lances 5 provided for the counter tube range or startup range, having signals that can be evaluated, documented and displayed on monitors, for example in a startup electronic monitoring unit 51. The intermediate range channels UD with lances and detectors 7 can likewise have corresponding monitors 52. These channels furthermore have a device 53 for stepwise reduction of the measuring sensitivity of these channels. This device 53 is indicated in FIG. 5A and FIG. 5B by a symbol of a divider. A corresponding reduction factor for the sensitivity of operating staff is changed, for example, as soon as it can be detected on the monitors 52 that the corresponding signal of the UD channels exceeds the currently set sensitivity range (see FIG. 2). The conventional monitoring of the startup procedure through the use of the UD channels includes a logic circuit monitoring device 54 which outputs a signal in each case upon overshooting of the sensitivity range (which is measurable, for example, in each case by one limit value monitor 55 for each of the three UD channels illustrated). If, in the case of the installed number n of UD channels (in this case: n=3), at least one prescribed number m (in this case: m=2) reports such range overshooting, a corresponding xe2x80x9cm from nxe2x80x9d circuit (in this case: a xe2x80x9ctwo from threexe2x80x9d circuit 56) outputs a signal a with the aid of which an intervention is made into the control of the reactor. This reactor control is symbolized in FIG. 5A and FIG. 5B as a corresponding control and regulation device 60 having a controller 61 and an actual-value arithmetic circuit 62. For example, a control center can be provided with a display or monitor 83 for the reactor power. A +/xe2x88x92 switch is operated by the staff in order to insert the control elements into the core or extract them from it when a lower or higher desired power value is targeted. All desired values and parameters for the planned control of the reactor are set at this device 60. Actuating signals for the corresponding manipulated variables of the reactor operation are largely program-controlled in the device 60 and formed automatically through the use of a plenitude of actual values determined in the reactor and actuator check-back signals which are symbolized in FIG. 5 by actual values S1, S2 and S3 that are formed by sensors and detectors of the LD channels. The actuating signals themselves are relayed to a corresponding actuating device 70 which, in addition to other actuating devices (for example coolant pumps for a coolant loop), chiefly controls drives for inserting and extracting the control rods into and from the reactor. Therefore, if the monitoring device 54 of the UD channels supplies a corresponding signal a in the prior art, an intervention is made into the reactor operation controlled by the device 60. That is done, for example, by having the controller 61 change to a shutdown program, or by using a computer initially to change directly in the device 70 from the programmed control of the control rods to shutting down the reactor by inserting the control rods. The LD channels are illustrated in FIG. 5A and FIG. 5B by the corresponding instrument lances and detectors 9. The signals of defective sensors are suppressed in a corresponding electronic evaluation device 80 by using plausibility criteria, in order to network in a corresponding signal interconnection unit 81 only plausible measuring signals from the sensors distributed over the volume of the reactor core, and to respectively form in a subsequent summing circuit 82 an aggregate signal for each LD channel which respectively detects the local distribution of the power in the reactor as a signal S1, S2 and S3 for the power of the reactor. Each signal S1, S2 and S3 of these redundant power range channels is monitored in a limit value monitor 84 for overshooting of the limit value M(LD). If at least two of the three signals (in general: at least one number m from the number n of the limit monitor signals) exceed this limit value, an appropriate evaluation circuit 85, that is to say a xe2x80x9ctwo from threexe2x80x9d circuit (in general: xe2x80x9cm from nxe2x80x9d circuit) outputs a corresponding signal b. With the aid of the signal b it is possible to activate a shutdown program (or another suitable countermeasure) in the control device 60 or, as illustrated in FIG. 5A and FIG. 5B, to intervene directly in the actuating device 70 for the control elements, in order to interrupt the normal startup procedure and, if appropriate, to initiate an emergency shutdown. It is not illustrated that signals, combined by the signal interconnection unit 81 in power distribution channels, from detectors of the instrument lances 9 are likewise monitored in three redundant channels for local oscillations. They can likewise trigger the signal b, in order to detect the three-dimensional power distribution. The power range monitoring by the evaluation device 80 therefore serves the purpose of redundantly monitoring corresponding measuring signals, in accordance with the prior art. The measuring signals are continuously formed by a plurality of power range channels and detect the reactor power. However, such a redundancy with the aid of the signals S1, S2 and S3 is also provided for the monitoring according to the invention in the counter tube range. The actuating device 70 has a corresponding input A with the aid of which it is possible, at least during the startup operation, to apply a trigger signal supplied by an appropriate device to the actuating device 70. Therefore, in accordance with FIG. 6 each of these signals S1, S2 and S3 is assigned a device C1, C2 and C3 corresponding to FIG. 3. Consequently, their signals A1, A2 and A3 and, if appropriate, their signals Axe2x80x21, Axe2x80x22 and Axe2x80x23 as well, are also likewise combined by an xe2x80x9cm from nxe2x80x9d circuit 91 or 92 to form a corresponding signal A0 or Axe2x80x20 with the aid of which the trigger signal A can be set. The reactor power is therefore continuously detected through the use of measuring signals of a plurality of power range channels, and redundantly monitored. The invention can be used to replace the conventional monitoring of the startup through the use of the signal a by monitoring through the use of a signal A, and A1, A2, A3 or A0, respectively, that is to say by a monitoring which uses other sensors and other channels. However, this monitoring of the startup is advantageously used as diversitary monitoring. Thus, in addition to the monitoring through the use of the power range channels and their detectors 9, the signals of the neutron flux detectors 7 are also monitored through the use of the signal a and, if appropriate, used for a countermeasure in order to limit the rise in power. These additionally used neutron flux detectors already monitor the neutron flux of the reactor core according to the prior art in the manner shown in FIG. 2, for observance of a current maximum value which is prescribed as a function of operation, as long as the rods are drawn out of the core and the reactor power is still in the intermediate range. This maximum value is virtually the upper limit of the measuring range respectively prescribed in FIG. 2 for each sensitivity stage. In accordance with the invention, it is possible to achieve a high reliability in the counter tube range on the basis of a prescribed redundancy. In accordance with FIG. 5B, for example, such a redundancy resides in the fact that a plurality of redundantly operating logic circuits connected to an evaluation circuit process the signals of the power range channels. The further redundancy due to the use as a diversitary system is important for the monitoring device 54. The actuator 70 for the control rods can therefore be activated not only by the signal of a monitoring device according to FIG. 3 or FIG. 6, but additionally also by the monitoring device 54 which is connected to the neutron flux detectors 7 and is constructed by using other physical principles and detectors. |
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042749204 | claims | 1. A water-cooled nuclear reactor which operates at an elevated temperature and pressure during normal operation having passive emergency shutdown and core cooling capability, comprising: a reactor core comprising a plurality of columnar arrangements of nuclear fuel elements disposed during normal operation within a pressure vessel in a chain reacting relationship, said pressure vessel containing at least sufficient energy to disperse the bulk of the core fuel content by driving it from the core; a plurality of hollow fuel dispersal guide conduits penetrating said pressure vessel at locations beneath said columnar arrangements and generally aligned therewith, said dispersal guide conduits being generally divergent at least along portions of their length substantially immediately beneath said pressure vessel for receiving nuclear fuel elements therefrom in the event of emergency shutdown and for guiding and dispersing said fuel elements whereby said fuel elements become spaced sufficiently to assume a non-critical dispersed configuration; an external heat sink of capacity adequate to absorb the residual heat contained in the reactor core and spontaneously generated by said nuclear fuel elements after shutdown, said heat sink being disposed beneath said reactor core so that gravitational attraction aids in driving said nuclear fuel elements along said dispersal guide conduits into said dispersed configuration for cooling said nuclear fuel elements in the event of emergency shutdown; means positioned in series with each guide conduit for communicating said conduit with said pressure vessel for initiating transfer of fuel elements from said pressure vessel into said associated dispersal guide conduit in the event of emergency shutdown; and means positioned in series with each guide conduit and adjacent said heat sink for permitting selective withdrawal of fuel elements from, and/or loading fuel elements into, said associated conduit. a reactor core comprising a plurality of columnar arrangements of nuclear fuel elements disposed during normal operation within a pressure vessel in a chain reacting relationship, said pressure vessel containing at least sufficient energy to disperse the bulk of the core fuel content by driving it from the core; a plurality of hollow fuel dispersal guide conduits configured to diverge to a non-critical dispersed configuration penentrating said pressure vessel at locations beneath said columnar arrangements and generally aligned with said arrangements for receiving nuclear fuel elements therefrom in the event of emergency shutdown and for guiding and dispersing said fuel elements whereby said fuel elements become spaced sufficiently to assume a non-critical configuration. an external heat sink in the form of a body of water of capacity adequate to absorb the residual heat contained in the reactor core and spontaneously generated by said nuclear fuel elements after shutdown, said heat sink being disposed beneath said reactor core so that gravitational attraction aids in driving said nuclear fuel elements along said dispersal guide conduits into said dispersed configuration, and for cooling said nuclear fuel elements by conduction and convection in the event of emergency shutdown; means positioned in series with each guide conduit between said pressure vessel and said heat sink communicating the conduit with said pressure vessel for initiating transfer of fuel elements from said reactor vessel into said associated dispersal guide conduit in the event of emergency shutdown; and means positioned in series with each guide conduit and positioned beneath said heat sink in a reactor refueling station for permitting selective withdrawal of fuel elements from, and/or loading fuel elements into, said associated conduits. a reactor core comprising a plurality of columnar arrangements of nuclear fuel elements disposed during normal operation within a pressure vessel in a chain reacting relationship, said pressure vessel containing at least sufficient energy to disperse the bulk of the core fuel content by driving it from the core; a plurality of generally divergent hollow fuel dispersal guide thimbles in the form of hollow tubular members penetrating said pressure vessel at locations beneath said columnar arrangements and generally aligned with said arrangements for receiving nuclear fuel elements therefrom in the event of emergency shutdown and for guiding and dispersing said fuel elements whereby said fuel elements become spaced sufficiently to assume a non-critical configuration; an external heat sink of capacity adequate to absorb the residual heat contained in the reactor core and spontaneously generated by said nuclear fuel elements after shutdown, said heat sink being disposed sufficiently adjacent said reactor core to permit the energy contained in said vessel to drive said fuel elements along said dispersal guide thimbles into a non-critical dispersed configuration and to cool said fuel elements by conduction and convection by said heat sink; means positioned in series with each guide thimble on a first side of said heat sink between said pressure vessel and said heat sink for communicating the thimble said pressure vessel for initiating transfer of fuel elements from said pressure vessel into said associated dispersal guide thimbles in the event of emergency shutdown; and means positioned in series with each guide thimble on a second side of said heat sink distant from said pressure vessel for selectively withdrawing fuel elements from and/or loading fuel elements into said associated thimble. 2. A reactor according to claim 1 wherein said heat sink comprises a body of water accommodated within the seismic support region of the reactor, and emergency gating means are provided for permitting gravitational replenishment of evaporated water while preventing the flow of said water out of the heat sink in the event of fuel dispersal guide rupture. 3. A nuclear reactor according to claim 1 including guide means within said pressure vessel for constraining the fuel elements comprising said columnar arrangements in alignment with said fuel dispersal guide conduits while permitting upward and downward slidable movement thereof along said guide means and into or out of said dispersal conduits. 4. A nuclear reactor according to claim 1 further comprising longitudinally straight-edged guide means radially positioned around each of said columnar arrangements for constraining the fuel elements of said arrangements in alignment with said fuel dispersal guide conduits while permitting upward and downward slidable movement along said straight edged guide means and into or out of said dispersal guide conduits. 5. A nuclear reactor according to claim 1 wherein a plurality of fuel elements comprising a plurality of columnar arrangements, each has a generally longitudinally extending hollow interior portion and is disposed in said reactor core on a guide member inserted into said hollow interior portion for constraining said fuel elements in alignment with said fuel dispersal guide conduits while allowing upward and downward slidable movement thereof along said guide member and into or out of said dispersal guide conduits. 6. A nuclear reactor according to claim 5 wherein a pluraltiy of said columnar arrangements comprises one or more hollow dummy elements positioned at the levels of one or more respective core support frames for limiting horizontal movement and vibration of said columnar arrangement. 7. A nuclear reactor according to claim 5 wherein the uppermost element of said columnar arrangement comprises a hollow dummy spacer with an enlarged funnel-shaped receptacle for facilitating engagement with said guide member during fuel loading. 8. A nuclear reactor according to claim 1 wherein the portion of the reactor beneath said columnar arrangement is provided with a surface shaped to funnel molten material from said columnar arrangement into said dispersal guide conduits. 9. A nuclear reactor according to claim 1 wherein said means for initiating transfer of said fuel elements from said core into each dispersal guide conduit comprise a piston member positioned within, and slidable along, the hollow interior of said conduit and held in position at the pressure vessel by pressurized fluid for supporting said fuel elements during normal operation of the reactor. 10. A nuclear reactor according to claim 9 wherein said pressurized fluid used to hold said piston in position is comprised of reactor coolant quality water. 11. A nuclear reactor according to claim 9 wherein said piston member comprises sealing means for sealing said fuel dispersal guide conduit from said reactor pressure vessel during normal core operation. 12. A nuclear reactor according to claim 9 wherein said guide conduits comprise sealing shoulders for receiving said piston in sealing engagement while said piston member is held in position for supporting said fuel elements during normal core operation. 13. A nuclear reactor according to claim 1 wherein a plurality of said fuel elements comprise clad fuel elements of nuclear fuel having a maximum longitudinal dimension not exceeding about three times the maximum transverse dimension in order to facilitate entry into said fuel dispersal guide conduits. 14. A nuclear reactor according to claim 1 wherein a hollow tubular member of neutron absorbing material is disposed around at least a portion of the emergency dispersal path of said fuel elements. 15. A nuclear reactor according to claim 1 wherein a plurality of said fuel elements having hollow interiors are provided with a plurality of longitudinally extending concave grooves for permitting cladding expansion with minimal reduction of cladding wall thickness upon fuel swelling. 16. A nuclear reactor according to claim 15 wherein a plurality of said fuel elements having hollow interiors are provided with a cladding tightly compressed onto a fuel pellet for facilitating heat transfer from fuel to coolant. 17. A nuclear reactor according to claim 16 wherein a plurality of said fuel elements with compressed cladding are provided with cladding which is of reinforced thickness as compared to the thickness of cladding required to obtain comparable heat transfer in a corresponding, non-compressed structure. 18. A nuclear reactor according to claim 1 wherein a plurality of said fuel elements are provided with a plurality of longitudinally extending concave grooves for permitting cladding expansion with minimal reduction of cladding wall thickness upon fuel swelling. 19. A nuclear reactor according to claim 18 wherein a plurality of said fuel elements are provided with a cladding tightly compressed onto a fuel pellet for facilitating heat transfer from fuel to coolant. 20. A nuclear reactor according to claim 19 wherein a plurality of said fuel elements with compressed cladding are provided with cladding which is of reinforced thickness as compared to the tickness of cladding required to obtain comparable heat transfer in a corresponding, non-compressed structure. 21. A nuclear reactor according to claim 1 wherein a plurality of said fuel elements are assembled into a multi-element cluster. 22. A nuclear reactor according to claim 1 comprising a plurality of normally open isolation valves for permitting fuel discharge into said fuel dispersal guide conduits. 23. A reactor according to claim 1 including a reactor refueling station coupled to said fuel dispersal guide conduits for on-line refueling. 24. A reactor according to claim 1 wherein said means associated with each conduit for communicating said conduit with said pressure vessel for initiating transfer of fuel elements comprises a pressure barrier to provide sealing engagement between said reactor pressure vessel and said dispersal guide conduits during normal reactor operation. 25. A nuclear reactor according to claim 1 wherein a plurality of said columnar arrangements of nuclear fuel elements comprise one or more dummy spaces for supporting said nuclear fuel elements in the operative region of the reactor core. 26. A water-cooled nuclear reactor which operates at an elevated temperature and pressure during normal operation having passive emergency shutdown and core cooling capability, comprising. 27. A nuclear reactor according to claim 26 further comprising a refueling station positioned beneath said heat sink for withdrawal of fuel elements from, and/or loading fuel elements into, said associated conduits. 28. A water-cooled nuclear reactor which operates at an elevated temperature and pressure during normal operation having passive emergency shutdown and core cooling capability, comprising: |
054209021 | description | Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is seen a fuel assembly for a boiling water reactor in cross section, which includes 11 rows and columns of fuel rods. A rectangular fuel assembly cross section is assumed, and fuel rods located in corners ("corner rods") are shown at reference numeral 1, fuel rods ("peripheral rods") adjacent to a fuel assembly channel ("water channel" WC) are shown at reference numeral 2, fuel rods adjacent to a rectangular coolant pipe ("water pipe" WR) disposed in the center of a rod cluster or bundle are shown at reference numeral 3, and remaining "inner rods" are shown at reference numeral 4. A foot that closes off the channel WC and covers the cluster is located below the plane of the drawing and has flow openings through which water acting as a coolant and moderator is introduced into the fuel assembly from below. The water emerges again at corresponding flow openings from a plate in a fuel assembly head that covers the channel WC and the rod cluster above the plane of the drawing. The fuel rods 1-4 are each guided by one mesh or opening of a grid-like spacer. These meshes, which define the spacer, in this preferred embodiment are formed by cans 5 that serve as inner ribs and surround the inner rods 4, the peripheral rods 2 and the corner rods 1. The cans 5 are laterally surrounded by outer peripheral ribs 6 and 7 and inner peripheral ribs 8 and 9. At least one support spring 10 is disposed in each mesh for pressing the corresponding fuel rod against opposed knobs or bumps 10'. The drawing also shows that the walls of the water channel WC have a normal thickness only at their corners and at the foot of the fuel assembly, while in middle parts the wall thickness is reduced. Under some circumstances, grooves or slight protrusions could be provided along the inner wall, acting as flow trippers. Otherwise, the channel walls are flat. Tabs 11 and 11' are respectively disposed on the outer peripheral ribs 6 and 7 and on the inner peripheral ribs 8 and 9. The tabs begin at the upper edge of these ribs and are bent into interstices between peripheral ribs. FIGS. 2 and 3 show that each of the support springs 10 which are each disposed in a respective one of the meshes guiding a fuel rod, has an upper bearing surface 12 and a lower bearing surface 13. According to FIG. 3, the bearing surfaces 12 and 13 rest on a front side 14 of a rib 15 surrounding the mesh. The front side 14 faces toward the rod. These bearing surfaces 12 and 13 are joined to two flat legs 17 and 18, each of which adjoins one bearing surface, through a singly bent, resilient middle part 16 facing toward the fuel rod. The bearing surface 12 changes into a spring end 19, which advantageously protrudes beyond a contacting edge of the front side 14 of the rib 15. A corresponding spring end 20 on the bearing surface 13 is advantageously bent around a contacting edge of the rib 15 and extends on the back side of the rib 15 as far the spring end 19, where these two spring ends are joined to one another. Since the two spring ends are fastened together, the rib 15 is enclosed by the spring. Therefore, the spring cannot fall off. On the back side of the rib 15 facing away from the rod, the two spring ends encompassing the rod may form a spring part 28 ("double spring") that is mirror-symmetrical to the bearing surfaces 12 and 13 and the singly bent middle part 16. The front side 14 also has a protrusion 21 pointing to the bent middle part 16, and if the spring is constructed with mirror symmetry, then a protrusion 22 pointing to the mirror-image spring part 28 is also provided on the back side. As FIG. 4 shows, the spring 10 may encompass two cans that are fastened together and form the meshes of two adjacent fuel rods. Accordingly, a further inner rib 23, which is also enclosed by the spring and on which the adjacent rod is supported, rests on the inner rib 15. FIG. 4 also shows two bearing elements 25, 26 that are constructed as rigid knobs on the rib 15. The disposition of the bearing elements 25, 26 on the rib 15 is shown in FIGS. 5 and 6. FIGS. 7 and 8 clearly show that in two different configurations of the fuel rods in a boiling water fuel assembly with a rectangular cross section, the outer peripheral ribs 6 and 7 are joined only by the inner ribs in the form of the cans 5, so that these ribs 6 and 7 form an outer frame around the spacer that has a gap with an opening angle .alpha. that is close to 90.degree.. An edge of the rib 15 resting on the spring end 20 (and naturally the corresponding edge of the rib 21 as well in the case of a double-sided spring) has a recess 24 formed therein that is engaged by the spring end 20. This prevents lateral shifting of the spring. An equivalent recess may be provided on both edges of the rib 15, as FIG. 5 shows. While FIGS. 1-9 show spacers in which all of the inner ribs are constructed as cans and all of the fuel rods are seated in these cans, FIG. 10 shows an embodiment in which only the inner rods are seated in meshes formed by inner ribs 30, 31, while these inner ribs are fastened, for instance by welding, to outer ribs 32, 33, so that the meshes with the peripheral rods are each formed partly by inner ribs and partly by peripheral ribs. Once again, however, the peripheral ribs may be joined together only by inner ribs. This is possible especially if knobs 35 are provided on the corners of these outer rods. The knobs fix the outer rods relative to a fuel assembly channel that rests on them. One such knob is shown in FIG. 11 as well as in the sectional views of FIGS. 12 and 13, which are taken along the lines XII--XII and XIII--XIII, and which include two halves 37, 38 of a toroidal bead that are shaped as mirror images of one another. The toroidal bead is formed onto an outer rib 39. The springs or tabs 10 that were already mentioned in conjunction with FIG. 1 may also have small knobs 40 seen in FIGS. 14-16, which are located downstream of the knob halves in terms of the flow and stabilize the position of the spacer without substantially contributing to the flow resistance. FIG. 17 shows that lower edges 50 of the outer peripheral ribs 6 and 7 cover corresponding lower edges of the inner peripheral ribs, because these edges are located practically in the same plane. The same is true for the upper edges of the tabs 10 and 11 at the inner and outer peripheral ribs. However, between these tabs, the outer peripheral ribs 6 and 7 are lower than the inner peripheral ribs that have upper edges 51 which are therefore visible in FIG. 17. As compared with the base of the tabs 10 at the outer peripheral ribs, the base of the tabs 11 on the inner peripheral ribs is accordingly smaller, and correspondingly the edges located between the tabs of the inner peripheral ribs are wider. They accordingly form suitable stop surfaces, against which corresponding stops 60 that are welded to the coolant pipe WR and are seen in FIG. 19 can strike, without substantially hindering the coolant at the fuel rods seated there. It is therefore unnecessary to make the stops 60 protrude so far beyond the spacer that they would have to strike the inner ribs (such as a can 61 in FIG. 19) secured to the inner peripheral rib. Instead, the stops 60 do not touch any inner ribs. These stops are then advantageously each provided in the middle of a pipe wall, and two stops at diametrically opposed points on the channel are sufficient. FIG. 17 also shows corner tabs 62 at the inner peripheral ribs, and FIG. 19 shows that two such corner tabs 62, 63 of abutting peripheral ribs are welded together. Although a corner tab of this kind only has a narrow connection to the respective inner peripheral rib because of the tight space available, nevertheless it is secured against being torn off by being welded to another corner tab. The overall result is that a stable edge is created at the inner peripheral ribs. FIG. 20 shows that each inner peripheral rib 8, 9 of FIG. 19 has one contact part 70, 71 springing back toward an adjacent fuel rod, or in other words pointing away from the pipe wall. These parts are formed by upper and lower peripheral parts. A can-like inner rib 72 is fastened to this peripheral part, for instance by welding, forming a mesh that guides the fuel rod. FIG. 20 shows that an adequate stop surface area for the stop 60 fastened to the pipe wall WR is still assured in that case. As FIG. 21 shows, a middle part 78 may also be provided between corresponding upper and lower peripheral parts 75, 76, which are required for fastening a can-like inner rib 77. A spacer element supporting the peripheral rib against the wall of the pipe WC may be disposed on this middle part 78. A spring 79 is provided as the spacer element and is constructed similarly to the spring of FIG. 3, except that in this case no symmetrical embodiment is needed, nor is it necessary to have a stop that prevents over-stretching of the spring when a fuel rod is inserted. The various provisions shown herein taking an especially advantageous spacer for a boiling water fuel assembly and can-like inner ribs as an example, are largely independent of one another. Accordingly, they need not be employed in combination. As a result, individual provisions according to the invention may also be applied to spacers having square meshes, for example, without departing from the scope of the invention. The embodiment of the spring which was described in detail at the outset may also be employed in fuel assemblies for pressurized water reactors. |
description | The present application claims priority from Japanese Patent Application No. 2011-189316 filed Aug. 31, 2011. 1. Technical Field This disclosure relates to a target supply unit. 2. Related Art In recent years, semiconductor production processes have become capable of producing semiconductor devices with increasingly fine feature sizes, as photolithography has been making rapid progress toward finer fabrication. In the next generation of semiconductor production processes, microfabrication with feature sizes at 60 nm to 45 nm, and further, microfabrication with feature sizes of 32 nm or less will be required. In order to meet the demand for microfabrication with feature sizes of 32 nm or less, for example, an exposure apparatus is needed in which a system for generating extreme ultraviolet (EUV) light at a wavelength of approximately 13 nm is combined with a reduced projection reflective optical system. Three kinds of systems for generating EUV light are known in general, which include a Laser Produced Plasma (LPP) type system in which plasma is generated by irradiating a target material with a laser beam, a Discharge Produced Plasma (DPP) type system in which plasma is generated by electric discharge, and a Synchrotron Radiation (SR) type system in which orbital radiation is used. A target supply unit according to one aspect of this disclosure may include: a nozzle unit having a through-hole to allow a target material to be outputted therethrough; a cover provided to cover the nozzle unit, the cover having a through-hole to allow the target material to pass therethrough; and a discharge device configured to pump out gas inside a space defined by the cover. A target supply unit according to another aspect of this disclosure may include: a nozzle unit having a through-hole to allow a target material to be outputted therethrough; an electrode provided to face the nozzle unit; a voltage generator configured to apply a voltage between the target material and the electrode; and a discharge device configured to pump out gas in at least a space between the nozzle unit and the electrode. A target supply unit according to yet another aspect of this disclosure may include: a nozzle unit having a through-hole to allow a target material to be outputted therethrough; a plurality of electrodes provided in a direction in which the target material travels; an electrical insulator for holding the plurality of electrodes; at least one voltage generator configured to apply a voltage between the plurality of electrodes; a cover provided to cover the nozzle unit, the plurality of electrodes, and the electrical insulator, the cover having a through-hole to allow the target material to pass therethrough; and a discharge device configured to pump out gas in a space defined by the cover. Hereinafter, selected embodiments of this disclosure will be described in detail with reference to the accompanying drawings. The embodiments to be described below are merely illustrative in nature and do not limit the scope of this disclosure. Further, the configuration(s) and operation(s) described in each embodiment are not all essential in implementing this disclosure. Note that like elements are referenced by like reference numerals and characters, and duplicate descriptions thereof will be omitted herein. Contents 1. Overview 2. Terms 3. Overview of EUV Light Generation System 3.1 Configuration 3.2 Operation 4. Chamber Including Electrostatic-Pull-Out Type Target Supply Unit 4.1 Configuration 4.2 Operation 5. Electrostatic-Pull-Out Type Target Supply Unit 5.1 Configuration 5.2 Operation 5.3 Effect 6. Target Supply Unit Including Acceleration Electrode 6.1 Configuration 6.2 Operation 6.3 Effect 7. Target Supply Unit Including Cover for Shielding Nozzle Unit 7.1 Configuration 7.2 Operation and Effect 8. Target Supply Unit Including Cover for Shielding Reservoir and Nozzle Unit 8.1 Configuration 8.2 Operation and Effect 9. Target Supply Unit Including Position-Adjustable Cover 9.1 Configuration 9.2 Operation and Effect1. Overview In an LPP type EUV light generation apparatus, a target may be supplied from a target supply unit in the form of droplets toward a plasma generation region inside a chamber. The target material may be irradiated with a pulse laser beam when the target material reaches the plasma generation region. Upon being irradiated with the pulse laser beam, the target material may be turned into plasma, and EUV light may be emitted from the plasma. In order to stably supply the target material to the plasma generation region, the target material may be charged by applying a high voltage between the target material inside the target supply unit and an electrode provided so as to face a nozzle unit of the target supply unit, and the trajectory of the target material may be controlled by causing an electric field to act on the target material. However, when a high voltage exceeding a withstand voltage is applied between the target material and the electrode, a dielectric breakdown (spark discharge) may occur. When the dielectric breakdown occurs, leakage current may flow inside the chamber, and the voltage between the target material and the electrode may become unstable. As a result, a charge given to the target material may vary, and controlling the trajectory of the charged target material may become difficult. Accordingly, charged targets may not be stably supplied to the plasma generation region. According to one aspect of this disclosure, gas located in a space between an electrode and the nozzle unit, through which the target material is outputted, may be pumped out of the space. With the gas being pumped out, the withstand voltage across the space may be increased, whereby the dielectric breakdown may be suppressed. 2. Terms Terms used in this application may be interpreted as follows. “Debris” may include neutral particles, of the target material supplied into the chamber, that have not been turned into plasma and ion particles emitted from the plasma, and may be a substance that causes contamination or damage to an optical element. 3. Overview of EUV Light Generation System 3.1 Configuration FIG. 1 schematically illustrates the configuration of an exemplary LPP type EUV light generation system. An EUV light generation apparatus 1 may be used with at least one laser apparatus 3. Hereinafter, a system that includes the EUV light generation apparatus 1 and the laser apparatus 3 may be referred to as an EUV light generation system 11. As illustrated in FIG. 1 and described in detail below, the EUV light generation system 11 may include a chamber 2, a target supply unit 26, and so forth. The chamber 2 may be sealed airtight. The target supply unit 26 may be mounted to the chamber 2 so as to penetrate a wall of the chamber 2. A target material to be supplied by the target supply unit 26 may include, but is not limited to, tin, terbium, gadolinium, lithium, xenon, or any combination thereof. The chamber 2 may have at least one through-hole formed in its wall, and a pulse laser beam 32 may travel through the through-hole into the chamber 2. Alternatively, the chamber 2 may be provided with a window 21, through which the pulse laser beam 32 may travel into the chamber 2. An EUV collector mirror 23 having a spheroidal surface may, for example, be provided inside the chamber 2. The EUV collector mirror 23 may have a multi-layered reflective film formed on the spheroidal surface thereof. The reflective film may include a molybdenum layer and a silicon layer laminated alternately. The EUV collector mirror 23 may have a first focus and a second focus, and may be positioned such that the first focus lies in a plasma generation region 25 and the second focus lies in an intermediate focus (IF) region 292 defined by the specification of an external apparatus, such as an exposure apparatus 6. The EUV collector mirror 23 may have a through-hole 24 formed at the center thereof, and a pulse laser beam 33 may travel through the through-hole 24 toward the plasma generation region 25. The EUV light generation system 11 may further include an EUV light generation controller 5 and a target sensor 4. The target sensor 4 may have an imaging function and detect at least one of the presence, the trajectory, and the position of a target 27. Further, the EUV light generation system 11 may include a connection part 29 that allows the interior of the chamber 2 and the interior of the exposure apparatus 6 to be in communication with each other. A wall 291 having an aperture may be provided inside the connection part 29, and the wall 291 may be positioned such that the second focus of the EUV collector mirror 23 lies in the aperture formed in the wall 291. The EUV light generation system 11 may also include a laser beam direction control unit 34, a laser beam focusing mirror 22, and a target collector 28 for collecting targets 27. The laser beam direction control unit 34 may include an optical element for defining the direction into which the pulse laser beam 32 travels and an actuator for adjusting the position and the orientation (posture) of the optical element. 3.2 Operation With continued reference to FIG. 1, a pulse laser beam 31 outputted from the laser apparatus 3 may pass through the laser beam direction control unit 34 and be outputted therefrom as the pulse laser beam 32 after having its direction optionally adjusted. The pulse laser beam 32 may travel through the window 21 and enter the chamber 2. The pulse laser beam 32 may travel inside the chamber 2 along at least one beam path, be reflected by the laser beam focusing mirror 22, and strike at least one target 27 as a pulse laser beam 33. The target supply unit 26 may be configured to output the target(s) 27 toward the plasma generation region 25 inside the chamber 2. The target 27 may be irradiated with at least one pulse of the pulse laser beam 33. Upon being irradiated with the pulse laser beam 33, the target 27 may be turned into plasma, and rays of light including EUV light 251 may be emitted from the plasma. The EUV light 251 may be reflected selectively by the EUV collector mirror 23. EUV light 252 reflected by the EUV collector mirror 23 may travel through the intermediate focus region 292 and be outputted to the exposure apparatus 6. The target 27 may be irradiated with multiple pulses included in the pulse laser beam 33. The EUV light generation controller 5 may be configured to integrally control the EUV light generation system 11. The EUV light generation controller 5 may be configured to process image data of the target 27 captured by the target sensor 4. Further, the EUV light generation controller 5 may be configured to control at least one of the timing at which the target 27 is outputted and the direction into which the target 27 is outputted. Furthermore, the EUV light generation controller 5 may be configured to control at least one of the timing at which the laser apparatus 3 oscillates, the direction in which the pulse laser beam 32 travels, and the position at which the pulse laser beam 33 is focused. It will be appreciated that the various controls mentioned above are merely examples, and other controls may be added as necessary. 4. Chamber Including Electrostatic-Pull-Out Type Target Supply Unit 4.1 Configuration FIG. 2 is a partial sectional view illustrating the configuration of an EUV light generation apparatus according to a first embodiment. As shown in FIG. 2, a laser beam focusing optical system 22a, the EUV collector mirror 23, the target collection unit 28, and a beam dump 44 may be provided inside the chamber 2. The chamber 2 may include a member, such as an electrically conductive member, formed of an electrically conductive material, for example, a metal material. The chamber 2 may further include an electrically non-conductive member. In that case, the wall of the chamber 2 may be constituted by the electrically conductive member, and the electrically non-conductive member(s) may be provided inside the chamber 2. A plate 42 may be attached to the chamber 2, and a plate 43 may be attached to the plate 42. The EUV collector mirror 23 may be attached to the plate 42 through an EUV collector mirror mount 41. The laser beam focusing optical system 22a may include an off-axis paraboloidal mirror 221, a flat mirror 222, and holders for the respective mirrors 221 and 222. The off-axis paraboloidal mirror 221 and the flat mirror 222 may be mounted on the plate 43 through the respective mirror holders such that a laser beam reflected sequentially by these mirrors is focused in the plasma generation region 25. The beam dump 44 may be fixed to the chamber 2 through a beam dump support member 45 so as to be positioned on an extension of the beam path of the laser beam traveling toward the plasma generation region 25. The target collector 28 may be provided in the chamber 2 downstream from the plasma generation region 25 in the direction in which the target 27 travels. The chamber 2 may include the window 21 (laser beam port) and the target supply unit 26. The details of the target supply unit 26 will be given later. Electrically conductive metal or the like may be used as the target material. In the embodiments disclosed in this specification, tin (Sn), whose melting point is 232° C., may, for example, be used as the target material. Further, a gas supply device 46, a discharge device 47, and a pressure sensor 48 may be connected to the chamber 2. A beam steering unit 34a and the EUV light generation controller 5 may be provided outside the chamber 2. The beam steering unit 34a may include high-reflection mirrors 341 and 342, holders (not shown) for the respective mirrors 341 and 342, and a housing in which the mirrors 341 and 342 are disposed. The EUV light generation controller 5 may include an EUV light generation control device 51, a target control device 52, and a chamber pressure control device 56. The chamber pressure control device 56 may respectively be connected to the gas supply device 46, the discharge device 47, and the pressure sensor 48 through respective signal lines. 4.2 Operation A buffer gas and/or an etching gas may be introduced into the chamber 2. The buffer gas may be introduced to reduce the amount of debris, which is generated when the target material is irradiated with the laser beam, being deposited on the EUV collector mirror 23. The etching gas may be introduced to etch the debris deposited on the EUV collector mirror 23. Argon (Ar), neon (Ne), helium (He), or the like may be used as the buffer gas. Hydrogen (H2), hydrogen bromide (HBr), hydrogen chloride (HCl), or the like may be used as the etching gas. The gas supply device 46 may be configured to supply a hydrogen gas so as to flow along the reflective surface of the EUV collector mirror 23. With this configuration, tin (Sn) deposited on the surface of the EUV collector mirror 23 may be etched through a reaction expressed as follows:Sn (solid)+2H2 (gas)→SnH4 (gas) The discharge device 47 may be configured to discharge gas, such as hydrogen (H2) and tin hydride (SnH4) generated as tin is etched, from the chamber 2. The chamber pressure control device 56 may be configured to control the gas supply device 46 and the discharge device 47 based on a control signal from the EUV light generation control device 51 and a detection signal from the pressure sensor 48. By controlling the gas supply device 46 and the discharge device 47, the chamber pressure control device 56 may retain the gas pressure of the buffer gas and/or the etching gas inside the chamber 2 at predetermined pressure. The target supply unit 26 may be configured to charge the target material and supply the charged target material to the plasma generation region 25. A laser beam outputted from the laser apparatus 3 may be reflected sequentially by the high-reflection mirrors 341 and 342, and enter the laser beam focusing optical system 22a through the window 21. The laser beam that has entered the laser beam focusing optical system 22a may be reflected sequentially by the off-axis paraboloidal mirror 221 and the flat mirror 222. The EUV light generation control device 51 may be configured to output a target output signal to the target control device 52 and a laser beam output signal to the laser apparatus 3. Through these signals, the target material outputted from the target supply unit 26 may be irradiated with the laser beam at a timing at which the target material reaches the plasma generation region 25. Upon being irradiated with the laser beam, the target material may be turned into plasma, and EUV light may be emitted from the plasma. The emitted EUV light may be reflected by the EUV collector mirror 23, focused in the intermediate focus region 292, and outputted to an exposure apparatus. 5. Electrostatic-Pull-Out Type Target Supply Unit 5.1 Configuration FIG. 3 is a sectional view illustrating the target supply unit shown in FIG. 2 and the peripheral components. As shown in FIG. 3, the target supply unit 26 may include a reservoir 61, a nozzle unit (target output unit) 62, an electrode 63, a heater 64, an electrical insulator 65, a pull-out electrode 66, an aperture member 67, a discharge device 71, and a pressure sensor 72. The reservoir 61 and the nozzle unit 62 may be formed integrally or separately. The reservoir 61 may be formed of an electrically non-conductive material, such as synthetic quartz, alumina, or the like. The reservoir 61 may store tin serving as the target material. The heater 64 may be mounted around the reservoir 61 to heat the reservoir 61 so that tin inside the reservoir 61 is kept in a molten state. The heater 64 may be used with a temperature sensor (not shown) configured to detect the temperature of the reservoir 61, a heater power supply (not shown) configured to supply electric current to the heater 64, and a temperature controller (not shown) configured to control the heater power supply based on the temperature detected by the temperature sensor. The target material may be outputted toward the plasma generation region 25 through the nozzle unit 62. The nozzle unit 62 may have a through-hole (orifice) 62a formed therein, through which the target material is outputted. The through-hole 62a in the nozzle unit 62 may be in communication with the interior of the reservoir 61. The nozzle unit 62 may have a tip portion projecting from an outer surface so that an electric field is enhanced at the target material in the tip portion of the nozzle unit 62. The aforementioned through-hole (orifice) 62a may be formed in the tip portion of the nozzle unit 62. The cylindrical electrical insulator 65 may be attached to the nozzle unit 62. The pull-out electrode 66 may be held in the electrical insulator 65. The electrical insulator 65 may provide electrical insulation between the nozzle unit 62 and the pull-out electrode 66. The pull-out electrode 66 may be provided so as to face the outer surface of the nozzle unit 62 in order to cause an electric field to act therebetween. With this configuration, the target material may be pulled out through the orifice 62a in the nozzle unit 62. The pull-out electrode 66 may have a through-hole 66a formed therein to allow charged targets 27 to pass therethrough. The aperture member 67 may be provided downstream from the pull-out electrode 66 in a trajectory of the target 27, and may be fixed to the electrical insulator 65. The aperture member 67 may have a through-hole 67a formed therein to allow the target 27 to pass therethrough. A discharge port 65a may be formed in a portion of the electrical insulator 65, the portion being located between a part at which the electrical insulator 65 is connected to the nozzle unit 62 and a part at which the aperture member 67 is connected to the electrical insulator 65. The discharge port 65a may be connected to a discharge pipe 65b. The discharge pipe 65b may be connected to the discharge device 71 provided outside of the chamber 2. Further, the pressure sensor 72 may be connected to the electrical insulator 65 through a communication path 65c which is in communication with the interior of the electrical insulator 65. The pressure sensor 72 may be provided outside the chamber 2. The target supply unit 26 may further include a target pressure adjuster 53, an inert gas cylinder 54, and a voltage generator 55. The inert gas cylinder 54 may be connected to the target pressure adjuster 53 through a pipe for supplying the inert gas. The target pressure adjuster 53 may be in communication with the interior of the reservoir 61 through another pipe for supplying the inert gas. 5.2 Operation The reservoir 61 may be heated by the heater 64 to a temperature equal to or higher than 232° C. (melting point of tin). Being heated to the above temperature, the target material may be stored inside the reservoir 61 in a molten state. The target control device 52 may be configured to output a target generation signal to the voltage generator 55. In accordance with the target generation signal, the voltage generator 55 may apply a pulsed voltage between the electrode 63, which is in contact with the target material inside the reservoir 61, and the pull-out electrode 66. With the pulsed voltage being applied, Coulomb force may be generated between the target material and the pull-out electrode 66. As a result, the target material may be pulled out through the through-hole 62a in the nozzle unit 62, and a charged target 27 may be generated. The target pressure adjuster 53 may be configured to adjust the pressure of the inert gas supplied from the inert gas cylinder 54 as necessary, and pressurize the target material inside the reservoir 61. Being pressurized by the inert gas, the target material may project slightly from the tip portion of the nozzle unit 62. Then, the electric field may be enhanced at the target material projecting from the tip portion. As the electric field is enhanced at the target material, stronger Coulomb force may act between the target material and the pull-out electrode 66. The target control device 52 may be configured to control the target pressure adjuster 53 and the voltage generator 55 such that the target 27 is generated at a timing specified by the EUV light generation control device 51. Wiring connected to one of the output terminals of the voltage generator 55 may be connected to the electrode 63 through an airtight terminal, or feedthrough provided in the reservoir 61. Wiring connected to the other output terminal of the voltage generator 55 may be connected to the pull-out electrode 66 through a feedthrough provided in the chamber 2 and a through-hole formed in the electrical insulator 65. The voltage generator 55 may be configured to generate a pulsed voltage to cause the Coulomb force to act between the target material and the pull-out electrode 66 under the control of the target control device 52. For example, the voltage generator 55 may generate a voltage that varies in pulses between the reference potential (0 V) and a potential P1, which is higher than the reference potential. In this case, the reference potential may be applied to the pull-out electrode 66, and the potential P1 may be applied to the electrode 63. Alternatively, the voltage generator 55 may be configured to generate a voltage that varies in pulses between the potential P1 and a potential P2, which is higher than the potential P1. In this case, the potential P1 may be applied to the target material through the electrode 63, and the potential P2 may be applied to the pull-out electrode 66. With this configuration, a pulsed voltage may be applied between the target material and the pull-out electrode 66. Alternatively, when the nozzle unit 62 is formed of an electrically conductive material, such as metal, the voltage generator 55 may apply a pulsed voltage between the nozzle unit 62 and the pull-out electrode 66. The discharge device 71 may be configured to pump out gas located in a space inside the electrical insulator 65 through the discharge port 65a and the discharge pipe 65b. Pressure in the space inside the electrical insulator 65 may be measured by the pressure sensor 72, and obtained data may be inputted to the chamber pressure control device 56. The chamber pressure control device 56 may be configured to control the operation of the discharge device 71 based on the data inputted from the pressure sensor 72. 5.3 Effect In order to generate targets 27 with stable size, under stable timing and with stable charge amount, a voltage applied between the electrodes may be stable. However, when the buffer gas and/or the etching gas are/is present inside the chamber 2, a withstand voltage between the electrodes tends to be decreased, whereby a dielectric breakdown is more likely to occur. When the dielectric breakdown occurs, a predetermined voltage may not be applied between the electrodes, and at least one of the size, the output timing, and the charge amount of the targets 27 may become unstable. In other instances, the target 27 may not be outputted. A spark discharge may occur when an electron accelerated through an electric field collides with a gas molecule to ionize the gas. Accordingly, when the number of gas molecules decreases, the collision becomes less likely to occur. On the other hand, when the number of gas molecules increases, the gas molecules may not be accelerated sufficiently prior to the collision. Thus, the spark discharge may become less likely to occur in either case. However, when a large number of gas molecules are present in the chamber 2, transmittance of the EUV light may decrease, whereby the efficiency of the EUV light generation apparatus may be reduced. Thus, the dielectric breakdown may preferably be suppressed by pumping out gas located inside the chamber 2. There are cases where a buffer gas and/or an etching gas are/is supplied into the chamber 2, and the chamber 2 may not be kept under vacuum. Therefore, in the first embodiment, while the interior of the chamber 2 is kept at a predetermined gas pressure, gas in a space around the pull-out electrode 66 may be pumped out locally. As a result, the number of gas molecules in the aforementioned space may be reduced, whereby the dielectric breakdown may be suppressed. According to the first embodiment, the dielectric breakdown may be suppressed by pumping out gas that is located in the space inside the electrical insulator 65. As a result, the voltage applied between the target material and the pull-out electrode 66 may be stabilized, whereby the targets 27 may be supplied into the chamber 2 stably. 6. Target Supply Unit Including Acceleration Electrode 6.1 Configuration FIG. 4 is a sectional view illustrating a target supply unit according to a second embodiment and the peripheral components. In the second embodiment, a predetermined potential may be applied to the aperture member 67, whereby the aperture member 67 may serve as an acceleration electrode. The electrically conductive member (i.e., the wall) of the chamber 2 may be connected electrically to the reference potential of the voltage generator 55, or may be grounded. The aperture member 67 may be formed of an electrically conductive material, and may be connected electrically to the reference potential. The configuration pertaining to pumping out gas located in the space inside the electrical insulator 65 may be similar to that of the first embodiment. 6.2 Operation The voltage generator 55 may apply a predetermined potential P2, such as 10 kV, to the pull-out electrode 66. Further, the voltage generator 55 may, in the initial state, retain a potential applied to the target material at a potential P1. When the target material is to be pulled out, the voltage generator 55 may raise the potential applied to the target material from the potential P1 to another predetermined potential, for example, 20 kV. Through this operation, a positively charged target 27 may be pulled out through the nozzle unit 62. The target 27 may be pulled out toward the pull-out electrode 66, and may pass through the through-hole 66a formed in the pull-out electrode 66. Thereafter, the target 27 may be accelerated toward the aperture member 67, at which the reference potential is applied. In this way, the target 27 may be accelerated through a potential gradient formed along a path from the nozzle unit 62 to the aperture member 67 via the pull-out electrode 66, and may pass through the through-hole 67a in the aperture member 67. In the path of the target 27 that has passed through the through-hole 67a, the potential gradient may be gradual since the wall of the chamber 2 is connected to the reference potential. Accordingly, after passing through the through-hole 67a, the target 27 may travel inside the chamber 2 with a kinetic momentum at the time of passing through the through-hole 67a. 6.3 Effect According to the second embodiment, a voltage may be present between the aperture member 67 and the pull-out electrode 66, but a dielectric breakdown may be suppressed by pumping out gas located in the space inside the electrical insulator 65. Further, with this configuration, the speed of the target 27 may be controlled with precision. 7. Target Supply Unit Including Cover for Shielding Nozzle Unit 7.1 Configuration FIG. 5 is a sectional view illustrating a target supply unit according to a third embodiment and the peripheral components. In the third embodiment, a cover 81 may be attached on the inner surface of the wall of the chamber 2. The cover 81 may be provided so as to cover the leading end portion of the target supply unit 26 including at least the electrical insulator 65. The cover 81 may have a through-hole 81a formed therein to allow the targets 27 to pass therethrough. The cover 81 may be formed of an electrically conductive material, such as metal, and directly connected to the electrically conductive member (i.e., the wall) of the chamber 2. Alternatively, the cover 81 may be connected electrically to the wall of the chamber 2 through an electrically conductive connection member, such as a wire. The wall of the chamber 2 may be connected electrically to the reference potential of the voltage generator 55, or may be grounded. The cover 81 may cover a part of the reservoir 61, the nozzle unit 62, the electrical insulator 65, and the pull-out electrode 66 inside the chamber 2. Further, the cover 81 may preferably cover the aperture member 67, which may serve as an acceleration electrode, inside the chamber 2. The aperture member 67 may be connected electrically to the wall of the chamber 2 through a through-hole formed in the electrical insulator 65. The reservoir 61 may be mounted to the chamber 2 through a flange 84. The flange 84 may be formed of an electrically non-conductive material. A space defined by the cover 81 and the wall of the chamber 2, and optionally the flange 84, may be in communication with the discharge device 71 provided outside the chamber 2. 7.2 Operation and Effect The cover 81 may shield electrically non-conductive materials, such as the electrical insulator 65, from charged particles emitted from plasma generated in the plasma generation region 25. Gas in the space defined by the cover 81 and the wall of the chamber 2, and optionally the flange 84, may be pumped out by the discharge device 71. The gas being pumped out from the aforementioned space, occurrence of a dielectric breakdown around the pull-out electrode 66 and the aperture member 67 may be suppressed. Further, even when the reservoir 61 is formed of an electrically conductive material, occurrence of a dielectric breakdown between the reservoir 61 and the chamber 2 may be suppressed if the flange 84 is formed of an electrically non-conductive material. 8. Target Supply Unit Including Cover for Shielding Reservoir and Nozzle Unit 8.1 Configuration FIG. 6 is a sectional view illustrating a target supply unit according to a fourth embodiment and the peripheral components. In the fourth embodiment, a cover 85 may cover the reservoir 61, the nozzle unit 62, the electrical insulator 65, and the pull-out electrode 66. The cover 85 may further cover the aperture member 67, deflection electrodes 70 which will be described later, and a temperature sensor 73. As shown in FIG. 6, primary constituent elements of the target supply unit 26, such as the reservoir 61 and so forth, may be housed in a shielding container which includes the cover 85 and a lid 86 attached to the cover 85. The cover 85 may be mounted to the wall of the chamber 2. The cover 85 may have a through-hole 85a formed therein to allow the targets 27 to pass therethrough. The lid 86 may seal the opening in the cover 85 outside the chamber 2. The reservoir 61 may be mounted to the cover 85 through the lid 86. The cover 85 may be formed of an electrically conductive material, such as metal, and directly connected to the wall of the chamber 2. Alternatively, the cover 85 may be connected electrically to the wall of the chamber 2 through an electrically conductive connection member, such as a wire. The wall of the chamber 2 may be connected electrically to the reference potential of the voltage generator 55, or may be grounded. An electrically non-conductive material, such as mullite, may be used as a material for the lid 86. Multiple deflection electrodes 70 may be provided downstream from the aperture member 67 in the direction in which the target 27 travels. In the example shown in FIG. 6, two pairs of multiple deflection electrodes 70 are indicated. The deflection electrodes 70 may be held by the electrical insulator 65. The heater 64 may be mounted on the outer surface of the reservoir 61. The heater 64 may be used with the temperature sensor 73 configured to detect the temperature of the reservoir 61. A heater power supply 58 may be configured to supply electric current to the heater 64, and a temperature controller 59 may be configured to control the heater power supply 58 based on the temperature detected by the temperature sensor 73. Wiring of the pull-out electrode 66 and wiring of the deflection electrodes 70 may be connected to the voltage generator 55 and a deflection electrode voltage generator 57, respectively, through respective through-holes formed in the electrical insulator 65 and a relay terminal 90a provided in the lid 86. Wiring of the aperture member 67 may be connected electrically to the cover 85 through a through-hole formed in the electrical insulator 65, or may be connected to the voltage generator 55 through wiring (not shown) and the relay terminal 90a. Wiring of the electrode 63 may be connected to the voltage generator 55 through a relay terminal 90b provided in the lid 86. Wiring of the heater 64 and wiring of the temperature sensor 73 may be connected to the heater power supply 58 and the temperature controller 59, respectively, through a relay terminal 90c provided in the lid 86. A space inside the shielding container, which includes the cover 85 and the lid 86, and a space outside the reservoir 61, may be in communication with the discharge device 71 provided outside the chamber 2 through a connection port 71a. The electrical insulator 65 may have an opening 65d formed therein to facilitate pumping of gas in the space inside the electrical insulator 65. Although not shown in FIG. 6, an inert gas cylinder may be connected to the target pressure adjuster 53 through a pipe to supply an inert gas. 8.2 Operation and Effect As the electric current flows in the heater 64 from the heater power supply 58, the reservoir 61 and the target material inside the reservoir 61 may be heated. The temperature controller 59 may be configured to receive a control signal from the EUV light generation control device 51 and a detection signal from the temperature sensor 73, and control the electric current to be supplied from the heater power supply 58 to the heater 64. The temperature of the reservoir 61 may be controlled to a temperature equal to or higher than the melting point of tin so that tin serving as the target material is retained in a molten state. The target control device 52 may be configured to output a target generation signal to the voltage generator 55. Then, a charged target 27 may be pulled out through the nozzle unit 62, and passed through the through-hole in the pull-out electrode 66. The target 27 that has passed through the through-hole in the pull-out electrode 66 may be accelerated through an electric field between the pull-out electrode 66 and the aperture member 67, to which the reference potential is applied. Then, the target 27 may pass through the through-hole in the aperture member 67. The deflection electrodes 70 may cause an electric field to act on the target 27 that has passed through the through-hole in the aperture member 67 to thereby deflect the direction of the target 27. When the target 27 needs to be deflected, the target control device 52 may output a control signal to the deflection electrode voltage generator 57 to control a potential difference between each pair of the deflection electrodes 70. The deflection electrode voltage generator 57 may be configured to apply a voltage between each pair of the deflection electrodes 70. The target 27 may be deflected based on a control signal from the EUV light generation control device 51. Various signals may be transmitted between the EUV light generation control device 51 and the target control device 52. For example, the EUV light generation control device 51 may obtain information on the trajectory of the target 27 from a target sensor (not shown), and calculate a difference between the obtained trajectory and an ideal trajectory. Further, the EUV light generation control device 51 may be configured to send a signal to the target control device 52 to control a voltage applied between the deflection electrodes 70 so that the aforementioned difference becomes smaller. Here, the target 27 that has passed through the two pairs of the deflection electrodes 70 may pass through the through-hole 85a formed in the cover 85. The cover 85 may shield electrically non-conductive members, such as the electrical insulator 65, from charged particles emitted from plasma generated in the plasma generation region 25. Gas in the space defined by the cover 85 and the lid 86 may be pumped out by the discharge device 71. As a result of this configuration, occurrence of a dielectric breakdown around the pull-out electrode 66, the aperture member 67, and the deflection electrodes 70 may be suppressed. 9. Target Supply Unit in Which Position-Adjustable Cover 9.1 Configuration FIG. 7 is a sectional view illustrating a target supply unit according to a fifth embodiment and the peripheral components. In the fifth embodiment, the cover 85 may be mounted to the chamber 2 through an XY-moving stage 88. As shown in FIG. 7, a through-hole 2a may be formed in the wall of the chamber 2. The cover 85 that houses the reservoir 61, the nozzle unit 62, and the electrical insulator 65 may pass through the through-hole 2a. A portion of the cover 85 where the through-hole 85a is formed may be located inside the chamber 2. The connection port 71a that is in communication with the discharge device 71 and the lid 86 may be located outside the chamber 2. A flange 85b may be provided on the cover 85 at a portion between a part where the through-hole 85a is formed and a part where the connection port 71a or the lid 86 is located. The flange 85b may be connected to the wall of the chamber 2 through a flexible pipe 89 outside the chamber 2. More specifically, one end of the flexible pipe 89 may be fixed airtight to the wall of the chamber 2 around the through-hole 2a, and the other end of the flexible pipe 89 may be fixed airtight to the flange 85b. The flexible pipe 89 may be connected in the aforementioned manner between the wall of the chamber 2 and the flange 85b to seal the chamber 2 airtight. The flexible pipe 89 may be a bellows that may withstand the stress exerted by the difference in pressure inside and outside the chamber 2. In this way, the cover 85 and the wall of the chamber 2 may be connected so that the cover 85 may be movable relative to the chamber 2 while the chamber 2 is kept sealed airtight. The XY-moving stage 88 may be connected between the wall of the chamber 2 and the flange 85b outside the flexible pipe 89. Although not shown in FIG. 7, an inert gas cylinder may be connected to the target pressure adjuster 53 through a pipe to supply an inert gas. 9.2 Operation and Effect With the above configuration, the chamber 2 may be retained at low pressure, and the cover 85 may be held movably by the XY-moving stage 88. Further, gas in the space defined by the cover 85 and the lid 86 and outside the reservoir 61 may be pumped out by the discharge device 71. With this operation, occurrence of a dielectric breakdown around the pull-out electrode 66, the aperture member 67, and the deflection electrodes 70 may be suppressed. The above-described embodiments and the modifications thereof are merely examples for implementing this disclosure, and this disclosure is not limited thereto. Making various modifications according to the specifications or the like is within the scope of this disclosure, and other various embodiments are possible within the scope of this disclosure. For example, the modifications illustrated for particular ones of the embodiments can be applied to other embodiments as well (including the other embodiments described herein). The terms used in this specification and the appended claims should be interpreted as “non-limiting.” For example, the terms “include” and “be included” should be interpreted as “including the stated elements but not limited to the stated elements.” The term “have” should be interpreted as “having the stated elements but not limited to the stated elements.” Further, the modifier “one (a/an)” should be interpreted as at least one or “one or more.” |
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056051719 | abstract | An illumination source comprising a porous silicon having a source of electrons on the surface and/or interticies thereof having a total porosity in the range of from about 50 v/o to about 90 v/o. Also disclosed are a tritiated porous silicon and a photovoltaic device and an illumination source of tritiated porous silicon. |
claims | 1. A charged particle lithography system for transferring a pattern onto the surface of a target, comprising:a beam generator for generating a plurality of charged particle beamlets, the plurality of beamlets defining a column;a plurality of aperture array elements comprising a first aperture array, a blanker array, a beam stop array, and a projection lens array;wherein each aperture array element comprises a plurality of apertures arranged in a plurality of groups, the apertures for letting the beamlets pass through the aperture array element;wherein the groups of apertures of each aperture array element form beam areas distinct and separate from a plurality of non-beam areas formed between the beam areas and containing no apertures for passage of the beamlets;wherein the beam areas of the aperture array elements are aligned to form beam shafts, each comprising a plurality of beamlets, and the non-beam areas of the aperture array elements are aligned to form non-beam shafts not having beamlets present therein; andwherein the first aperture array element is provided with cooling channels adapted for transmission of a cooling medium for cooling the first aperture array element, the cooling channels being provided in the non-beam areas of the first aperture array element. 2. The system of claim 1, wherein the first aperture array element comprises a plate having a thickness in a direction of the axis of the column and a width in a direction perpendicular to the axis of the column, wherein the apertures are formed through the thickness of the plate in the non-beam areas of the plate, and the cooling channels are formed internally in the non-beam areas of the plate and extend in a direction of the width of the plate. 3. The system of claim 1, wherein the first aperture array element comprises a plate having a thickness in a direction of the axis of the column and a width in a direction perpendicular to the axis of the column, wherein the apertures are formed through the thickness of the plate in the non-beam areas of the plate, and the cooling channels are formed in external elements attached to the plate in the non-beam areas and extending in a direction of the width of the plate, the cooling channels adapted for providing structural support for the first aperture array element. 4. The system of claim 1, wherein the cooling medium comprises water. 5. The system of claim 1, further comprising a coolant system for flowing the cooling medium through the cooling channels, the coolant system being adapted to produce turbulent flow of the cooling medium through the cooling channels. 6. The system of claim 1, wherein the first aperture array is made from a monolithic plate of material in which the apertures and cooling channels are formed. 7. The system of claim 1, wherein the first aperture array is made from a plate of Tungsten. 8. The system of claim 1, wherein the first aperture array is made from a plate of Copper or Molybdenum. 9. The system of claim 1, wherein the plurality of aperture array elements further comprise a current limiting aperture array and a condenser lens array, each comprising a plurality of apertures arranged in a plurality of groups, the apertures for letting the beamlets pass through the aperture array elements, and wherein the groups of apertures of each aperture array element form beam areas distinct and separate from a plurality of non-beam areas formed between the beam areas and containing no apertures for passage of the beamlets, and wherein the beam areas of the aperture array elements are aligned to form beam shafts, each comprising a plurality of beamlets, and the non-beam areas of the aperture array elements are aligned to form non-beam shafts not having beamlets present therein. 10. The system of claim 1, wherein the first aperture array element comprises an integral current limiting aperture array, the apertures of the first aperture array element having a narrowest portion recessed below the upper surface of the first aperture array element. 11. The system of claim 1, wherein the first aperture array element is provided with a curved upper surface facing towards the beam generator. 12. The system of claim 11, wherein the first aperture array element is subdivided into alternating aperture-free areas and aperture areas, each aperture area comprising a plurality of apertures, and wherein the curved upper surface encompasses a plurality of the aperture-free areas and aperture areas. 13. The system of claim 11, wherein the curved upper surface of the first aperture array element forms a raised dome-shaped area protruding above the upper surface towards the beam generator. 14. The system of claim 11, wherein the curved upper surface of the first aperture array element forms a dome-shaped depression in the upper surface area facing the beam generator. 15. The system of claim 11, wherein the system has an optical axis and the curved surface is shaped according to a cosine function centred around the optical axis. 16. The system of claim 11, wherein the circumference of the curved surface is substantially larger than the height of the curved surface. 17. An aperture array element adapted for use in a charged particle lithography system for generating a plurality of beamlets for transferring a pattern onto the surface of a target, the aperture array comprising a plurality of apertures arranged in a plurality of groups, the apertures for letting the beamlets pass through the aperture array element;wherein the groups of apertures form beam areas distinct and separate from a plurality of non-beam areas formed between the beam areas and containing no apertures for passage of the beamlets; andwherein the first aperture array element is provided with cooling channels adapted for transmission of a cooling medium for cooling the first aperture array element, the cooling channels being provided in the non-beam areas of the first aperture array element. 18. The aperture array element of claim 17, wherein the first aperture array element comprises a plate having a thickness and a width, wherein the apertures are formed through the thickness of the plate in the non-beam areas of the plate, and the cooling channels are formed internally in the non-beam areas of the plate and extend in a direction of the width of the plate. 19. The aperture array element of claim 17, wherein the first aperture array element comprises a plate having a thickness and a width, wherein the apertures are formed through the thickness of the plate in the non-beam areas of the plate, and the cooling channels are formed in external elements attached to the plate in the non-beam areas and extending in a direction of the width of the plate, the cooling channels adapted for providing structural support for the first aperture array element. 20. The aperture array element of claim 17, wherein the cooling medium comprises water. 21. The aperture array element of claim 17, further comprising a coolant system for flowing the cooling medium through the cooling channels, the coolant system being adapted to produce turbulent flow of the cooling medium through the cooling channels. 22. The aperture array element of claim 17, wherein the first aperture array is made from a monolithic plate of material in which the apertures and cooling channels are formed. 23. The aperture array element of claim 17, wherein the first aperture array is made from a plate of Tungsten. 24. The aperture array element of claim 17, wherein the first aperture array is made from a plate of Copper or Molybdenum. 25. The aperture array element of claim 17, wherein the first aperture array element comprises an integral current limiting aperture array, the apertures of the first aperture array element having a narrowest portion recessed below the upper surface of the first aperture array element. 26. The aperture array element of claim 17, wherein the aperture array element comprises a plate, the plate being provided with a curved upper surface facing towards a beam direction. 27. The aperture array element of claim 26, wherein the aperture array element is subdivided into alternating aperture-free areas and aperture areas, each aperture area comprising a plurality of apertures, and wherein the curved upper surface encompasses a plurality of the aperture-free areas and aperture areas. 28. The aperture array element of claim 26, wherein the curved upper surface of the aperture array forms a raised dome-shaped area protruding above the upper surface towards the charged particle source. 29. The aperture array element of claim 26, wherein the curved upper surface of the aperture array forms a dome-shaped depression in the upper surface area facing the charged particle source. 30. The aperture array element of claim 26, wherein the system has an optical axis and the curved surface is shaped according to a cosine function centred around the optical axis. 31. The aperture array element of claim 26, wherein the circumference of the curved surface is substantially larger than the height of the curved surface. 32. A charged particle beam generator, comprising:a charged particle source adapted for generating a diverging charged particle beam;a collimating system for refracting the diverging charged particle beam, the collimating system comprising a first electrode; andan aperture array element according to claim 17, the aperture array element forming a second electrode;wherein the system is adapted for creating an accelerating electric field between the first electrode and the second electrode. |
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041785240 | summary | BACKGROUND OF THE INVENTION This invention relates to electrical power sources and more particularly to a radioisotope photoelectric generator which may be used as a self-contained remote power source. Heretofore, various types of electrical generators and batteries have been used for various uses. Remote power sources include such diverse sources as conventional batteries, radioisotope thermoelectric generators (RTG), and solar cell arrays. Batteries have the obvious disadvantage of limited amounts of energy available and thus they must be recharged periodically. Solar cell arrays have many significant advantages (particularly for spacecraft) but even when used with storage cells they cannot supply reliable power on earth because of cloud cover and similar problems. A highly reliable system for remote applications such as the recent space missions to the outer planets or for remote military requirements is the RTG. The RTG is the most similar power source to the present invention. Both use, as their energy sources, radioactive isotopes and therefore have the potential for very high reliability. The RTG, however, has the disadvantages of low efficiency (.about.4%) and very high operating temperature required and, for applications where a high voltage source is required, a further conversion from the low voltages produced by the RTG (generally less than 100 volts) to the desired high voltage is required at a further sacrifice in efficiency. As used in this disclosure "high Z" or "high atomic number" is defined as Z>46 and "low Z" or "low atomic number" is defined as Z<23. SUMMARY OF THE INVENTION This device is a remote electrical generator which makes use of radioactive isotopes in combination with high-Z and low-Z materials to generate an electrical output voltage. X rays or gamma rays emitted by the radioactive isotopes produce electrons in the high Z material which are captured by the low-Z material to produce the output energy. The high-Z and low-Z materials alternate and are connected electrically together to form a "battery." The generator may be used as a high-voltage source or as a remote power source for space vehicles, unattended facilities such as buoys and weather stations, and for military and commercial uses. The device does not depend upon any outside sources; it has long life; it has no moving parts; and it has high reliability and trouble-free operation. |
047175311 | claims | 1. An article transport system, comprising: means for moving said article from a first location toward a second location substantially within a first plane; means for driving said moving means within said first plane; and pivotable means fixedly disposed within the vicinity of said second location for automatically moving said article from said first plane into a second plane in response to said driven movement of said moving means within said first plane. means upon said article for engaging said pivotable means. said moving means comprises a railroad transport car; and said article is a nuclear reactor fuel container pivotably mounted upon said transport car. said first plane is a horizontal plane; and said second plane is a vertical plane. the center of gravity of said fuel container is located relative to the pivotable axis of said fuel container so as to bias said fuel container toward said horizontal plane. means for moving said article between a first location substantially with a first plane; means for driving said moving means within said first plane; pivotable means fixedly disposed within the vicinity of said first location for automatically moving said article from said first plane into a second plane in response to said driven movement of said moving means within said first plane when said article is being transported from said second location to said first location; and pivotable means fixedly disposed within the vicinity of said second location for automatically moving said article from said first plane into a second plane in response to said driven movement of said moving means within said first plane when said article is being transported from said first location to said second location. said first plane is a horizontal plane; and said second plane is a vertical plane. means for moving said fuel container between said reactor containment handling pool and said spent fuel storage pool substantially within a first plane; means for driving said moving means within said first plane; pivotable first means fixedly disposed within said reactor containment handling pool for automatically moving said container from said first plane into a second plane in response to said driven movement of said moving means within said first plane when said fuel container is being transported from said spent fuel storage pool into said reactor container handling pool; and pivotable second means fixedly disposed within said spent fuel storage pool for automatically moving said container from said first plane into a second plane in response to said driven movement of said moving means within said first plane when said fuel container is being transported from said reactor containment handling pool into said spent fuel storage pool. said first plane is a horizontal plane within which said transport tube is disposed; said second plane is a vertical plane within which said fuel container is disposed for unloading spent fuel and loading fresh fuel when said fuel container is located within said spent fuel storage pool; and said second plane is a vertical plane within which said fuel container is disposed for unloading fresh fuel and loading spent fuel when said fuel container is located within said reactor containment handling pool. stop means disposed within said reactor containment handling and spent fuel storage pools for limiting the movement of said fuel container from said first plane to said second plane. means for moving said article from a first location toward a second location substantially within a first plane; means for driving said moving means within said first plane; pivotable bar means fixedly disposed within the vicinity of said second location for automatically moving said article from said first plane to a second plane in response to said driven movement of said moving means within said first plane; and slotted bracket means upon said article for engaging said pivotable bar means. said pivotable means is dependently supported at an elevational level disposed above said first plane of movement of said article; and said article engaging brackets are secured upon an upper surface portion of said article. said pivotable means is upstandingly supported at an elevational level disposed below said first plane of movement of said article; and said article engaging brackets are secured upon an undersurface portion of said article. means for biasing said pivotable means toward its initial position for engagement with said bracket means. said biasing means comprises counterweight means. said pivotable means comprises an arm, and said bar means comprises a pair of bars projecting laterally outwardly from said arm; and said bracket means, comprises a pair of laterally spaced slotted brackets for engaging said pair of bars. said pivotable means comprises a pair of laterally spaced arms, and said bar means comprises a laterally spaced pair of bars respectively mounted upon said pair of arms and projecting laterally inwardly toward each other; and said bracket means comprises a pair of laterally spaced slotted brackets for engaging said pair of bars. said brackets are disposed upon side surfaces of said article. stop means for defining said initial position of said pivotable means for engagement with said bracket means. 2. An article transport system as set forth in claim 1, further comprising: 3. An article transport system as set forth in claim 1, wherein: 4. An article transport system as set forth in claim 3, wherein: 5. An article transport system as set forth in claim 4, wherein; 6. An article transport system, comprising: 7. An article transport system as set forth in claim 6 wherein: 8. A transport system for use within a nuclear facility for transferring fuel containers through a transfer tube or conduit between a reactor containment handling pool and a spent fuel storage pool, comprising: 9. A transport system as set forth in claim 8, wherein: 10. A transport system as set forth in claim 8, further comprising: 11. An article transport system, comprising: 12. An article transport system as set forth in claim 11, wherein: 13. An article transport system as set forth in claim 11, wherein: 14. An article transport system as set forth in claim 11, further comprising: 15. An article transport system as set forth in claim 14, wherein: 16. An article transport system as set forth in claim 11, wherein: 17. An article transport system as set forth in claim 11, wherein: 18. An article transport system as set forth in claim 17, wherein: 19. An article transport system as set forth in claim 14, further comprising: |
claims | 1. A circumferential sampling tool for obtaining a sample from an interior wall of a tube comprising:a cylindrical body having a central axis;an aperture in the cylindrical body;a shaft disposed in the cylindrical body along the central axis;a first cutter operatively connected to the shaft for rotation therewith, the first cutter being movable radially between a retracted position where the first cutter is disposed inside the cylindrical body at a first distance from the central axis and an extended position where the first cutter extends at least in part through the aperture at a second distance from the central axis, the second distance being greater than the first distance;a first actuator operatively connected to the first cutter for moving the first cutter between the retracted position and the extended position as the shaft rotates, the first actuator mechanically biasing the first cutter toward the retracted position;a second cutter operatively connected to the shaft for rotation therewith and being disposed at an angle to first cutter, the second cutter being movable radially between a retracted position where the second cutter is disposed inside the cylindrical body at a third distance from the central axis and an extended position where the second cutter extends at least in part through the aperture at a fourth distance from the central axis, the fourth distance being greater than the third distance, the fourth distance being greater than the second distance; anda second actuator operatively connected to the second cutter for moving the second cutter between the retracted position and the extended position as the shaft rotates, the second actuator mechanically biasing the second cutter toward the retracted position, the second cutter being in the retracted position when the first cutter is in the extended position, and the first cutter being in the retracted position when the second cutter is in the extended position;wherein rotating the shaft causes the first cutter to move to the extended position thereby cutting a portion of the interior wall of the tube and then causes the second cutter to move to the extended position thereby cutting the sample from the interior wall of the tube from a location in the tube revealed by cutting the portion of the interior wall of the tube. 2. The tool of claim 1, wherein:the first actuator comprises a spring mechanically biasing the first cutter toward the retracted position; andthe second actuator comprises a spring mechanically biasing the second cutter toward the retracted position. 3. The tool of claim 2, further comprising a ramp disposed inside the cylindrical body along a circumferential portion thereof, the ramp being disposed opposite the aperture;wherein the first actuator further comprises a first roller, the first roller causing the first cutter to move to the extended position when the first roller rolls over the ramp; andwherein the second actuator further comprises a second roller, the second roller causing the second cutter to move to the extended position when the second roller rolls over the ramp. 4. The tool of claim 3, wherein a diameter of the first roller is greater than a diameter of the second roller. 5. The tool of claim 1, wherein the first cutter is wider than the second cutter. 6. The tool of claim 1, wherein an arc defined by the first cutter in the extended position as the shaft rotates is longer than an arc defined by the second cutter in the extended position as the shaft rotates. 7. The tool of claim 1, further comprising:a first receptacle connected to the first cutter for receiving the portion of the interior wall of the tube cut by the first cutter; anda second receptacle connected to the second cutter for receiving the sample cut by the second cutter. 8. The tool of claim 1, further comprising:at least one spring connected to the first cutter for biasing the first cutter against the interior wall of the tube when the first cutter is in the extended position; andat least one spring connected to the second cutter for biasing the second cutter against the interior wall of the tube when the second cutter is in the extended position. 9. The tool of claim 1, wherein the first cutter is disposed opposite the second cutter. 10. The tool of claim 9, further comprising at least one spring connected between the first cutter and the second cutter, the at least one spring biasing the first and second cutters away from each other. 11. The tool of claim 1, further comprising a motor disposed in the cylindrical body and operatively connected to the shaft for rotating the shaft. 12. A circumferential sampling tool for obtaining a sample from an interior wall of a tube comprising:a cylindrical body having a central axis;an aperture in the cylindrical body;a shaft disposed in the cylindrical body along the central axis;an extension ramp connected to the cylindrical body;a retraction ramp connected to the cylindrical body;a first cutter operatively connected to the shaft for rotation therewith, the first cutter being movable radially between a retracted position where the first cutter is disposed inside the cylindrical body at a first distance from the central axis and an extended position where the first cutter extends at least in part through the aperture at a second distance from the central axis, the second distance being greater than the first distance;a first actuator operatively connected to the first cutter for moving the first cutter between the retracted position and the extended position by interacting with the retraction ramp and the extension ramp respectively as the shaft rotates;a second cutter operatively connected to the shaft for rotation therewith and being disposed at an angle to first cutter, the second cutter being movable radially between a retracted position where the second cutter is disposed inside the cylindrical body at a third distance from the central axis and an extended position where the second cutter extends at least in part through the aperture at a fourth distance from the central axis, the fourth distance being greater than the third distance, the fourth distance being greater than the second distance; anda second actuator operatively connected to the second cutter for moving the second cutter between the retracted position and the extended position by interacting with the retraction ramp and the extension ramp respectively as the shaft rotates, the second cutter being in the retracted position when the first cutter is in the extended position, and the first cutter being in the retracted position when the second cutter is in the extended position;wherein rotating the shaft causes the first cutter to move to the extended position thereby cutting a portion of the interior wall of the tube and then causes the second cutter to move to the extended position thereby cutting the sample from the interior wall of the tube from a location in the tube revealed by cutting the portion of the interior wall of the tube. 13. The tool of claim 12, wherein the first actuator includes a first actuation bar disposed generally parallel to the central axis, the first actuation bar having a first roller at a first end thereof, a second roller at a second end thereof, and at least one third roller between the first and second ends thereof;wherein the second actuator includes a second actuation bar disposed generally parallel to the central axis, the second actuation bar having a fourth roller at a first end thereof, a fifth roller at a second end thereof, and at least one sixth roller between the first and second ends thereof;wherein the extension ramp extends generally parallel to the central axis toward the first cutter and the second cutter, and defines an arc about the central axis;wherein the retraction ramp extends generally parallel to the central axis toward the extension ramp, the first cutter and the second cutter, and defines an arc about the central axis;wherein the first and second cutters are disposed between the extension ramp and the retraction ramp in a direction parallel to the central axis;the tool further comprising:a first holder connected to the first cutter, the first holder having at least one slot defined therein at an angle to the central axis, the at least one slot of the first holder receiving the at least one third roller therein; anda second holder connected to the second cutter, the second holder having at least one slot defined therein at an angle to the central axis, the at least one slot of the second holder receiving the at least one sixth roller therein;wherein when the first roller rolls over the extension ramp, the at least one third roller moves in the at least one slot of the first holder causing the first holder to move radially away from the central axis thereby causing the first cutter to move to the extended position;wherein when the second roller rolls over the retraction ramp, the at least one third roller moves in the at least one slot of the first holder causing the first holder to move radially toward the central axis thereby causing the first cutter to move to the retracted position;wherein when the fourth roller rolls over the extension ramp, the at least one sixth roller moves in the at least one slot of the second holder causing the second holder to move radially away from the central axis thereby causing the second cutter to move to the extended position; andwherein when the fifth roller rolls over the retraction ramp, the at least one sixth roller moves in the at least one slot of the second holder causing the second holder to move radially toward the central axis thereby causing the second cutter to move to the retracted position. 14. The tool of claim 13, wherein the extension ramp has a first ramp portion and a second ramp portion, the first ramp portion being longer than the second ramp portion; andwherein the first roller rolls over the first ramp portion of the extension ramp and the fourth roller rolls over the second ramp portion of the extension ramp. 15. The tool of claim 13, wherein the retraction ramp has a first ramp portion and a second ramp portion, the first ramp portion being longer than the second ramp portion; andwherein the second roller rolls over the second ramp portion of the retraction ramp and the fifth roller rolls over the first ramp portion of the retraction ramp. 16. The tool of claim 12, wherein the first cutter is wider than the second cutter. 17. The tool of claim 12, wherein an arc defined by the first cutter in the extended position as the shaft rotates is longer than an arc defined by the second cutter in the extended position as the shaft rotates. 18. The tool of claim 12, further comprising:a first receptacle connected to the first cutter for receiving the portion of the interior wall of the tube cut by the first cutter; anda second receptacle connected to the second cutter for receiving the sample cut by the second cutter. 19. The tool of claim 12, further comprising:at least one spring connected to the first cutter for biasing the first cutter against the interior wall of the tube when the first cutter is in the extended position; andat least one spring connected to the second cutter for biasing the second cutter against the interior wall of the tube when the second cutter is in the extended position. 20. The tool of claim 12, further comprising a motor disposed in the cylindrical body and operatively connected to the shaft for rotating the shaft. |
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claims | 1. Soft X-ray microscope, comprising:a table (10);a housing (20) installed to the upper side of the table (10) and having a partition (22);a light source chamber (30) installed lower than the partition (22) of the housing (20) to project a light to liquid jetted under a high pressure to generate plasma;a mirror chamber (40), installed above the partition (22) of the housing (20), in which first and second mirror (410 and 430) are respectively installed to upper and lower sides of a holder (420) for storing a living sample, the soft X-ray generated by the plasma generated in the light source chamber (30) illuminates the living sample, and the soft X-ray penetrated the living sample is amplified to obtain an image in an image capturing chamber; andan image capturing chamber (50) installed to the upper side of the housing (20) to amplify a light image signal amplified through the mirror chamber (40) and to capture the light image on an external screen to allow distinguishing the light image from exterior;wherein the mirror chamber (40) comprises:a first base plate (440) fixed to the upper side of the partition (22) of the housing (20) and having a first transmission hole (442) formed in the central portion thereof;a first mirror (410) including a first transporting device (412) installed on the first base plate (440), and a condenser mirror (414) installed in the central portion of the first transporting device (412) to amplify the light and to illuminate the living sample;a second base plate (450) positioned above the first mirror (410), supported by a plurality of supporting rods (452) to maintain the distance from the first base plate (440), and having a second transmission hole (454) formed in the central portion thereof;a holder part (420) including a second transporting device (422) installed on the second base plate (450), and a coupling (426) for separating and coupling the holder (424) storing the living sample from and to the central portion of the second transporting device (422);a second mirror (430) including a third transporting device (432) installed on the second base plate (450), and a Fresnel diffraction zone plate (434) installed in the central portion of the third transporting device (432) and positioned above the holder (424); anda vacuuming device (460) for vacuuming the inside of the housing (20) having the mirror chamber (40) and for maintaining vacuum. 2. The soft X-ray microscope as claimed in claim 1, further comprising a telemicroscope (60) installed to the side of the light source chamber (30) to allow watching the procedure of projecting the soft X-ray to the high-pressure liquid to form the plasma from the exterior. 3. The soft X-ray microscope as claimed in claim 2, wherein, the light source chamber (30) comprises:a nozzle part (310) for jetting liquid nitrogen supplied from the exterior under a high pressure;a discharge part (320) provided opposite to the nozzle part (310) to suction the liquid nitrogen and to discharge the liquid nitrogen to the exterior;a light source (330) for projecting a light to the liquid nitrogen jetted from the nozzle part (310) to form the plasma; anda light source vacuum pump (340) for vacuuming the inside of the housing (20) in which the light source (30) is installed and for maintaining vacuum of the housing (20). 4. The soft X-ray microscope as claimed in claim 1, wherein, the light source chamber (30) comprises:a nozzle part (310) for jetting liquid nitrogen supplied from the exterior under a high pressure;a discharge part (320) provided opposite to the nozzle part (310) to suction the liquid nitrogen and to discharge the liquid nitrogen to the exterior;a light source (330) for projecting a light to the liquid nitrogen jetted from the nozzle part (310) to form the plasma; anda light source vacuum pump (340) for vacuuming the inside of the housing (20) in which the light source (30) is installed and for maintaining vacuum of the housing (20). 5. The soft X-ray microscope as claimed in claim 4, wherein the nozzle part (310) comprises:a capillary tube (312) for receiving the high-pressure nitrogen gas from the exterior to jet the high-pressure nitrogen gas; andan outer tube (314) for surrounding the outer circumference of the capillary tube (312) and for receiving the high-pressure liquid nitrogen from the exterior to be filled up and to liquefy the high-pressure nitrogen gas jet through the capillary tube (312). 6. The soft X-ray microscope as claimed in claim 4, wherein the light source (330) comprises a diode pump solid laser having an average power of 12 W and a repetition rate of 300 Hz. 7. The soft X-ray microscope as claimed in claim 4, wherein the light source vacuum pump (340) comprises a turbo molecular pump having a vacuum degree of more than 500 L/S. 8. The soft X-ray microscope as claimed in claim 1, further comprising a rod lock chamber (70) provided at the side of the mirror chamber (40) and to transport the holder (424) such that vacuum of the mirror chamber (40) is not damaged and the holder (424) storing the living sample is coupled with and separated from the coupling (426) of the holder part (420). 9. The soft X-ray microscope as claimed in claim 8, wherein the rod lock chamber (70) comprises a vacuuming device (72) for preventing vacuum generated in the mirror chamber (40) from being damaged when the holder (424) storing the living sample is separated from and coupled with the mirror chamber (40), wherein the vacuuming device (72) comprises a turbo molecular pump of 60 L/S and an ion pump of 30 L/S. 10. The soft X-ray microscope as claimed in claim 1, further comprising an optical aligning device (80) for checking whether the first mirror (410), the holder part (420), and the second mirror (430) are aligned in the optical axis direction, and for aligning the same. 11. The soft X-ray microscope as claimed in claim 1, wherein the condenser mirror (414) includes first and second oval-shaped hedrons (414a and 414b) symmetrical to each other and having an optical axis-directional length 136 mm, an inner diameter of 50 mm, and a depth of 42 mm, and a pin hole (414) formed in the center portion in the longitudinal direction; andthe first and second oval-shaped hedrons (414a and 414b) are formed by ovals having a longitudinal directional center as a focal point (P), a distance of 160 mm from the focal point (P) to another focal points (P′ and P″), and symmetrical to each other with respect to the central focal point (P). 12. The soft X-ray microscope as claimed in claim 1, wherein the Fresnel diffraction zone plate (434) is manufactured by forming gold (Au) having a thickness of 100 nm to 160 nm on a silicon nitride layer (Si3N4) substrate, and has an outmost zone width of 30 mm to 40 mm a diameter of 60 mm to 70 mm, and the number of Fresnel diffraction zone plate is 200 to 300. 13. The soft X-ray microscope as claimed in claim 1, wherein the vacuuming device (460) comprises at least one turbo molecular pump of 210 L/S and at least one ion pump of 120 L/S. 14. The soft X-ray microscope as claimed in claim 1, further comprising a filter (470) installed in the lower side of the first base plate (440) to filter the light transmitted to the mirror chamber (40) through the plasma generated by the light source chamber (30) and to separate the vacuum of the light source chamber (30) and the mirror chamber (40), and made of titanium. 15. The soft X-ray microscope as claimed in claim 1, further comprising a shielding device (480) installed to the lower side of the second base plate (450) to interrupt a direct light, which is not amplified by the condenser mirror (414), to directly illuminate the living sample when illuminating the illuminated through the condenser mirror (414), and including a through-hole (484a) formed in a supporting plate (484) supported by a fourth transporting device (482), and a focal point interrupting plate (486) installed in the center of the through-hole (484a) to interrupt the direct light. 16. The soft X-ray microscope as claimed in claim 1, wherein the image capturing chamber (50) comprises:a multi-channel plate (510) for converting a light image signal obtained through the light amplified by the second mirror (430) into an electric signal; anda CCD (520) for amplifying the electric signal converted by the multi-channel plate (510) and for converting the amplified electric signal into a visible light using a fluorophor such that the converted visible light forms an image on the external screen through an optical lens. 17. A soft X-ray microscope, comprising:a table (10);a housing (20) installed to the upper side of the table (10) and having a partition (22);a light source chamber (30) installed lower than the partition (22) of the housing (20) to project a light to liquid jelled under a high pressure to generate plasma;a mirror chamber (40), installed above the partition (22) of the housing (20), in which first and second mirror (410 and 430) are respectively installed to upper and lower sides of a holder (420) for storing a living sample, the soft X-ray generated by the plasma generated in the light source chamber (30) illuminates the living sample, and the soft X-ray penetrated the living sample is amplified to obtain an image in an image capturing chamber; andan image capturing chamber (50) installed to the upper side of the housing (20) to amplify a light image signal amplified through the mirror chamber (40) and to capture the light image on an external screen to allow distinguishing the light image from exterior;wherein the holder part (420) comprises:a holder (424) including:a sample part (4210) having sample windows (4212) made of a silicon nitride layer (Si3N4) with a thickness of 90 nm to 120 nm to cover ends of the living sample and viton plates (4214) for covering ends of the sample windows (4212);a sample plate (4220), on which the sample part (4210) is placed, haying a transmission hole (4222) formed in the center and a locking hook (4224) formed in a side;a cover plate (4230) for covering the upper side of the sample plate (4220) on which the sample part (4210 ) is placed and having a transmission hole (4232) formed in the center thereof; andan O-ring (4240) for maintaining sealing between the sample plate (4220) and the cover plate (4230);a coupling (426) including a plurality of supporting plates (426a) having a plurality of ball plungers (426b) to support outer circumference of the sample plate (4220), and an opened portion enabling the holder (424) to separate; anda second transporting device (422) provided at the side of the coupling (426) and transported in the three directions of the X-axis, the Y-axis, and the Z-axis by a motor. 18. A soft X-ray microscope, comprising:a table;a housing installed at an upper side of the table, and having a partition, the partition separating the housing into a mirror chamber disposed above the partition and a light source chamber disposed below the partition, the light source chamber projecting light to a liquid jetted under a high pressure to thereby generate plasma, a soft X-ray being generated from the plasma in the light source chamber, the mirror chamber having:a first base plate fixed to an upper side of the partition and having a first transmission hole formed in a central portion thereof;a first mirror including a first transporting device installed on the first base plate, and a condenser mirror installed at a central portion of the first transporting device, the condenser mirror amplifying the soft X-ray and illuminating and penetrating a living sample, to obtain an optical image signal;a second base plate positioned above the first mirror, and being supported by a plurality of supporting rods to separate the second base plate from the first base plate, and having a second transmission hole formed in a central portion thereof;a holder part for storing the living sample, the holder part having a second transporting device installed on the second base plate, and a coupling that respectively separates and couples a holder from and to the central portion of the second transporting device; anda second mirror disposed at an upper side of the holder, and including a third transporting device installed on the second base plate, and a Fresnel diffraction zone plate installed at a central portion of the third transporting device and positioned above the holder; anda vacuuming device for generating and maintaining a vacuum inside the mirror chamber; andan image capturing chamber installed at an upper side of the housing to amplify the optical image signal and to capture the optical image signal on an external screen to allow an image of the living sample to be viewed. 19. A soft X-ray microscope, comprising:a table;a housing installed at an upper side of the table, and having a partition, the partition separating the housing into a mirror chamber disposed above the partition and a light source chamber disposed below the partition, the light source chamber projecting light to a liquid jetted under a high pressure to thereby generate plasma, the mirror chamber having:a holder part for storing a living sample, the holder part including:a holder including:a sample part having a plurality of sample windows, each being made of a silicon nitride layer (Si3N4) with a thickness of 90 nm to 120 nm, to cover the living sample, and a plurality of viton plates for covering the sample windows;a sample plate, having an upper side on which the sample part is placed, a transmission hole formed in a center thereof, and a locking hook formed at a side thereof;a cover plate for covering the upper side of the sample plate, and having a transmission hole formed in a center thereof; andan O-ring for maintaining a seal between the sample plate and the cover plate;a coupling including a plurality of supporting plates, each having a ball plunger to support an outer circumference of the sample plate, and an opened portion enabling the holder to be separated therefrom; anda second transporting device provided at a side of the coupling and being movable along an X-axis, a Y-axis, and a Z-axis by a motor; anda first mirror disposed at a lower side of the holder, and a second mirror disposed at an upper side of the holder, a soft X-ray being generated from the plasma in the light source chamber, the soft X-ray being amplified by the first mirror and illuminating and penetrating the living sample, to obtain an optical image signal; andan image capturing chamber installed at an upper side of the housing to amplify the optical image signal and to capture the optical image signal on an external screen to allow an image of the living sample to be viewed. |
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044977700 | description | DESCRIPTION OF PREFERRED EMBODIMENT The support structure 1 is formed by a plurality of tubes 2 which may be square or rectangular in cross-section. Tubes 2 may be used to receive the nuclear waste. As an alternative, tubes 2 may be subdivided into several chambers 3 for receiving nuclear waste. The cross-section of these chambers depends upon the size of the radioactive material intended to be stored therein. Each storage tube 2 is closed at its lower end by a plate 4. Plates 4 are perforated as indicated at 5 to admit a cooling medium, such as water, to the fuel elements in tubes 2, or chambers 3, respectively. A plurality of bolts 6 is affixed to each tube 2 below plate 4. The number of bolts 6 should be at least three, and is preferably four, as shown in FIGS. 1 and 2. The longitudinal axes of bolts 6 intersect at a common point. To obtain as large a resistance as possible against tripping or tilting forces which may occur when storage tubes 2 are subjected to horizontal forces, bolts 6 are arranged in diagonal directions of storage tubes 2 near the edges of the latter. As shown in FIGS. 1 and 2, four support plates 7 are provided for supporting one storage tube 2, i.e. one plate 7 at each corner of one square or rectangular tube 2. Each of bolts 6 cooperates with one of support plates 7. Support plates 7 have upstanding portions 8. To be more specific, each supporting plate 7 has four upstanding portions 8 so that the entire structure 7,8 has the shape of a box having four side walls and being open at the top. Each upstanding portion 8--comparable to the side wall of a box open at the top thereof--is provided with a recess 9. Recesses 9 are open at the top and closed at the bottom thereof. The top ends, or open ends, of recesses 9 are slanting in opposite directions. Referring to FIG. 2, the open ends of recess 9 are both slanting, the left edge 10 bounding recess 9 slanting downwardly from left to right and the right edge 11 bounding recess 9 slanting downwardly from right to left. The surfaces 10,11 form guides for pins 6. Recess 9 has lower substantially circular bolt-bearing surfaces of which each is coaxially arranged to one of bolts 6. The diameter of these bolt-bearing surfaces is substantially equal to the diameter of bolts 6 to allow an easy engagement of bolt 6 with said bolt-bearing surfaces and at the same time allow a support of bolts 6 without play, or with a minimum of play. When assembling the structure shown in FIGS. 1 and 2, a crane lowers each storage tube 2 into position. Each bolt 6 engages one of recesses 9 and is guided by slanting surfaces 10,11 toward the closed ends of recesses 9 which are adapted to ultimately receive and support bolts 6. The upstanding portions 8 of base plates 7 have a nose-like projection 12 that projects at least to the vertical median plane of the bolt 6 which is lowered into the particular recess 9. These nose-like projections lock bolts 6 in position against the action of upward directed forces. The lateral or upstanding walls 8 are not necessarily planar as clearly shown in FIG. 1 of the drawings. When a tube is lowered into the position shown in FIG. 1, its weight is initially supported by surfaces 10,11, and thereafter the bolts 6 engage the circular or cylindrical portion of recesses 9. During this process of insertion of storage tubes 2 into recesses 9 the former are caused to perform a slight rotary motion that ends when bolts 6 are fixedly supported by walls 8. As shown in FIGS. 1 and 2, the four upstanding portions 8 of each base plate 7 each support one pin 6 forming part of, or affixed to, a different tube 2. All vertical forces resulting from the weight of tubes 2 and that of their content, as well as all horizontal impact forces are transmitted by way of bolts 6 to base plates 7. As shown in FIGS. 1 and 2, each storage tube 2 is held in position by four bolts each engaging one of four base plates 7, and each base plate 7 is adapted to lock in position four storage tubes 2. In FIGS. 3 and 4 the same reference numerals as in FIGS. 1 and 2 have been applied to indicate like parts. Thus FIGS. 3 and 4 require a description only to the extent that the structure shown therein differs from that shown in FIGS. 1 and 2. FIGS. 3 and 4 show one single storage tube 2 for receiving nuclear material. It may be subdivided into a plurality of compartments 3. The tubular member 2 is provided near the bottom thereof with a plurality of horizontal bolts or, to be more specific, four horizontal bolts. Their longitudinal axes intersect at one point designated by the reference letter M. Tubular storage member 2 is supported by a first circular plate 13 having a plurality of upstanding portions 8 which are angularly displaced, e.g. 90 deg. Each upstanding portion 8 of plate 13 supports one of bolts 6. Each upstanding portion 8 is provided at the upper edge thereof with an open recess 9 for the insertion of bolts 6. The details of this recess 9 are the same as shown in FIGS. 1 and 2. They include a relatively wide slanting entrance defined by surfaces or planes 10, 11 for inserting one of said plurality of bolts 6. Recesses 9 further include a substantially circular relatively narrow bottom portion coaxial with one of said plurality of bolts 6. The first circular plate 13 and a second circular plate 4 are arranged parallel to each other. The latter is perforated to allow admission of a cooling medium to the chambers inside of storage tube 2. A spacer to which reference numeral 14 has been applied spaces plates 4 and 13. The structure shown in FIGS. 1 to 4 is not limited to store fuel elements of a nuclear reactor. It may be used to store fuel elements which are still too highly radioactive and require an additional storage time before they can be shipped to a processing plant. The structure of FIGS. 1-4 may also be used for storing new radioactive fuel elements before they are used in a nuclear plant. |
060693614 | claims | 1. A solid state X-ray detector comprising: a first pixellated array sensor sensitive to a first predetermined bandwidth; a second pixellated array sensor sensitive to a second predetermined bandwidth arranged such that the pixels of the second sensor are facing the pixels of the first sensor; and at least one phosphorescent layer sensitive to X-rays and emitting light within the first predetermined bandwidth and the second predetermined bandwidth sandwiched in between the sensors; wherein the first and second bandwidths are different, there are at least two phosphorescent layers with an opaque mask sandwiched between them, such that there is first phosphorescent layer emitting the first bandwidth, and a second phosphorescent layer emitting the second bandwidth. a plurality of pixellated arrays each sensitive to a bandwidth selected for the pixellated array wherein the bandwidths are different, a phosphorescent dot pattern applied to at least one of the pixellated arrays and aligned with the pixels of that pixellated array, the dot pattern being sensitive to X-rays and emitting light in response to X-rays within the bandwidth of the sensor to which it is applied. a plurality of pixellated array sensors each sensitive to bandwidth selected for the sensor wherein the bandwidths are different, a phosphorescent layer applied to at least one of the sensors and aligned with the pixels of that sensor, the phosphorescent layer being sensitive to X-rays and emitting light in response to X-rays within the bandwidth of the sensor to which it is applied. 2. The X-ray detector of claim 1 wherein the first and second pixellated sensors have arrays offset with respect to each other. 3. The X-ray detector of claim 2 wherein the offset is essentially x/2, y/2 where x represents the distance between two adjacent pixels on the first sensor along the x axis and y represents the distance between two adjacent pixels on the first sensor along the y axis. 4. The X-ray detector of claim 1 further comprising at least one mask adjacent to at least one of the sensors, the mask being opaque to the light emitted by the phosphorescent layer and having a plurality of apertures that are aligned with pixels in the sensor. 5. The X-ray detector of claim 1 wherein the phosphorescent layer emitting to the first predetermined bandwidth or the second predetermined bandwidth is applied to the sensor in a phosphor dot pattern that is aligned with at least the pixels of either the first or second sensors. 6. The X-ray detector of claim 1 wherein the first and second pixellated layers are offset with respect to each other. 7. A solid state X-ray detector comprising: 8. The X-ray detector of claim 7 wherein at least two of the pixellated sensors are offset with respect to each other. 9. The X-ray detector of claim 8 wherein the offset is essentially ##EQU2## where x represents the distance between two adjacent pixels along the x axis on the first pixellated array, y represents the distance between two adjacent pixels along the y axis on the first pixellated array, n equals the total number of pixellated arrays in the sensor, and s equals the current pixellated array being offset. 10. The X-ray detector of claim 1 further comprising at least one mask adjacent to at least one of the sensors, the mask being opaque to the light emitted by the phosphorescent layer and having a plurality of apertures that are aligned with pixels in the sensor to which it is adjacent. 11. A solid state X-ray detector comprising: 12. The X-ray detector of claim 11 wherein at least two of the pixellated sensors are offset with respect to each other. 13. The X-ray detector of claim 11 wherein the offset is essentially ##EQU3## where x represents the distance between two adjacent pixels along the x axis on the first pixellated array, y represents the distance between two adjacent pixels along the y axis on the first pixellated array, n equals the total number of pixellated arrays in the sensor, and s equals the current pixellated array being offset. 14. The X-ray detector of claim 11 further comprising at least one mask adjacent to at least one of the sensors, the mask being opaque to the light emitted by the phosphorescent layer and having a plurality of apertures that are aligned with pixels in the sensor to which it is adjacent. |
description | This application claims priority to U.S. patent application Ser. No. 11/997,278, filed on Jan. 29, 2008, which is a United States national stage of Patent Cooperation Treaty application PCT/JP2005/013941, filed on Jul. 29, 2005, the contents of which are hereby incorporated by reference. 1. The Field of the Invention The present invention relates to a method for simply removing a low-concentration radioactive substance, a removing material suitable for the method, and a solvent composition for removing. 2. The Relevant Technology At present, it is considered that materials exposed to radiation in nuclear power plants or the like can be subjected to waste treatment when radioactive substance is removed at a low-concentration level of residual contamination with radioactive substance, where contamination is hardly observed (clearance level). However, since equipment and tools are used under the condition that they are reused, the following methods have been applied: a gentle removal method that does not damage materials; and a method of wiping contaminated parts with a cloth piece moistened with water or a removing agent with little chemical reaction (such as alcohol, acetone, and a synthetic detergent) (refer to Non-Patent Document 1). Even at present, the removal work is done mainly by wiping, for example, with a disposable towel like Kimtowel (manufactured by Nippon Paper Crecia Co., Ltd.) immersed with a 50% by volume aqueous solution of ethanol. However, the current removing material (Kimtowel immersed with a 50% by volume aqueous solution of ethanol) has insufficient removing performance and requires repeated wiping operations. In addition, the cleanliness after wiping depends largely on a worker's impression. Moreover, removing materials can be discarded only after they are dried. Therefore, the use of a disposable towel immersed with a 50% by volume aqueous solution of ethanol, which has poor drying properties, requires drying treatment of a wiped-off surface and the disposable towel before discarded, after wiping. Further, the 50% by volume aqueous solution of ethanol also has a problem against inflammability. Patent Document 1: Japanese Patent Application Laid-Open No. 5-508418. Patent Document 2: Japanese Patent No. 3482488. Patent Document 3: U.S. Pat. No. 5,466,877. Non-Patent Document 1: “RADIOISOTOPES” magazine, pp. 57-62, vol. 23, No. 12, (1974), issued by the Japan Radioisotope Association. The present invention was made in consideration of such a situation, and an object of the present invention is to provide a method for removing a low-concentration radioactive substance simply, a removing material suitable for the removal, and a solvent composition for removing. As a result of intensive research to solve the above problems, the present inventors have found that a removing material immersed with a solvent composition for removing, comprising at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone as a medium for transporting a radioactive substance is effective for the removal of a radioactive substance. A removing material in which a wipe substrate is immersed with a solvent composition for removing of the present invention (hereinafter referred to as “a removing wiper”) is particularly excellent in the removing performance (removal effect) and can substantially reduce the wiping work, which is currently performed repeatedly several times. Further, since the solvent composition for removing of the present invention has excellent drying properties, the time required for the drying that is currently performed after wiping work can be substantially shortened or omitted. Furthermore, since the removing solvent composition of the present invention is inflammability, it can also eliminate the danger of ignition. One aspect of the present invention is a solvent composition for removing radioactive substance characterized by comprising at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone as a medium for transporting the radioactive substance. Hydrofluoroether or perfluoroketone preferably has 4 to 8 carbon atoms. Specifically, the hydrofluorocarbon is preferably C2H2F10, C4H5F5, c-C5H3F7, or C7HF15. Further, the hydrofluoroether is preferably C4F9OCH3, C4F9OC2H5, C2HF4OC2H2F3, or F(CF(CF3)CF2O)CHFCF3. Furthermore, the perfluoroketone is preferably CF3CF2C(O)CF(CF3)2, (CF3)2CFC(O)CF(CF3)2, or (CF3)2CFCF2C(O)CF(CF3)2. The solvent composition for removing radioactive substance according to the present invention can further comprise at least one organic solvent selected from alcohol, ketone, ether, ester, hydrocarbon, halogenated hydrocarbon, glycol ether, or a silicone-based organic solvent. Among these, it is preferred that the composition comprise alcohol. As the alcohol, it is preferred to use methanol, ethanol, 1-propanol, 2-propanol, 1-butanol, 2-butanol, t-butanol, or a mixture thereof. The organic solvent can be contained in an amount of from 1 to 50% by weight based on the total weight of the removing solvent composition. Another aspect of the present invention is a material for removing radioactive substance, characterized in that the material is immersed with a removing solvent composition comprising at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone as a medium for transporting a radioactive substance. The removing material of the present invention can be prepared by immersing a wipe substrate with the solvent composition for removing of the present invention. It is preferred to use a nonwoven fabric as a wipe substrate. Further, it is preferred to use a wipe substrate comprising at least one selected from pulp, synthetic fiber, cellulose, and regenerated cellulose. Still another aspect of the present invention is a method for removing a radioactive substance, characterized by using at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone as a medium for transporting the radioactive substance. Further, the present invention is a method for removing a radioactive substance, characterized by comprising the steps of: bringing a surface of an article with the radioactive substance adhered thereto into contact with a removing material immersed with the solvent composition for removing of the present invention; and adsorbing the radioactive substance to the removing material, thereby recovering the radioactive substance. As the removing material, the removing wiper according to the present invention can be used. Hereinafter, the present invention is described in detail. The solvent composition for removing radioactive substance according to the present invention comprises at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone as a medium for transporting a radioactive substance. In the present invention, the optimum medium for transporting a radioactive substance is selected from at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone depending on the type of contamination, the type of contaminees, and the like. The solvent composition for removing is preferably a liquid at room temperature (has a boiling point at room temperature or above), preferably having a boiling point of 30° C. to 100° C., and it preferably has 4 to 8 carbon atoms. From a viewpoint of safety, the compound having low level of toxicity, preferably, 100 ppm or more of permissible concentration level (ppm (Vol)) is used. Further, similarly, the compound having low inflammability, preferably no flash point (according to JIS K2265) is used. Furthermore, the compound having a low global warming potential (GWP) is preferably used from an environmental point of view. An increase in the number of fluorine atoms in compound results in an increase in non-inflammability, and an increase in the molecular weight tends to raise a boiling point. Thus, a compound may be suitably selected according to the purpose. For example, in order to improve drying properties, a compound having a small molecular weight may be used, or it may be mixed with a highly volatile organic solvent or the like. (a) Hydrofluorocarbon (HFC) Examples of the hydrofluorocarbons used in the present invention include 1,1,1,2,2,3,4,5,5,5-decafluoropentane, 1,1,1,3,3-pentafluorobutane, 1,1,2,2,3,3,4-heptafluorocyclopentane, and 1H-perfluoroheptane. Among these hydrofluorocarbons, C5H2F10, C4H5F5, c-C5H3F7, or C7HF15 is preferred in terms of the removal effect and having a boiling point of from 30° C. to 100° C., no flash point, and low toxicity. The above hydrofluorocarbons may be used alone or in combination of two or more. These hydrofluorocarbons can be prepared by a known method; or those commercially available may be used; or they may be produced, for example, using a method described in Patent Document 1. (b) Hydrofluoroether (HFE) Examples of the hydrofluoroethers used in the present invention include CF3CF2CH2OCHF2, CF3CHFCF2OCH3, CF3CH2OCF2CH2F, CF3CHFCF2OCH2CF3, nonafluorobutyl methyl ether, nonafluorobutyl ethyl ether, 1,1,2,2-tetrafluoroethyl-2,2,2-trifluoroethyl ether, and 2H-perfluoro(5-methyl-3,6-dioxanonane). Among these hydrofluoroethers, C4F9OCH3, C4F9OC2H5, C2HF4OC2H2F3, or F(CF(CF3)CF2O)CHFCF3 is preferred in terms of the removal effect and having a boiling point of from 30° C. to 100° C., no flash point, and low toxicity. These hydrofluoroethers can be prepared by a known method; or those commercially available may be used; or they may be produced, for example, using a method described in Patent Document 2. These hydrofluoroethers may be used alone or in combination of two or more. (c) Perfluoroketone Examples of the perfluoroketones used in the present invention include CF3(CF2)5C(O)CF3, CF3CF2CF2C(O)CF2CF2CF3, CF3CF2C(O)CF(CF3)2, (CF3)2CFC(O)CF(CF3)2, (CF3)2CFCF2C(O)CF(CF3)2, CF3(CF2)2C(O)CF(CF3)2, CF3(CF2)3C(O)CF(CF3)2, CF3CF2C(O)CF2CF2CF3, and CF3OCF2C(O)CF(CF3)2. Among these hydrofluoroethers, CF3CF2C(O)CF(CF3)2 is preferred in terms of the removal effect and having a boiling point of from 30° C. to 100° C., no flash point, and low toxicity. These perfluoroketones can be prepared by a known method; or those commercially available may be used; or they may be produced, for example, using a method described in Patent Document 3. These perfluoroketones may be used alone or in combination of two or more. Further, in the present invention, hydrofluorocarbon, hydrofluoroether, and perfluoroketone used as a medium for transporting a radioactive substance may be used alone or in combination of two or more. To the solvent composition for removing of the present invention, in order to further improve the removal performance, may be added an organic solvent such as alcohol, ketone, ether, ester, hydrocarbon, halogenated hydrocarbon, glycol ether, and a silicone-based organic solvent. Examples of alcohol include methanol, ethanol, 1-propanol, 2-propanol, 1-butanol, 2-butanol, and t-butanol. Examples of ketone include acetone and methyl ethyl ketone. Examples of ether include diethyl ether. Examples of ester include methyl acetate and ethyl acetate. Examples of hydrocarbon include hexane, heptane, and isooctane. Examples of halogenated hydrocarbon include trans-1,2-dichloroethylene and 1,1-dichloro-2,2,3,3,3-pentafluoropropane. Examples of silicone-based organic solvent include hexamethyldisiloxane. Examples of glycol ether include 1,2-diethoxyethane. These organic solvents may be used alone or in combination of two or more. Inflammable organic solvents such as alcohol, ether, and the like are preferably used in relatively low concentrations. The amount of these organic solvents to be added may be appropriately set in terms of inflammable, compatibility, and the like, but these organic solvents can be added in a proportion of 1 to 50%, preferably 2 to 30%, more preferably 3 to 15%, by weight, relative to the total weight of the solvent composition for removing. When an alcohol is used as an organic solvent, an increase in the amount of alcohol to be added increases the removal effect, but it tends to increase the time until the used removing solvent composition dries. Therefore, it is preferred to add an alcohol in a proportion of 2 to 30%, more preferably 3 to 15%, by weight, relative to the total weight of the removing solvent composition. Next, the material for removing the radioactive substance of the present invention is described. The material for removing radioactive substance of the present invention is characterized by comprising at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone. The removing material of the present invention is preferably a removing wiper prepared by immersing a wipe substrate with the above removing solvent composition of the present invention. Note that, in the present invention, a “wiper” means the generic name of what is used for wiping the surface of an article. The wipe substrate is not particularly limited as far as it holds a liquid removing solvent composition of the present invention and comprises a material that can be used for wiping the surface of an article. However, it is preferred to use the one comprising of at least one selected from pulp, synthetic fiber, cellulose, and regenerated cellulose in terms of availability and cost. The form of a wipe substrate is not particularly limited as far as the substrate is processed from the above materials, but it is preferred to use a form that can maintain a certain degree of strength when it is used for wiping. It is preferred to use a nonwoven fabric in that it has high wiping effect and fibers are hard to remain. The nonwoven fabric to be used is not particularly limited, but the most suitable one can be selected depending on the type of contamination, the type of contaminees, and the like. Examples include pulp, pulp/synthetic fiber, pulp/rayon, pulp/synthetic fiber/rayon, rayon, rayon/synthetic fiber, pulp/lyocell, pulp/synthetic fiber/lyocell, lyocell, lyocell/synthetic fiber, synthetic fiber, and cotton yarn. Examples of synthetic fiber include polyethylene terephthalate, polybutylene terephthalate, nylon and/or polyolefines such as polypropylene, polyethylene, and poly-4-methyl-1-pentene. The thickness of a nonwoven fabric can be suitably selected depending on the application of the removing wiper of the present invention, and it is generally preferably from about 10 μm to about 3 mm. Further, the mass per unit area of a nonwoven fabric can be suitably selected depending on the application, and it is generally preferably from 10 to 500 g/m2. A method for manufacturing a nonwoven fabric used for the removing wiper of the present invention is not particularly limited, but the nonwoven fabric can be manufactured by generally used methods such as water jetting, needle punching, stitch bonding, chemical bonding, thermal bonding, spun-bonding, meltblowing, and wet process. Further, the wipe substrate for the removing wiper of the present invention is not limited to a fabric form as described above, but a wipe substrate having a porous structure such as sponge may be used. In the present invention, a method for immersing a wipe substrate with a solvent composition for removing is not particularly limited, but it can be performed by a generally used method, for example, by immersing a wipe substrate in a solvent composition for removing or by spraying a solvent composition for removing onto a wipe substrate. A method for removing a radioactive substance according to the present invention is characterized by using at least one selected from hydrofluorocarbon, hydrofluoroether, and perfluoroketone as a medium for transporting a radioactive substance. As a medium for transporting a radioactive substance, a suitable composition can be appropriately selected and used according to the description about the solvent composition for removing of the present invention as described above. Further, the present invention is a method for removing a radioactive substance, characterized by comprising the steps of: bringing a surface of an article with a radioactive substance adhered thereto into contact with a removing material immersed with the solvent composition for removing of the present invention; and adsorbing the radioactive substance to the removing material, thereby recovering the radioactive substance. As a removing material, the removing wiper according to the present invention can be used. In the step of bringing a surface of an article with a radioactive substance adhered thereto into contact with a removing material immersed with the solvent composition for removing of the present invention, the method for bringing the removing material into contact is not particularly limited, but a larger amount of the radioactive substance can be adsorbed to the removing material as the area of contact with the surface of the article with the radioactive substance adhered thereto becomes larger. Next, the evaluation method of the solvent composition for removing of the present invention is described below. (Evaluation of the Removal Solvent Composition) The solvent composition for removing of the present invention was evaluated for the following points 1 to 3. 1. Test for confirming removal: Level of removal was evaluated by measuring the amount of hematite (Fe2O3), which is recognized as a simulated material of a radioactive contaminant, adhering to a wipe substrate immersed with a solvent composition for removing. Specifically, the weight (A) of a wipe substrate (Sontara (registered trademark), a rayon/polyester mixed product manufactured by Du Pont Kabushiki Kaisha) or Kimtowel (100% pulp, manufactured by Nippon Paper Crecia Co., Ltd.) having an area of 70 cm2 was measured first. Then, a wipe substrate immersed with a solvent composition for removing shown in Table 2 was attached to a fixture and a weight (500 g) was put on the fixture. The wipe substrate was moved 500 mm on a surface to be removed (No. 1 finished-surface of SUS 304 stainless steel; weight of a simulated material adhered: 0.3 mg/cm2) which is previously applied, as a radioactive contaminant, with hematite (Fe2O3) (obtained by heat-treating iron (III) oxide manufactured by Kanto Chemical Co., Inc. at 600° C.), which is recognized as a simulated material of a radioactive contaminant. Then, the wipe substrate was removed from the fixture and dried for two days at room temperature, and the weight (B) of the wipe substrate after drying was measured. The ratio of the contaminant adhered to the wipe substrate [(B−A)/area of wipe substrate (70 cm2)] was determined from the difference of the weight (B−A) measured in this way. 2. Drying test: The wipe substrate was immersed with a solvent composition for removing, put on a balance at room temperature, and measured for the time until it dries. Thus, drying properties were evaluated. 3. Non-inflammability test: The flame of a lighter was brought close to a glass petri dish in which a solvent composition for removing was put, and non-inflammability was evaluated by whether the solvent composition ignites or not. Examples of the present invention are described below, but the present invention is not limited to the inventions disclosed in Examples. In Examples, the solvent compositions for removing shown in Table 1 were evaluated. TABLE 1Solvent compositions for removingTrade name1,1,1,2,2,3,4,5,5,5-Vertrel (registered trademark) XF,decafluoropentanemanufactured by Du Pont-MitsuiFluorochemicals Company, Ltd.Mixture of 96 wt % ofVertrel (registered trademark) XE,1,1,1,2,2,3,4,5,5,5-decafluoropentanemanufactured by Du Pont-Mitsuiand 4 wt % of ethanolFluorochemicals Company, Ltd.Nonafluorobutyl methyl etherNovec 7100 (registered trademark)manufactured by 3M LimitedNonafluorobutyl ethyl etherNovec 7200 (registered trademark)manufactured by 3M LimitedMixture of 90 wt % ofVertrel (registered trademark)1,1,1,2,2,3,4,5,5,5-decafluoropentaneX-E10, manufactured by Du Pont-and 10 wt % of ethanolMitsui Fluorochemicals Company,Ltd. The test for confirming removal was performed using Vertrel (registered trademark) XF, Vertrel (registered trademark) XE, Vertrel (registered trademark) X-E10, Novec 7100 (registered trademark) (manufactured by 3M Limited), and Novec 7200 (registered trademark) (manufactured by 3M Limited) as solvent compositions for removing; and a nonwoven fabric (Sontara (registered trademark), a rayon/polyester mixed product manufactured by Du Pont Kabushiki Kaisha) as wipe substrates. The results are shown in Table 2. The test for confirming removal was performed by the same operation as in Example 1 except that a 50% by volume aqueous solution of ethanol was used as a solvent composition for removing and Kimtowel was used as a wipe substrate. The results are shown in Table 2. TABLE 2Wipe substrate contamination ratioSolvent compositions for removingThe amount wiped off, mg/cm2Vertrel (registered trademark) XF0.679Vertrel (registered trademark) XE0.834Vertrel (registered trademark) X-E101.393Novec 7100 (registered trademark)1.061Novec 7200 (registered trademark)0.89050 vol % aqueous ethanol solution0.636 The drying test was performed using Vertrel (registered trademark) XF, Vertrel (registered trademark) XE, and Vertrel (registered trademark) X-E10 as solvent compositions for removing and Kimtowel (manufactured by Nippon Paper Crecia Co., Ltd.) as a wipe substrate. Kimtowel cut to a 50 mm square (0.05 g) was immersed in Vertrel (registered trademark) XF, Vertrel (registered trademark) XE, or Vertrel (registered trademark) X-E10 for 1 minute. Then, the resulting Kimtowel was transferred to a balance and measured for the time until its weight returns to the initial weight of the wipe substrate. The drying time of the solvent composition for removing is shown in Table 3. In this Comparative Example, the same operation as in Example 2 was performed using a 50% by volume aqueous solution of ethanol instead of the solvent composition for removing. Drying time is shown in Table 3. TABLE 3Drying timeSolvent compositions for removingThe amount wiped off, mg/cm2Vertrel (registered trademark) XF57Vertrel (registered trademark) XE68Vertrel (registered trademark) X-E1013050 vol % aqueous ethanol solution2803 At room temperature, 20 ml of Vertrel (registered trademark) X-E10 was put in a glass petri dish having an inner diameter of 85 mm. When the flame of a lighter was brought close to the upper surface of the petri dish, the flame went out. At room temperature, 20 ml of a 50% by volume aqueous solution of ethanol was put in a glass petri dish having an inner diameter of 85 mm in the same manner as in Example 3. When the flame of a lighter was brought close to the upper surface of the petri dish, the solution continued burning on the liquid surface with a blue flame even after the flame of the lighter is removed. The present invention provides a solvent composition that exhibits excellent removal effect in the work for removing a radioactive substance from the equipment and the like with the radioactive substance adhered thereto in nuclear power plants, hospitals, airplanes, and the like. In addition, since the solvent composition has excellent evaporation properties from the equipment and the like after the removal work, it allows the treatment after removal to be done easily. |
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047626712 | abstract | Injection device used when a reactor is shut down to remove radioactive substances from within the gap between the internal piping and the inside of the nozzles of a RPV by injecting fluid towards the inside of these nozzles. The injection device is equipped with an injection nozzle that injects fluid towards the end of the gap, a casing that supports this injection nozzle, a suspension mechanism whereby the casing is movably suspended from above the RPV, a water feed device that supplies high-pressure fluid to the injection nozzle, a sensor or the like that senses when the casing is positioned directly above the internal piping and a fixing mechanism that temporarily fixes the casing to the inside wall of the RPV. |
claims | 1. A system, comprising:a charged particle source comprising a tip;a charged particle column;a detector configured to obtain an image of the tip of the charged particle source;a moveable optical reflective element having a first position in the charged particle column and a second position outside the charged particle column; anda positioning device configured to move the optical reflective element between its first and second positions, the positioning device being configured to move in a first plane and in a second plane perpendicular to the first plane,wherein:the tip is configured to emit light when heated;the tip is configured to emit charged particles during use of the system; andthe moveable optical reflective element is configured in the first position to reflect light from the charged particle source to the detector to obtain the image. 2. The system of claim 1, wherein in the second position, the optical reflective element cannot reflect light passing through the charged particle column to the detector. 3. The system of claim 1, further comprising:a first actuator configured to move the positioning device in the first plane; anda second actuator configured to move the positioning device in the second plane. 4. The system of claim 1, wherein the charged particle source is configured so that during use at least some of the charged particles generated by the charged particle source pass through the charged particle column and wherein the charged particle source is configured so that, when it emits light, the light goes into the column and can be reflected by the optical reflective element when it is in the first position. 5. The system of claim 1, wherein the charged particle column is an ion beam column. 6. A system, comprising:a charged particle source comprising a tip;a charged particle column having an axis;a detector configured to obtain an image of the tip of the charged particle source; andan optical reflective element positioned within the charged particle column and displaced off-axis with respect to the axis of the charged particle column, the optical reflective element being coupled to the charged particle column and configured to reflect light from the charged particle source to the detector,wherein the optical reflective element is fixed with respect to the charged particle column. 7. The system of claim 6, further comprising:a housing; anda mount fixedly mounting the optical reflective element to the housing,wherein the mount, the optical reflective element, the charged particle source and the charged particle column are housed within the housing. 8. The system of claim 6, wherein the charged particle source is configured so that during use at least some of the charged particles generated by the charged particle source pass through the charged particle column without interacting with the optical reflective element and wherein the charged particle source is configured so that, when it emits light, the light goes into the column and can be reflected by the optical reflective element. 9. The system of claim 6, wherein the charged particle column is an ion beam column. 10. A method, comprising:emitting light from a tip of a charged particle source so that the light enters a charged particle column;reflecting at least a portion of the light in the charged particle column to a detector;obtaining an image of the tip of the charged particle source based on detected light; andbased on the image of the tip, modifying a shape of the tip. 11. The method of claim 10, wherein modifying the shape of the tip comprises changing at least one parameter selected from the group consisting of a temperature of the tip of the charged particle source, a gas pressure of a chamber housing the charged particle source, and an intensity of light emitted by the charged particle source. 12. The method of claim 11, further comprising, based on the detected light, increasing at least one parameter selected from the group consisting of a charged particle source temperature and a gas pressure in a chamber housing the charged particle source. 13. The method of claim 10, wherein the charged particle source is a gas field ion source. 14. The method of claim 10, comprising making a tip of a charged particle source. 15. The method of claim 14, wherein the charged particle source is a gas field ion source. 16. The method of claim 10, further comprising, before reflecting at least the portion of the light in the charged particle column to the detector:mounting an optical reflective element on a positioning device;moving the positioning device in a first plane; andmoving the positioning device in a second plane which is perpendicular to the first plane. |
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description | This application claims priority under 35 U.S.C. § 119(e) from provisional patent Application No. 60/562,449, filed Apr. 14, 2004. The 60/562,449 Application is incorporated herein by reference. The present invention relates to a suit designed to protect the wearer from explosions and designed to allow remote retrieval of the wearer. The disclosed invention is a protective suit with an internal harness that connects to a flexible tether. The suit also employs an attached respirator tie down that eliminates the need for a second harness, and wrist and ankle closures to stop explosive gasses or other flammable material from entering the interior of the suit and igniting. The suit also employs a removable hood that provides protection from high temperatures, prevents gas buildup in the hood, and can be easily removed. A number of patents have separately dealt with suits to protect the wearer from fires and harnesses to extract the wearer from a dangerous area or retard a fall. The prior art has not integrated a fire and explosion protection suit with a built-in extraction harness. A number of patents teach safety harnesses. U.S. Pat. No. 2,979,153 to Hoagland et al. teaches the use of an internal harness which tightens onto the limbs of the wearer when used, which could cause further injury to the wearer. U.S. Pat. No. 3,973,643 to Hutchinson teaches a detachable waist harness in a fireman's turn-out coat. U.S. Pat. No. 4,273,216 to Weissmann teaches a harness mounted to the outside of a jacket. U.S. Pat. No. 4,682,671 to Hengstenberger et al. Teaches a harness loop that wraps under the arms and behind the head, and a jacket. U.S. Pat. No. 4,854,418 also to Hengstenberger et al. teaches the same harness and jacket with the addition of a crotch strap. Neither of the Hengstenberger et al. patents teach the use of a full body extraction harness integrated with the interior of a flash suit. It will also be appreciated that the harness loop arrangement of Hengstenberger et al. Is prone to causing neck injuries when in use. U.S. Pat. No. 5,960,480 to Neustater et al. teaches a harness inside a coverall. Like the harness of Hoagland, the harness fits loose most of the time, but cinches tight during a fall. U.S. Pat. No. 6,256,789 to Young et al. teaches a fall arresting harness integrated into a garment, in order to maximize the surface area acted on by the harness. The arrangement of the self-tightening harness is similar to those taught by Hoagland and Neustater et al. Several patents assigned to E. I. Du Pont de Nemours and Company (“DuPont”) relate to fire resistant suits. None of these patents teach the use of an integral extraction harness. U.S. Pat. No. 5,048,124 to Lewis Jr. et al. teaches “Easy Access Protective Coveralls”, constructed of a shell to withstand high temperatures and laminated with a liquid impervious layer, and a multilayer liner. U.S. Pat. No. 5,050,241 to Flowers et al. Teaches a multilayer outer shell that has a vapor-permeable, liquid impermeable sheet sandwiched between a woven sheet and an insulating inner layer. U.S. Pat. No. 5,279,287 to Wiseman Sr. teaches a suit, similar to the suit disclosed in Lewis et al., made up of woven fabric with an aluminum layer adhered to it, and includes a detachable head and respirator covering. U.S. Pat. No. 6,490,733 teaches the use of a harness in a pant portion of a suit, but again this would not provides sufficient protection in a closed area where combustible gasses are present and could built up under a suit that was made of two separate garments. U.S. Pat. No. 6,487,725 teaches the storage of a lanyard in the harness, but does not provide for use of the lanyard without obstructing the work of the wearer. Existing flash protection suits consist of a garment of one layer of flame protective cloth and a separate external fall harness layered on over the garment. The use of an external harness is inconvenient and can cause an explosion if the metal buckles and clips of the harness create a spark. These suits are used in situations where combustible gasses are present or build up, such as inside large pipelines or tanks, and where the risk of explosion is very high. In the case of an explosion, the person wearing the suit must be protected from possible burns due to the high temperatures. In addition, the force of the explosion will often render the person wearing the suit unconscious. The suit must also provide a way to retrieve the person wearing the suit without endangering the lives of those attempting the retrieval. Current suits with integrated harnesses place the harness on the exterior of the suit, which prevents a secure harness attachment to the wearer and allows the harness to shift and move. Moreover, an external harness decreases the effectiveness of the flame proof material by cinching and bunching the garment material against the wearer and thereby decreasing the thickness of the insulating material, squeezing out insulating air pockets, and allowing heat to penetrate the garment more quickly. At this time there is no garment on the market that addresses the issues of multi-layered flash fire protection, retrievable built-in one-piece harness with lanyard, and respirator hose tie-down, in one protective garment eliminating the need for separate garments and harnesses to protect the worker. At present, available off the shelve Fire Resistant (FR) Flash/Coverall garments merely meets the three-second test criteria to qualify as Fire Resistant material. These off the shelve single layer FR, natural or aramid, Personal Protective Equipment garments provide some thermal protection for a person engulfed in a gas vapor ignition for less than one to two seconds. Any exposure to the flash-fire over the three-seconds exposes the worker to significant thermal burns to the body and head. Within the field, it is highly desirable and sought-after to provide a flash suit capable of thermal protection for a person engulfed in a gas vapor ignition for at least eight seconds. Needed is a multi-layer system that will provide added protection to the worker. A multi-layer flash suit must also provide unrestricted movement and comfort compared to single layer garments. The present invention provides an article of clothing that provides protection from high temperatures due to flames, explosions, or combustion. Another purpose of the invention is to provide substantial protection to the wearer from burn injury for an eight second time period. Another purpose of the invention is to allow retrieval of the wearer. Another purpose of the invention is to provide wrist and ankle opening seals to prevent gas or explosive material from building up in the suit and causing an explosion internal to the suit. Another purpose of the invention is to provide an internal harness that will allow remote retrieval of the wearer in case of accident. Another purpose of the invention is the provision of a flash suit that allows retrieval of the wearer from the source of the flames without having to endanger the rescue personnel. Another purpose of the invention is to provide an internal harness that cannot cause sparks and create the risk of explosion. Another purpose of the invention is to provide an integrated respirator tie down so that an additional harness is not needed. Another purpose of the invention is to provide a removable hood that protects the head and neck of the wearer from burns, but does not allow gas build up in the hood. Another purpose of the invention is to provide a storage pouch for a flexible retrieval lanyard so that the lanyard can be easily stored with the suit. Another purpose of the invention is to provide a grounding lead to the suit, further preventing the possibility of spark in the hazard area. FIG. 1a shows a front view of the flash suit 1. The garment is a single-piece suit having a torso portion 10, an opening collar 11, a front opening 12, arms 13, wrist closure tabs 14, waistband 15, legs 17, ankle closure tabs 18, and a hook-and-loop panel 16 mounted on the ankle. The hook-and-loop panels described herein are complimentary panels of hook panels and loop panels which, when pressed together, stick together. These hook-and-loop fasteners are commonly referred to by the trademark “Velcro”. In this application, the choice of hook or loop panel for a particular closure is not important, so the complimentary panels are referred to by the same reference number, 16. The wrist closure tab 14 is placed near the end of the arm or sleeve 13, and by pulling the tab 14 and fastening it to its respective panel 16 (the wrist panel 16 is not seen from the front of the garment and is not shown in FIG. 1, but may be seen in FIG. 2), the wrist portion of the sleeve 13 may be wrapped tightly around the wearer's wrist. Similarly, the ankle tab 18 is located at the bottom of the garment's trouser leg 17 and may be pulled and fastened to a panel 16 on the leg 17, thereby tightly wrapping the ankle portion of the leg 17 around the wearer's leg. By providing these closure tabs, 14 and 18, the suit can restrict gases from entering the interior of the garment. The danger of a gas build-up in the interior of a flash suit is that, should and explosion or fire take place, the gases inside the suit could be ignited. The hook-and-loop panels 16 are constructed from a fire resistant material capable of withstanding high heat without melting are catching fire. One such commercially available product is Aplix® #820 hook and loop from Aplix Inc. The hook panel is constructed of a woven fire resistant base, such as the Nomex® manufactured by E.I. Du Pont de Nemours and Company. The Nomex® material is discussed in more detail below, in connection with the materials used to construct the outer shell of the flash suit 1. Atop the Nomex® base of the hook panel are mounted 6.5 mil nylon monofilament hooks. The loop panel is constructed of a woven Nomex® base, atop which is mounted unnapped Nomex® hoops. The described hook-and-loop material is non-flammable and non-melting. Each of the panels is individually resistant to high temperatures and, when fastened together, are resistant to higher temperatures. As shown in FIG. 11, the closure tabs, 14 and 18, are constructed of two layers of fire resistant 7.5 oz Nomex® Yellow Tab fabric, stitched together, turned and top stitched. A loop panel 16, constructed as described above, is sewn to the tab, 14 or 18. FIG. 1b shows detail of the front opening 20 of the suit 1. The opening 20 extends down from the collar 11 and provides an opening sufficient to allow the wearer to put on and take off the suit with ease. Preferably, the opening 20 extends below the waistband 15 to a point near the wearer's crotch, which affords convenient access and day-to-day practicality. A zipper 21 allows the opening 20 to be securely closed and easily opened. The zipper is 21 preferably a heavy duty, number 10, nylon, one-way zipper. A storm flap 12 is provided on the exterior of the torso portion 10 to create a seal over the zippered opening 20. The storm flap 12 has a hook-and-loop strip 22 mounted to the interior of the flap 12. A complimentary hook-and-loop strip 23 is mounted to the torso 10, so that, when pressed together, the flap 12 will be securely sealed to the torso 10. As described above with respect to the wrist and ankle closure tabs, 14 and 18, the hook-and-loop strips, 22 and 23, should be of a non-flammable and non-melting construction, such as the Aplix® #820 hook & loop product. When properly sealed, the storm flap 12 restricts gases and flames from entering the suit. The collar 11 is also provided with a hook-and-loop fastener to seal the collar 11 around the wearer's neck. A “Nero” collar 11 is formed when the collar's closure tab 19 fastens over the collar. The tab 19 has a non-flammable and non-melting hook-and-loop panel 24 on its interior surface which secures to a complimentary panel 25 on the collar. As with the wrist and ankle tabs, 14 and 18, the collar tab 19 can seal over the wearer's neck and restrict the entry of gases and flames. FIG. 2 shows the rear, exterior portion of the suit 1. On the upper portion of the rear torso 10, an exterior pouch 32 is provided for storing an emergency remote retrieval lanyard 75. The pouch is closed with a upper flap 33, and the flap is sealed with complimentary non-flammable and non-melting hook-and-loop panels 36, such as the type described above. The construction of the harness 61 (as seen, e.g., in FIG. 6), and lanyard 75 (as seen in FIG. 10), are described in greater detail below. The pouch 32 is designed to store a lanyard 75. For example, as seen in FIG. 10, a woven lanyard 75 may be coiled 77 for storage in the pouch 32. Lower on the rear portion is a tie-down D-ring 35 mounted on a D-ring panel 34. The tie-down ring 35 is suitable for securing a respirator (not shown), and is also suitable as a grounding point to prevent the build-up of static electricity. As seen in FIG. 12a, the tie-down ring may be formed of a D-ring 35 held by a non-flammable and non-melting web strap 36. Preferably, the strap 36 is constructed of commercially available two-inch natural Kevlar® webbing, having a thickness of about 0.062 inch, a weight of about 1.8 ounces per yard, a ground warp count of type 964 Kevlar® and Kevlar® catchcord of about 272, a binder warp count of type 964 Kevlar® of about 62, and breaking strength of about 7,300 pounds. Kevlar® is described in U.S. Pat. No. 5,050,241 as poly(p-phenylene terephthalamide) fiber, and is commercially available from E. I. Du Pont de Nemours and Company. Kevlar® has properties of high strength and fire resistance. The tie-down ring strap 36 is held in place by a tie-down panel, as seen in FIG. 12b. The strap is sewn together 80 to form a loop to hold the D-ring 35. Upper and lower ends of the strap, 78 and 79, are box X stitched 71 to the panel 34. FIG. 8b shows an exemplary box X stitch. Between the waistband panel 15 of the outer suit 10 and the tie-down panel 34, the upper end of the strap 78 is secured with box X stitch 71, the lower end of the strap 79 is secured with box X stitch 71, and the loop 36 emerges from the tie-down panel 34 from slot 81. The tie-down panel 34 is stitched to the suit 10. The tie-down ring 35 is available for securing a piece of equipment, such as a respirator, and eliminates the need for a separate harness. The elimination of external harnesses increases the fire protection effectiveness of the suit 1. When an external harness is present, the material of the suit 1 is bunched together and compressed underneath the harness. As noted above, an external harness decreases the effectiveness of flame proof material by cinching and bunching the garment material against the wearer, thereby decreasing the thickness of the insulating material, squeezing out insulating air pockets, and allowing heat to penetrate the garment more quickly. Thus, use of the integrated tie-down ring 35 eliminates the need for an additional external harness and increases the fire protective ability of the suit 1. FIG. 3a shows a pattern for the panels that make up the outer shell of flash suit 1. The panel's are taken from a fire-resistant material, preferably Nomex® IIIA® 7.5 ounce per square yard (“ospy”), or Nomex® 7.5 ospy. Nomex® is described in U.S. Pat. No. 5,050,241 as poly(m-phenylene isophthalamide) fiber, and is a trademarked material owned by E. I. Du Pont de Nemours and Company (“Du Pont”) and commercially available. As more fully described in U.S. Pat. No. 6,132,476 to Lunsford et al., the Nomex® fibers are an aromatic polyamide, which are formed by reactions of aromatic diacid chlorides with aromatic diamines to produce amide linkages in an amide solvent, and referred to by the generic term aramid fiber. Aramid fibers are typically available in meta-type fibers composed of poly(m-phenylene isophthalamide), referred to as meta-aramid fibers, and para-type fibers composed of poly(p-phenyleneterephthalamide), referred to as para-aramid fibers. Meta-aramid fibers are currently available from Du Pont in several forms under the NOMEX® trademark. NOMEX IIIA®, sometimes referred to as NOMEX T-462®, is 93% NOMEX®, 5% KEVLAR®, and 2% carbon core nylon. The 7.5 opsy NOMEX IIIA® provides fire resistance and light weight. Referring again to FIG. 3a, Panel 41 is the rear portion of the outer torso 10. Panel 46 is the waistband 15. Panel 43 is one side of the front outer shell torso 10 and leg 17, and panel 42 is the other side. Panel 48 is the collar 11. Panel 47 is the storm flap 12. Panels 49 and 50 are the arm or sleeve portions 13. Panel 44 is the rear portion of one of the legs 17, and panel 45 is the rear portion of the other leg. Panel 34 is the tie-down panel. Panel 51 is the back reinforcement panel. Panels 14 and 18 are each halves of the tab closures. FIG. 3b shows an alternative pattern for the panels that make up the outer shell of flash suit 1. It will be appreciated to those familiar with the construction of garments, that the patterns shown in FIGS. 3a and 3b allow each of the panels to be cut from a single sheet of material. Many alternative patterns are possible. The flash suit 1 is constructed of an outer shell, as seen in FIGS. 1a through 3b, and an inner liner, as seen in the pattern shown in FIG. 4. The inner liner is constructed from panels of thermal insulating material, such as 3-layer E-89® NOMEX®/KEVLAR® quilted fabric, a mixture of meta-aramid and para-aramid fibers, commercially available from DuPont. Panel 53 is the rear portion of the inner liner torso 10. Panel 54 is one side of the front inner liner torso 10 and leg 17, and panel 55 is the other side. Panels 58 and 59 are the arm or sleeve portions 13. Panel 56 is the rear portion of one of the legs 17, and panel 57 is the rear portion of the other leg. As with the pattern shown in FIG. 3a, it will be appreciated to those familiar with the construction of garments, that the pattern shown in FIG. 3a allows each of the panels to be cut from a single sheet of material. Many alternative patterns are possible. FIGS. 5 through 9a show the interior, integral harness 61. The harness 61 is constructed from 6000 psi nylon seat belt webbing sewn with KEVLAR® thread. All high stress areas of the harness 61 are sewn with double box X stitches 71 (as seen in FIG. 8b). The harness 61 extends outside of the suit through slot 73 in back reinforcement panel 51. The segment of harness that protrudes from the suit is constructed of natural KEVLAR® webbing, as described above with respect to the tie-down web strap 35. FIG. 5 shows a front view of the harness 61. A right-side strap 62 runs over the right shoulder and straight down to the waist, meeting the waist strap 64 on the right side from the middle of the torso 10. Symmetrically, the left-hand strap 63 runs over the shoulder on the left side. A right leg band 65 encircles the right thigh of the leg 17, and a left leg band 66 encircles the left thigh. FIG. 6 shows a rear view of the harness 61. The right-side strap 62 and left-side strap 63 run over the shoulders and straight down, past the waist strap 64, and down to the thigh bands, 65 and 66. An upper back reinforcement strap 67 runs horizontally across from the right and left shoulder straps, 62 and 63. A right-side reinforcement strap 68 runs diagonally from the point where the right-side strap 62 meets the waistband 64 up to the upper reinforcement strap 67, then out through slot 73 to the exterior of the suit where it folds over on itself to form a loop 74, then runs back down diagonally to the point where the left-side strap 63 meets the waistband 64. The double cross-hatching of the loop 74 and reinforcement straps 68 and 69 indicates that these portions of the harness 61 are constructed of fire-resistant Kevlar® webbing material. A left-side reinforcement strap 69 runs from the middle of the upper back strap 67 diagonally down to the point where the left-side strap 63 meets the waistband 64. FIG. 7 provides a view of the harness 61 with the front and rear portions of the suit 1 spread open. This figure shows the continuous construction of the right and left straps, 62 and 63. This figure shows how the right and left straps, 62 and 63, run from the waist strap 64 in the front, over the shoulders, past waist strap 64 in the rear, and down to the leg bands 65 and 66. FIGS. 8a and 8b show how high-stress points in the harness 61 are reinforced by double box X stitching. Thus, where the right and left straps, 62 and 63, meet the upper back reinforcement strap 67, double box X stitching 71 reinforces the junctions. Also, where the right and left diagonal reinforcement straps, 68 and 69, approach the upper back strap 67, double box X stitching reinforces the harness 61 at the back reinforcement panel 51. Also, where the right and left diagonal straps, 68 and 69, meet the upper back strap 67, double box X stitching reinforces the junction. Finally, it can be seen that, where the harness exits rear portion of the suit 1 through slot 73, double box X stitching fastens the right and left diagonal straps, 68 and 69, so that they form a harness loop 74 (seen in FIG. 9a). As noted above, this harness loop portion 74 is constructed of natural KEVLAR® webbing, as described above with respect to the tie-down web strap 35. Referring to FIG. 9a, it may be seen that the harness loop 74 exits the interior of the suit 1 through slot 73, located near the shoulder blades of the wearer. A back reinforcement panel 51 on the interior of the suit provides additional strength at the point where the harness 61 exits the slot 73. From FIG. 9b it will be seen that the back reinforcement panel 51 is an eight inch square piece. The slot 73 is a rectangle, two-and-a-half inches by three-quarters of an inch in dimension, which is sewn and topstitched, including topstitching around the edges. The slot 73 is located high at the rear of the torso 10 so that pulling forces from a lanyard 75 (not shown in FIG. 9a) are exerted to the harness loop 74 then to the harness 61, thereby pulling suspending the wearer from a point near the head. The harness 61 distributes forces to the thighs, waist and chest. If dragged by a lanyard 75, the wearer will be pulled head first, which is the most efficient manner in emergency operations for which the suit 1 is designed. The harness 61 and loop 74 arrangement prevents the wearer from being dragged or suspended sideways, which may injure the wearer and increase the likelihood that the wearer will get stuck when being pulled or dragged. FIG. 10 shows the connection of the harness loop 74 and lanyard 75. One end of the lanyard is threaded through the harness loop 74, then made to form its own loop 76 by folding it over and stitching it with a double box X stitch 71 to the lanyard 75. As described above, FIG. 11 shows how the lanyard 75 may be stored in the rear pouch 32 (not shown) by coiling 77 the lanyard. FIG. 13 shows the detachable Hood 91, which consists of double layered NOMEX® knit around the face opening 93, a three layer hood 92, described below, and a single layer NOMEX® drape 94 which covers the shoulders of the wearer. The hood 92 has three layers: an outer layer formed of NOMEX® knit, a middle layer formed of 3-layer E-89® Nomex®/Kevlar® quilted, and an inner layer of PBI® knit (all commercially available from DuPont). It has been discovered that a separate hood has advantages of suits with integral hoods. Most significantly, the separate hood prevents the build up of flammable and explosive gases inside the suit 1, thereby decreasing the risk of ignition within the suit. The separate hood 91 may be used with convention face protection and is easily put on and taken off. FIG. 14 shows a table of test results 101 for the flash suit 1 of the disclosed invention, as well as test results for currently available flash suits and test results without protective clothing (nude burn, cotton clothing, and cotton/nylon blends of clothing). The table indicates the time, in seconds, the test lasted and the percentage of 2nd and 3rd degree burns, as well as the total percentage of burning. The test results showed that the flash suit construction of the disclosed invention lasted 8 seconds without any burning reaching the inside of the suit. This result surpassed all other commercially available flash suits. The drawings and description set forth here represent only some embodiments of the invention. After considering these, skilled persons will understand that there are many ways to make a flash suit according to the principles disclosed. The inventors contemplate that the use of alternative structures, materials, or manufacturing techniques, which result in a flash suit according to the principles disclosed, will be within the scope of the invention. |
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claims | 1. A scanning electron microscope comprising:an electron source;a magnetic field type objective lens for focusing a primary electron beam emitted from the electron source on a specimen and having a lens gap opening toward the specimen;a first deflector for deflecting the primary electron beam off-axis of the objective lens; anda second deflector, comprising an electrostatic deflector, for deflecting the primary electron beam, so that a deflecting force of the objective lens against the primary electron beam is canceled when the primary electron beam is deflected off-axis of the objective lens,wherein the second deflector creates an electric field that suppresses off-axis deviation of the primary electron beam being caused by a magnetic field created by the objective lens. 2. The scanning electron microscope according to claim 1, wherein the first deflector serves also as a scanning deflector for scanning the primary electron beam. 3. The scanning electron microscope according to claim 1, wherein the second deflector is an octupole deflector. 4. The scanning electron microscope according to claim 3, wherein:the octupole deflector has an insulating base plate provided with a primary electron beam passing aperture and insulating slits formed so as to extend radially from the electron beam passing aperture, andopposite surfaces of a part of the base plate around the electron beam passing aperture and side surface of the electron beam passing aperture and the insulating slits are coated with conductive films. 5. The scanning electron microscope according to claim 4, wherein:the insulating base pate has a conductive, cylindrical part formed around the primary electron beam passing aperture, andthe conductive, cylindrical part of the insulating base plate is inserted in a primary electron beam passing aperture of the objective lens. 6. The scanning electron microscope according to claim 3, wherein:the octupole deflector has a part inserted in a primary electron beam passing aperture of the objective lens, anda shielding electrode is disposed so as to screen partly a deflecting electric field created by the octupole deflector. 7. A scanning electron microscope comprising:an electron source;a magnetic field type objective lens for focusing a primary electron beam emitted from the electron source on a specimen and having a lens gap opening toward the specimen;a first deflector for deflecting the primary electron beam off-axis of the objective lens; anda second deflector, comprising an electrostatic deflector, for deflecting the primary electron beam, so that an aberration created by a deflecting force of the objective lens against the primary electron beam is cancelled when the primary electron beam is deflected off-axis of the objective lens,wherein the second deflector creates an electric field that suppresses off-axis deviation of the primary electron beam being caused by a magnetic field created by the objective lens. 8. The scanning electron microscope according to claim 7, wherein the first deflector serves also as a scanning deflector for scanning the primary electron beam. 9. The scanning electron microscope according to claim 7, wherein the second deflector is an octupole deflector. 10. The scanning electron microscope according to claim 9, wherein:the octupole deflector has an insulating base plate provided with a primary electron beam passing aperture and insulating slits formed so as to extend radially from the electron beam passing aperture, andopposite surfaces of a part of the base plate around the electron beam passing aperture and side surface of the electron beam passing aperture and the insulating slits are coated with conductive films. 11. The scanning electron microscope according to claim 10, wherein:the insulating base plate has a conductive, cylindrical part formed around the primary electron beam passing aperture, andthe conductive, cylindrical part of the insulating base plate is inserted in a primary electron beam passing aperture of the objective lens. 12. The scanning electron microscope according to claim 9, wherein:the octupole deflector has a part inserted in a primary electron beam passing aperture of the objective lens, anda shielding electrode is disposed so as to screen partly a deflecting electric field created by the octupole deflector. |
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051223341 | summary | FIELD OF THE INVENTION The present invention relates to zirconium-base alloys and structural components made thereof for use in nuclear reactors. BACKGROUND OF THE INVENTION Various zirconium alloys are used as structural components in the nuclear industry. The most commonly used alloys, Zircaloy-2 and Zircaloy-4, contain strong alpha stabilizers tin and oxygen, plus the beta stabilizers iron, chromium and nickel. These alloys are generally forged in the beta region, then solution treated at about 1065.degree. C. (1950.degree. F.) and water quenched. Subsequent hot working and heat treating is done in the alpha region (below 790.degree. C.) to preserve a fine, uniform distribution of intermetallic compounds which results from solution treating and quenching. Corrosion resistance in steam and hot water depends on the distribution of the intermetallic compounds. Another significant commercial zirconium alloy is Zr-2.5Nb. The mechanical and physical properties of Zr-2.5Nb are similar to those of the Zircaloys but the corrosion resistance is slightly inferior to that of the Zircaloys. In zirconium, the low-temperature alpha phase has a close-packed hexagonal crystal structure which transforms to a body-centered-cubic structure at about 870.degree. C. (1600.degree. F.). The transformation temperature is affected by even small amounts of impurities such as oxygen. Alpha-stabilizing elements raise the temperature of the allotropic alpha-to-beta transformation. The alpha-stabilizing elements include Al, Sb, Sn, Be, Pb, Hf, N, O and Cd. Beta-stabilizing elements lower the alpha-to-beta transformation temperature. Typical beta-stabilizers include Fe, Cr, Ni, Mo, Cu, Nb, Ta, V, Th, U, W, Ti, Mn, Co and Ag. Low-solubility intermetallic compound formers such as C, Si and P readily form stable intermetallic compounds and are relatively insensitive to heat treatment. In addition to being an alpha-stabilizing element, oxygen is also used for solid-solution strengthening of zirconium. The oxygen content of Kroll process sponge generally varies from about 500 to 2000 ppm depending on the number of purification steps and the effectiveness of each step. Crystal bar zirconium generally contains less than 100 ppm oxygen. For instance, Table 5.10 of The Metallurgy Of Zirconium, by B. Lustman et al., McGraw-Hill Book Co., Inc., 1955, sets forth a typical analysis of Westinghouse crystal-bar zirconium having 200 ppm oxygen, 200 ppm Fe, 30 ppm Si, 30 ppm Al, 40 ppm Hf, less that 0.5 ppm Cu, 10 ppm Ti, less than 50 ppm Ca, less than 10 ppm Mn, less than 10 ppm Mg, less than 10 ppm Pb, less than 10 ppm Mo, 30 ppm Ni, 30 ppm Cr, less than 10 ppm Sn, 10 ppm N, 20 ppm H and 100 ppm C and elements not detected included Ga, Co, W, Au, Ag, Ta, Cb, B, V, P, Bi, Cd, Y, Yb, In, Ir, As, Os, Lu and Na. Gallium is used predominantly in the electronics industry where it is combined with elements of Group III, IV or V of the periodic table to form semiconducting materials. Gallium in aluminum causes severe intergranular corrosion of the aluminum. Zirconium alloys are disclosed in U.S. Pat. Nos. 3,148,055; 4,584,030; 4,707,330; 4,717,434; 4,751,045; 4,778,648; 4,810,461; 4,863,679; 4,908,071; 4,938,920; 4,938,921; and 4,963,316. U.S. Pat. No. 4,659,545 discloses a zirconium-based nuclear fuel rod cladding. U.S. Pat. No. 3,777,346 discloses a tension band for suspending rotatable mechanisms of measuring instruments, the tension bands being composed of Ti, Zr and Hf alloys which may also contain up to 15 atomic percent of non-transition metals such as Al, Sn, In, Ga or Cu. SUMMARY OF THE INVENTION The present invention provides a zirconium-base alloy having improved creep strength, the alloy comprising Zr and an amount of Ga effective to improve creep strength of the alloy. The alloy can contain up to 1 wt. % Ga. For instance, Ga can be present in amounts up to 0.5 wt. % or up to 0.25 wt. % such as 0.1 to 0.25 wt. %. According to one aspect of the invention, Ga is present in amounts of 0.25-0.5 wt. %. The alloy can also contain oxygen. For instance, the oxygen content can be up to 0.5 wt. %. In particular, oxygen can be present in amounts of 0.1-0.25 wt. % or 0.12-0.18 wt. %. The alloy can also include Sn. For instance, Sn can be present in amounts of up to 1 wt. %. In particular, Sn can be present in amounts of 0.1-0.7 wt. % or 0.25-0.5 wt. %. The alloy can also contain at least one of Fe, Cr and V. For instance, the total amount of Fe, Cr and V can be up to 1 wt. %. In particular, Fe can be present in amounts of up to 0.5 wt. % such as 0.1-0.5 wt. % or 0.25-0.4 wt. %. Cr can be present in amounts of up to 0.5 wt. % such as 0.1-0.5 wt. % or 0.15-0.25 wt. %. V can be present in amounts of up to 0.5 wt. % such as 0.15-0.4 wt. % or 0.2-0.3 wt. %. According to one aspect of the invention, the alloy consists essentially of 0.25-0.5 wt. % Ga, 0.1-0.25 wt. % oxygen, 0.1-0.7 wt. % Sn, 0.1-0.5 wt. % Fe, 0.15-0.4 wt. % V, 0-0.5 wt. % Cr, balance Zr and unavoidable impurities. For instance, the oxygen can be present in amounts of 0.12-0.18 wt. %, Sn can be present in amounts of 0.25-0.5 wt. %, Fe can be present in amounts of 0.25-0.4 wt. %, V can be present in amounts of 0.2-0.3 wt. % and Cr can be present in amounts of 0.15-0.25 wt. %. In accordance with another aspect of the invention, a structural component for use in nuclear reactors is provided wherein the component comprises a zirconium-base alloy including an amount of Ga effective to improve creep strength of the alloy. The alloy can include up to 1 wt. % Ga, up to 0.5 wt. % oxygen, up to 1 wt. % Sn and up to 1 wt. % in total of Fe, Cr and V. More particularly, the alloy can consist essentially of 0.25-0.5 wt. % Ga, 0.1-0.25 wt. % oxygen, 0.1-0.7 wt. % Sn, 0.1-0.5 wt. % Fe, 0.15-0.4 wt. % V, 0-0.5 wt. % Cr, balance Zr and unavoidable impurities. The oxygen can be present in amounts of 0.12-0.18 wt. %, Sn can be present in amounts of 0.25-0.5 wt. %, Fe can be present in amounts of 0.25-0.4 wt. %, V can be present in amounts of 0.2-0.3 wt. % and Cr can be present in amounts of 0.15-0.25 wt. %. The structural component can comprise a component of a fuel assembly such as a fuel tube, spacers, springs, etc. In the case of a fuel tube, the alloy can be used with a Ga-free inner liner, such as a zirconium liner, or the tube can be liner-free in which case the entire tube can be of the Zr-Ga alloy. |
claims | 1. A fuel assembly comprising a plurality of first fuel rods and a plurality of second fuel rods having a length shorter than a length of the plurality of first fuel rods, said first and said second fuel rods being arranged in a fuel rod array of 10 rows by 10 columns; and a plurality of water rods occupying regions allowing arrangement of 8 fuel rods, said second fuel rods being not arranged in an outermost tier of the fuel rod array, which satisfies the following conditions: B xe2x89xa760, 15 xe2x89xa6n xe2x89xa620( n : integer), Awr/Ach xe2x89xa60.149, Lp/Lf xe2x89xa711/24, Awr/Ach xe2x89xa7(3.00xc3x9710 xe2x88x924 xc3x97n 2 +6.00xc3x9710 xe2x88x924 xc3x97n xe2x88x921.2xc3x9710 xe2x88x922 )xc3x97( Lp/Lf xe2x88x921)+1.75xc3x9710 xe2x88x921 , and Awr/Ach xe2x89xa6(8.63xc3x9710 xe2x88x924 xc3x97n 2 xe2x88x926.09xc3x9710 xe2x88x922 xc3x97n +1.33xc3x9710 1 )xc3x97( Lp/Lf xe2x88x928.32xc3x9710 xe2x88x921 ) where Awr is a total sum of horizontal sectional areas of said water rods, Ach is a horizontal sectional area of a coolant flow passage in a bottom portion of said fuel assembly, Lf is an effective fuel length of said first fuel rods, n is the number of said second fuel rods, Lp is an effective fuel length of said second fuel rods, and B (GWd/t) is an average unloading burn-up. 2. A fuel assembly according to claim 1 , wherein some of said second fuel rods are arranged adjacent to said water rods and some of said second fuel rods are arranged in a second tier from the outer side in the fuel rod array. claim 1 3. A fuel assembly according to claim 1 , wherein each of said water rods has a horizontal sectional area arranged in a region allowing arrangement of four fuel rods, and the number of said water rods is two. claim 1 |
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description | This invention relates, in general, to perforating a wellbore that traverses a fluid bearing subterranean formation using shaped charges and, in particular, to an apparatus and method for dynamically adjusting the center of gravity of a perforating apparatus. Without limiting the scope of the present invention, its background will be described with reference to perforating a subterranean formation with a shaped charge perforating apparatus, as an example. After drilling the various sections of a subterranean wellbore that traverses a formation, individual lengths of relatively large diameter metal tubulars are typically secured together to form a casing string that is positioned within the wellbore. This casing string increases the integrity of the wellbore and provides a path for producing fluids from the producing intervals to the surface. Conventionally, the casing string is cemented within the wellbore. To produce fluids into the casing string, hydraulic opening or perforation must be made through the casing string, the cement and a short distance into the formation. Typically, these perforations are created by detonating a series of shaped charges located within the casing string that are positioned adjacent to the desired formation. Specifically, one or more charge carriers are loaded with shaped charges that are connected with a detonating device, such as detonating cord. The charge carriers are then connected within a tool string that is lowered into the cased wellbore at the end of a tubing string, wireline, slick line, coil tubing or the like. Once the charge carriers are properly positioned in the wellbore such that the shaped charges are adjacent to the formation to be perforated, the shaped charges are detonated. Upon detonation, the shaped charges create jets that blast through scallops or recesses in the carrier. Each jet creates a hydraulic opening through the casing and the cement and enters the formation forming a perforation. It has been found, however, that it is sometimes desirable to perforate a wellbore in a particular direction or range of directions relative to the wellbore. For example, in a deviated, inclined or horizontal well, it is frequently beneficial to form perforations in the upward direction, the downward direction or both. Attempts have been made to achieve this goal of perforating wells in particular directions. One method of orienting perforating charges downhole requires the charges to be rigidly mounted in a gun carrier so that they are pointed in the desired directions relative to the carrier. The gun carrier is then conveyed into a wellbore and either laterally biased physically to one side of the wellbore so that the gun carrier seeks the lower portion of the wellbore due to gravity, or the gun carrier is rotatably supported with its center of gravity laterally offset relative to the wellbore. This method relies on the gun carrier rotating in the wellbore, so that the gun carrier may be oriented relative to the force of gravity. Frequently, such orienting rotation is unreliable due to friction between the gun carrier and the wellbore, debris in the wellbore or the like. More recently, the assignee of the present invention has developed a perforating gun that includes a tubular gun carrier, multiple perforating charges, multiple charge mounting structures and multiple rotating supports. This internally oriented perforating apparatus has successfully provided increased reliability in orienting perforating charges to shoot in the desired directions in a well. In this design, the direction or directions of the perforations is established when the gun is assembly in its manufacturing facility. It has been found, however, that in certain installations, it is necessary to avoid shooting in a particular direction or directions. For example, one or more communication conduits or controls lines may extend along the exterior of the casing string. During installation, these conduits commonly become wound around the casing string such that the exact location of these lines can only determined after installation by, for example, logging the well. A need has therefore arisen for an apparatus and method operable to achieve reliable downhole orientation of the shaped charges in a perforating apparatus such that the shaped charges shoot in desired directions. In addition, a need has arisen for such an apparatus and method operable to achieve reliable downhole orientation of the shaped charges in a perforating apparatus such that the shaped charges do not shoot in undesired directions. The present invention disclosed herein comprises an apparatus and method for dynamically adjusting the center of gravity of a perforating apparatus. The apparatus and method of the present invention are operable to achieve reliable downhole orientation of shaped charges in a perforating apparatus such that the shaped charges shoot in desired directions. In addition, apparatus and method of the present invention are operable to achieve reliable downhole orientation of shaped charges in a perforating apparatus such that the shaped charges do not shoot in undesired directions In one aspect, the present invention is directed to a perforating apparatus used to perforate a subterranean well. The perforating apparatus includes a generally tubular gun carrier having a charge holder rotatably mounted therein. At least one shaped charge is mounted in the charge holder and is operable to perforate the well upon detonation. A dynamically adjustable weight system is operably associated to the charge holder. The dynamically adjustable weight system is operable to adjust the center of gravity of the charge holder such that gravity will cause the charge holder to rotate within the gun carrier to position the at least one shaped charge in a desire circumferential direction relative to the well prior to perforating. In one embodiment, the dynamically adjustable weight system includes a plurality of discrete weights that are individually coupled to the charge holder at a plurality of longitudinal locations. In this embodiment, for each of the longitudinal locations, the charge holder may include a plurality of circumferentially distributed openings such as uniformly distributed openings at between about 15 and 60 degree increments. Alternatively, for each of the longitudinal locations, the charge holder may include a circumferentially extending slot that may extend circumferentially between about 90 and 180 degrees. In another embodiment, the dynamically adjustable weight system includes a plurality of longitudinally extending tubes operable to contain a weighted material therein. In a further embodiment, the dynamically adjustable weight system includes weights formed from a malleable material. In yet another embodiment, the dynamically adjustable weight system includes a weight tube that is rotatable relative to the charge holder. In any of these embodiments, the at least one shaped charge may include a plurality of shaped charges that may be positioned in the charge holder to fire in substantially the same circumferential direction or the shaped charges may be positioned in the charge holder to fire in multiple circumferential directions. In another aspect, the present invention is directed to a perforating apparatus used to perforate a subterranean well. The perforating apparatus includes a generally tubular gun carrier having a charge tube rotatably mounted therein. The charge tube includes a plurality of circumferentially extending slots. At least one shaped charge is mounted in the charge tube and is operable to perforate the well upon detonation. A dynamically adjustable weight system is coupled to the charge tube. The dynamically adjustable weight system includes a plurality of discrete weights that are coupled to the charge tube at the slots such that the circumferential location of the weights is adjustable along the length of the slots to adjust the center of gravity of the charge tube such that gravity will cause the charge tube to rotate within the gun carrier to position the at least one shaped charge in a desired circumferential direction relative to the well prior to perforating. In one embodiment, adjacent slots in the charge tube extend in circumferentially opposite directions. In another embodiment, the weights are attached to the charge tube using bolts that are selectively slidable within the slots. In another aspect, the present invention is directed to a method of perforating a subterranean well. The method includes identifying at least one undesired circumferential direction associated with a perforating interval in the well; adjusting components of a dynamically adjustable weight system to change the center of gravity of a charge holder rotatably mounted within a gun carrier; positioning the gun carrier within the perforating interval in the well; gravitationally aligning a least one shaped charge mounted in the charge holder in at least one desired circumferential direction relative to the well that does not correspond with the at least one undesired circumferential direction; and firing the at least one shaped charge to perforate the well in the at least one desired circumferential direction. The method may also include relocating discrete weights circumferentially about the charge holder. This may be accomplished by relocating the discrete weights relative to circumferentially distributed openings in the charge holder or relocating the discrete weights relative to circumferentially extending slots in the charge holder. Alternatively, the method may include changing the amount of weighted material in at least one longitudinally extending tube, reshaping malleable material disposed within the charge holder or rotating a weight tube relative to the charge holder. While the making and using of various embodiments of the present invention are discussed in detail below, it should be appreciated that the present invention provides many applicable inventive concepts which can be embodied in a wide variety of specific contexts. The specific embodiments discussed herein are merely illustrative of specific ways to make and use the invention, and do not delimit the scope of the present invention. Referring initially to FIG. 1, a plurality of apparatuses for dynamically adjusting the center of gravity of perforating apparatuses operating from an offshore oil and gas platform are schematically illustrated and generally designated 10. A semi-submersible platform 12 is centered over a submerged oil and gas formation 14 located below sea floor 16. A subsea conduit 18 extends from deck 20 of platform 12 to wellhead installation 22 including subsea blow-out preventers 24. Platform 12 has a hoisting apparatus 26 and a derrick 28 for raising and lowering pipe strings such as work sting 30. A wellbore 32 extends through the various earth strata including formation 14. A casing 34 is cemented within wellbore 32 by cement 36. Work string 30 includes various tools such as a plurality of perforating apparatuses or guns 38. When it is desired to perforate casing 34, work string 30 is lowered through casing 34 until the perforating guns 38 are properly positioned relative to formation 14. Thereafter, the shaped charges within the string of perforating guns 38 are sequentially fired, either in an uphole to downhole or a downhole to uphole direction. Upon detonation, the liners of the shaped charges form jets that create a spaced series of perforations extending outwardly through casing 34, cement 36 and into formation 14, thereby allow fluid communication between formation 14 and wellbore 32. In the illustrated embodiment, wellbore 32 has an initial, generally vertical portion 40 and a lower, generally deviated portion 42 which is illustrated as being horizontal. It should be noted, however, by those skilled in the art that the apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention is equally well-suited for use in other well configurations including, but not limited to, inclined wells, wells with restrictions, non-deviated wells, multilateral wells and the like. In addition, even though an offshore operation has been depicted in FIG. 1, the apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention is equally well-suited for use in onshore operations. Work string 30 includes a packer 44 that may be sealingly engaged with casing 34 and is illustrated in the vertical portion 40 of wellbore 32. At the lower end of work string 30 is the gun string including the plurality of perforating guns 38, a ported nipple 46 and a fire head 48. In the illustrated embodiment, perforating guns 38 include internal orientation features which allow for reliable rotation of the charge tube within the gun carrier as described in U.S. Pat. No. 6,595,290 issued to Halliburton Energy Services, Inc. on Jul. 22, 2003, which is hereby incorporated by reference for all purposes. Referring now to FIG. 2, therein is depicted a perforating apparatus that includes an apparatus for dynamically adjusting the center of gravity of the perforating apparatus of the present invention that is generally designated 100. In the following description of apparatus 100 as well as the other apparatuses and methods described herein, directional terms such as “above”, “below”, “upper”, “lower” and the like are used for convenience in referring to the illustrations as it is to be understood that the various embodiments of the invention may be used in various orientations such as inclined, inverted, horizontal, vertical and the like and in various configurations, without departing from the principles of the invention. Gun 100 includes a plurality of shaped charges 102 that are securably mounted in a charge holder that is depicted as charge tube 104. Charge tube 104 is rotatably mounted within gun carrier 106. Preferably, charge tube 104 is made from cylindrical tubing, but it should be understood that it is not necessary for charge tube 104 to be tubular or have a cylindrical shape in keeping with the principles of the invention. Charge tube 104 includes multiple supports 108 that allow charge tube 104 to rotate within gun carrier 106. This manner of rotatably supporting charge tube 104 prevents charges 102 or any other portion of charge tube 104 from contacting the interior of gun carrier 106. Each of the supports 108 includes rolling elements or bearings 110 contacting the interior of gun carrier 106. For example, bearings 110 could be ball bearings, roller bearings, plain bearings or the like. Bearings 110 enable supports 108 to suspend charge tube 104 in gun carrier 106 and permit rotation thereof. In addition, optional thrust bearings 112 may be positioned between each end of charge tube 104 and gun carrier 106 such that thrust bearings 112 contact devices 114 attached at each end of gun carrier 106. Each device 114 may be tandems that are used to couple two guns to each other, a bull plug used to terminate a gun string, a firing head, or any other type of device which may be attached to gun carrier 106. As with bearings 110 described above, thrust bearings 112 may be any type of bearings. Thrust bearings 112 support charge tube 104 against axial loading within gun carrier 106, while permitting charge tube 104 to rotate within gun carrier 106. Charge tube 104, charges 102 and other portions of gun 100 supported in gun carrier 106 by the supports 108 including, for example, a detonating cord 116 extending to each of the charges and portions of the supports themselves, are parts of an overall rotating assembly 118. By offsetting a center of gravity 120 of assembly 118 relative to a longitudinal rotational axis 122 of bearings 110, assembly 118 is biased by gravity to rotate to a specific position in which the center of gravity 120 is located directly below the rotational axis 122. Assembly 118 may, due the construction of the various elements thereof, initially have the center of gravity 120 in a desired position relative to charges 102. However, to ensure that charges 102 are directed to shoot in respective predetermined directions, the center of gravity 120 may be repositioned using a dynamically adjustable weight system that is depicted as weights 124. In the illustrated embodiment, on the left side of FIG. 2, weights 124 are added to assembly 118 to direct the charges 102 to shoot upward, while on the right side of FIG. 2, weights 124 are added to assembly 118 to direct the charges 102 to shoot downward. As discussed in greater detail below, weights 124 may be otherwise positioned to direct the charges 102 to shoot in any desired direction, or combination of directions and to avoid shooting in undesired directions. Gun carrier 106 is provided with reduced wall thickness portions 126, which extend circumferentially about carrier 106 outwardly overlying each of the charges 102. Thus, as the charges 102 rotate within carrier 106, they remain directed to shoot through the portions 126. The reduced wall thickness portions 126 may be formed on carrier 106 by rolling, forging, lathe cutting or any other suitable technique. Referring next to FIGS. 3A and 3B, therein are depicted side and cross sectional views of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 130. Apparatus 130 includes a charge holder depicted as charge tube 132 which houses a plurality of shaped charges 134. In the illustrated embodiment, shaped charges 134 are configured in a 180 degree phased pattern, however, those skilled in the art will appreciate that any number of alternative phased patterns of the shaped charges are possible and are considered within the scope of the present invention. Apparatus 130 also includes a dynamically adjustable weight system depicted as weights 136. In the illustrated embodiment, each of the weights 136 includes a threaded portion that is operable to receive therein a complementary threaded bolt 138. Weights 136 are accordingly attached to charge tube 132 by passing the shaft portion of a bolt 138 through one of a plurality of openings 140 in charge tube 132 and then rotatably coupling that bolt 138 to one of the weights 136. As illustrated, each longitudinal location of charge tube 132 that is designed to receive a weight 136 has eight openings 140 that are circumferentially spaced apart at 45 degree increments. It should be understood by those skilled in the art, however, that any number of openings having any desired circumferentially spacing both uniform and nonuniform is possible and is considered within the scope of the present invention, so long as the structural integrity of charge tube 132 is maintained. For example, it may be desirable to have openings that are circumferentially spaced uniformly around a charge tube at between about 15 and about 60 degree increments. As used herein, the term dynamically adjustable refers to the ability to change the center of gravity of a perforating apparatus in the field as opposed to only as the perforating apparatus is manufactured. This ability provides the versatility to make adjustments to apparatus 130 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line disposed to the exterior of the casing string. Continuing with this example, if one or more control lines are position to the exterior of the casing string, it is imperative to avoid causing damage to the control lines during the perforating process. As these control lines commonly take on a spiral configuration around the casing string during installation, the actual location of the control lines must be determined prior to perforating the well by, for example, logging the well. Once the circumferential location of the control lines is known for each depth of the well, the present invention allows field personnel to custom design the perforating gun string such that the control lines can be avoided and the well can be perforated in the desired directional orientations. In the illustrated embodiment, this is accomplished by repositioning the weights 136 relative to any one of the respective openings 140 circumferentially spaced around charge tube 132. For example, if charge tube 132 were installed within a gun carrier as configured in FIG. 3B and deployed in a horizontal well, weights 136 would cause charge tube 132 to rotate to the position depicted in FIG. 3B wherein shaped charges 134 would fire at 0 and 180 degrees in the well. If weights 136 were each moved to the next adjacent position, shaped charges 134 would fire at 45 and 225 degrees in the well. Likewise, if weights 136 were each moved again to the next adjacent position, shaped charges 134 would fire at 90 and 270 degrees in the well. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Even though FIGS. 3A-3B have depicted apparatus 130 as having one weight positioned between adjacent shaped charge, it should be understood by those skilled in the art that no particular relationship is required between the number of weights and the number of shaped charges in a given perforating apparatus. The number and configuration of the weights and shaped charges will vary based upon factors such as the desired shots per foot, the diameter of the charge tube, the explosive mass of the charges, the size of the weights, the spacing between charges and the like. The important factor is that the center of gravity is dynamically adjustable to cause the charge tube to rotate within the gun carrier to the desired position. Referring next to FIGS. 4A and 4B, therein are depicted side and cross sectional views of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 150. Apparatus 150 includes a charge holder depicted as charge tube 152 which houses a plurality of shaped charges 154. In the illustrated embodiment, shaped charges 154 are configured in a 180 degree phased pattern, however, those skilled in the art will appreciate that any number of alternative phased patterns of the shaped charges are possible and are considered within the scope of the present invention. Apparatus 150 also includes a dynamically adjustable weight system depicted as weights 156. In the illustrated embodiment, each of the weights 156 includes a threaded portion that is operable to receive therein a complementary threaded bolt 158. Weights 156 are accordingly attached to charge tube 152 by passing the shaft portion of a bolt 158 through a slot 160 in charge tube 152 and then rotatably coupling that bolt 158 to one of the weights 156. As illustrated, each longitudinal location of charge tube 152 that is designed to receive a weight 156 has a slot 160 that circumferentially traverses 180 degrees of charge tube 152. Adjacent slots 160 of apparatus 150 are configured such that they extend on opposite sides of charge tube 152. This design enhances the structural integrity of charge tube 152 and allows for infinite variability in the center of gravity of apparatus 150. In certain implementations weights 156 may be placed in each of the slots 160. In other implementations, it may be desirable to have weights 156 in every other slot 160 such that each of the weights 156 can be positioned at the same circumferential position. It should be understood by those skilled in the art that slots 160 could have other circumferential orientations and could have other relative spacing arrangement, both uniform and nonuniform, without departing from the principles of the present invention, so long as the structural integrity of charge tube 152 is maintained. As discussed above, the combination of slots 160 and weights 156 allow for dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 150 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by circumferentially repositioning the weights 156 along slots 160 by loosening bolts 158, sliding the weights 156 to the desired circumferential position and resecuring the weights 156 to charge tube 152 with the bolts 158. If charge tube 152 were installed within a gun carrier as loaded in FIG. 4B and deployed in a horizontal well, weights 156 would cause charge tube 152 to rotate to the position depicted in FIG. 4B wherein shaped charges 154 would fire at 0 and 180 degrees in the well. Repositioning of the weights 156 along slots 160, as described above, would allow for firing in any desired circumferential directions. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Referring next to FIG. 5, therein is depicted a cross sectional view of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 170. Apparatus 170 includes a charge holder depicted as charge tube 172 which houses a plurality of shaped charges (not pictured). Apparatus 170 also includes a dynamically adjustable weight system 174 that is depicted a plurality of tubes 176. Tubes 176 extend at least partially longitudinally within charge tube 172 and are operable to contain a weighted material such as a fluid or a solid. As illustrated, apparatus 170 includes seven tubular tubes 176 that are circumferentially distributed within charge tube 172 at 30 degree increments. It should be understood by those skilled in the art that tubes 176 could have other circumferential orientations, both uniform and nonuniform, within charge tube 172 without departing from the principles of the present invention. Likewise, even though tubes 176 are depicted as having a tubular cross section, tubes 176 could alternatively have other cross sections including, but not limited to, oval cross sections, rectangular cross sections, arc shaped cross sections and the like. In addition, those skilled in the art will recognize that not all of tubes 176 need to have the same cross section or be of the same size. In operation, dynamically adjustable weight system 174 of apparatus 170 allows field personnel to make dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 170 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by adding or reducing the weight within tubes 176 by, for example, adding or removing a fluid such as water from tubes 176. As the weight is adjusted in the various tubes 176, the desired downhole rotation of charge tube 172 can be achieved. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Referring next to FIG. 6, therein is depicted a cross sectional view of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 180. Apparatus 180 includes a charge holder depicted as charge tube 182 which houses a plurality of shaped charges (not pictured). Apparatus 180 also includes a dynamically adjustable weight system 184 that is depicted a plurality of tubes 186. Tubes 186 extend at least partially longitudinally along the exterior of charge tube 182 and are operable to contain a weighted material such as a fluid or a solid. As illustrated, apparatus 180 includes seven tubular tubes 186 that are circumferentially distributed within charge tube 182 at 30 degree increments. It should be understood by those skilled in the art that tubes 186 could have other circumferential orientations, both uniform and nonuniform, within charge tube 182 without departing from the principles of the present invention. Likewise, even though tubes 186 are depicted as having a tubular cross section, tubes 186 could alternatively have other cross sections including, but not limited to, oval cross sections, rectangular cross sections, arc shaped cross sections and the like. In addition, those skilled in the art will recognize that not all of tubes 186 need to have the same cross section or be of the same size. In operation, dynamically adjustable weight system 184 of apparatus 180 allows field personnel to make dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 180 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by adding or reducing the weight within tubes 186 by, for example, adding or removing a fluid such as water from tubes 186. As the weight is adjusted in the various tubes 186, the desired downhole rotation of charge tube 182 can be achieved. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Even though FIGS. 5 and 6 have depicted tubes located respectively inside and outside of a charge tube that are operable to receive a weighted material therein, those skilled in the art should recognize that alternate configurations could also be used and would be considered within the scope of the present invention including, but not limited to, forming one or more passageways in the wall of the charge tube or similar tubular operable to receive a weighted material therein. Referring next to FIGS. 7A and 7B, therein is depicted cross sectional views of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 190. Apparatus 190 includes a charge holder depicted as charge tube 192 which houses a plurality of shaped charges (not pictured). Apparatus 190 also includes a dynamically adjustable weight system 194 that is depicted as malleable weight members 196 that may be formed from a metal such as lead or a polymer. Malleable weight members 196 may extend at least partially longitudinally along the interior of charge tube 192 or may be discrete weight elements similar to weights 136 and 156 described above. As illustrated, each malleable weight member 196 is coupled to charge tube 192 using one or more bolts 198. In operation, dynamically adjustable weight system 194 of apparatus 190 allows field personnel to make dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 190 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by applying pressure or force to the malleable material that forms malleable weight members 196 using, for example, an adjustment tool that is sized to extend into charge tube 192. The location of at least a portion of the mass of malleable weight members 196 can them be adjusted, as seen in a comparison of FIGS. 7A and 7B, such that the desired downhole rotation of charge tube 192 can be achieved. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Referring next to FIGS. 8A-8G, therein are depicted various views of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 200. When assembled, apparatus 200 forms a rotating assembly 202 that is rotatably mounted in a gun carrier in a manner described above. Apparatus 200 includes a charge holder 204 that supports a plurality of shaped charges 206. Charge holder 204 is coupled to end plates 208. Each end plate 208 includes a plurality of notches 210 that are illustrated as being positioned circumferentially around end plates 208 at 60 degree increments, however, those skilled in the art will recognize that notches 210 could have alternate configurations including having different circumferential spacing. In addition, depending upon the length of charge holder 204, it may be desirable to have addition structures that are similar to end plates 208 positioned at intermediate locations along charge holder 204 between certain shaped charges 206. Apparatus 200 also includes a dynamically adjustable weight system depicted as weight tube 212. Weight tube 212 is formed from a substantially tubular member having a window 214, as best seen in FIG. 8E. In the illustrated embodiment, window 214 extends about 120 degrees circumferentially around weight tube 212, however, those skilled in the art will recognize that window 214 could have alternate configurations including having a different circumferential width or multiple window sections circumferential distributed around weight tube 212. Weight tube 212 includes circumferential end sections 216 that are sized to closely receive end plates 208. Weight tube 212 includes a plurality of rails 218 that are designed to mesh with notches 210 of end plates 208. In operation, the dynamically adjustable weight system of apparatus 200 allows field personnel to make dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 200 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by inserting charge holder 204 into weight tube 212 such that shaped charges 206 are oriented in the desired direction. For example, if charge holder 204 were installed within weight tube 212 as shown in FIG. 8F and deployed in a horizontal well, weight tube 212 would cause rotating assembly 202 to rotate to the position depicted in FIG. 8F wherein shaped charges 206 would fire at 0 degrees in the well. If charge holder 204 was rotated 60 degrees in either direction to realign rails 218 and notches 210, shaped charges 206 would fire at either 60 degrees or 300 degrees in the well. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Referring next to FIGS. 9A-9B, therein are depicted side and top views of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 220. When assembled, apparatus 220 forms a rotating assembly 222 that is rotatably mounted in a gun carrier in a manner described above via bearings 224. Apparatus 220 includes a charge holder 226 that supports a plurality of shaped charges 228. Apparatus 220 also includes a dynamically adjustable weight system depicted as weight tube 230. Weight tube 230 is formed from a partially tubular member. Charge holder 226 is selectively rotatable mounted within weight tube 230 such that charge holder 226 may be rotated about 120 degrees circumferentially within weight tube 230. In operation, the dynamically adjustable weight system of apparatus 220 allows field personnel to make dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 220 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by selectively releasing a connection such as a pin, a set screw or the like between charge holder 226 and weight tube 230 then rotating charge holder 226 such that shaped charges 228 are oriented in the desired direction. For example, if charge holder 226 was installed within weight tube 230 as shown in FIG. 9A and deployed in a horizontal well, weight tube 230 would cause rotating assembly 222 to rotate to the position depicted in FIG. 9A wherein shaped charges 228 would fire at 0 degrees in the well. If another circumferential direction is desired, however, charge holder 226 may be incrementally adjusted in certain embodiments or infinitely adjusted in other embodiments to any position between the locations of maximum travel which have been described above as approximately 60 degrees from vertical in either direction in the illustrated embodiment. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. Referring next to FIGS. 10A-10C, therein are depicted various views of an apparatus for dynamically adjusting the center of gravity of a perforating apparatus of the present invention that is generally designated 240. Apparatus 240 includes a charge holder depicted as a charge tube 242 that is rotatably mounted in a gun carrier in a manner described above via bearings 244, as best seen in FIGS. 10A and 10C. Charge tube 242 supports a plurality of shaped charges 246. Apparatus 240 also includes a dynamically adjustable weight system depicted as weight tube 250, as best seen in FIGS. 10B and 10C. Weight tube 250 is formed from a partially tubular member. Weight tube 250 is rotatable mounted within a swivel member 252 that is mounted within charge tube 242 such that weight tube 250 may be rotated about 120 degrees circumferentially within charge tube 242. One or more coupling members depicted as pins 254 are used to selectively prevent rotation of weight tube 250 relative to swivel member 252. In operation, the dynamically adjustable weight system of apparatus 240 allows field personnel to make dynamic adjustments in the center of gravity of a perforating apparatus in the field. This ability provides the versatility to make adjustments to apparatus 240 that will not only allow the field personnel to shoot in a desired direction but also prevent shooting in an undesired direction, such as in the direction of a control line or other hazard disposed to the exterior of the casing string or within the casing string. Specifically, in the illustrated embodiment, this is accomplished by selectively releasing the connection, such as pins 254, between weight tube 250 and swivel member 252 then rotating weight tube 250 relative to swivel member 252 such that weight tube 250 is positioned in the desired orientation relative to shaped charges 246. For example, if weight tube 250 was installed relative to shaped charges 246 as shown in FIGS. 10B-10C and deployed in a horizontal well, weight tube 250 would cause charge tube 242 to rotate to the position depicted in FIGS. 10B-10C wherein shaped charges 246 would fire at 0 degrees in the well. If another circumferential direction is desired, however, weight tube 250 may be incrementally adjusted in certain embodiments or infinitely adjusted in other embodiments to any position between the locations of maximum travel which have been described above as approximately 60 degrees from vertical in either direction in the illustrated embodiment. Accordingly, the directions the shaped charges will perforate the well may be dynamically adjusted by field personnel after the location of any wellbore hazards has been determined. While this invention has been described with reference to illustrative embodiments, this description is not intended to be construed in a limiting sense. Various modifications and combinations of the illustrative embodiments as well as other embodiments of the invention will be apparent to persons skilled in the art upon reference to the description. It is, therefore, intended that the appended claims encompass any such modifications or embodiments. |
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claims | 1. An apparatus to examine a target, comprising:an x-ray source configured to deliver a first x-ray beam towards the target;a device having an array of openings, wherein at least two of the openings are next to each other in a side-by-side configuration, the device located at an angle less than 180 degrees relative to a beam path of the first x-ray beam to receive a second x-ray beam resulted from an interaction between the first x-ray beam and the target; anda detector aligned with the device, the detector located at an angle less than 180 degrees relative to the beam path of the first x-ray beam to receive the second x-ray beam from the device after the second x-ray beam has exited through one or more of the openings of the device;wherein the apparatus further comprises a processing unit configured to:determine two or more peaks based on data received from the detector, the two or more peaks selected from the group consisting of a first k-alpha peak, a first k-beta peak, a first Compton scatter peak, a second k-alpha peak, a second k-beta peak, and a second Compton scatter peak, andcalculate a parameter indicating a cancer characteristic, wherein the processing unit is configured to calculate the parameter by computing a ratio between two of the two or more peaks. 2. The apparatus of claim 1, wherein the detector comprises a first detector element and a second detector element, the first detector element corresponding with a first one of the openings, the second detector element corresponding with a second one of the openings; andwherein the first detector element is configured to provide spectrum data in response to the first beam interacting with a first part of the target, and the second detector element is configured to provide second spectrum data in response to the first beam interacting with a second a second part of the target. 3. The apparatus of claim 1, wherein the processing unit configured to determine examination data based on signals generated by the detector, the examination data representing fluid build up, fluid retention, fluid wash out, rate of fluid build up, or rate of fluid wash out. 4. The apparatus of claim 1, further comprising an image re-constructor for reconstructing a volumetric image of the target using signals generated by the detector. 5. The apparatus of claim 1, wherein the angle is a value that is anywhere between 30 degrees and 170 degrees. 6. The apparatus of claim 1, wherein the second beam comprises a photo absorption component and a Compton scatter component. 7. The apparatus of claim 1, further comprising a collimator between the x-ray source and the target. 8. The apparatus of claim 1, wherein the detector comprises a photo spectrum sensitive detector. 9. The apparatus of claim 1, wherein each of the openings has an elongated shape. 10. The apparatus of claim 1, wherein the openings are arranged in a plurality of rows. 11. The apparatus of claim 1, wherein the device comprises a plurality of spaced apart plates that define the array of openings. 12. The apparatus of claim 1, wherein the first x-ray beam comprises a cone beam. 13. The apparatus of claim 1, wherein the first x-ray beam comprises a first fan beam for examining parts of the target that are in a first plane. 14. The apparatus of claim 13, wherein the x-ray source is configured to provide a second fan beam for examining parts of the target that are in a second plane. 15. The apparatus of claim 1, wherein the first x-ray beam comprises a first pencil beam for examining parts of the target that are along a first line. 16. The apparatus of claim 15, wherein the x-ray source is configured to provide a second pencil beam for examining parts of the target that are along a second line, the first line and the second line defining a plane. 17. The apparatus of claim 16, wherein the x-ray source is configured to provide a third pencil beam for examining parts of the target that are along a third line. 18. The apparatus of claim 1, wherein the parameter comprises information regarding a density of agent attached to a cancer site, information regarding angiogenesis, information regarding hypoxia, information regarding pathological functional marker(s), information regarding an identifier of cancer malignancy, information regarding an identifier of cancer growth, or information regarding an identifier of cancer growth rate. 19. The apparatus of claim 1, wherein the second x-ray beam has a first beam portion and a second beam portion travelling in different respective directions, and wherein the device is configured to allow the first beam portion to go through the one or more of the openings, and block the second beam portion. 20. The apparatus of claim 1, wherein the angle is approximately 90 degrees. 21. The apparatus of claim 1, wherein the angle is anywhere from 30 degrees to 100 degrees. 22. The apparatus of claim 1, wherein the processing unit is configured to calculate the parameter using an intensity of the first k-alpha peak, an intensity of the first k-beta peak, an intensity of the first Compton scatter peak, or two or more of the foregoing. 23. The apparatus of claim 1, wherein the processing unit is configured to calculate the parameter using an area under the first k-alpha peak, an area under the first k-beta peak, an area under the first Compton scatter peak, or two or more of the foregoing. 24. The apparatus of claim 1, wherein the processing unit is also configured to determine information regarding radiation treatment effectiveness, information regarding a progress of radiation treatment, or information regarding lack of a progress of radiation treatment. 25. An apparatus to examine a target, comprising:an x-ray source configured to deliver a first x-ray beam towards the target;a device having an array of openings, wherein at least two of the openings are next to each other in a side-by-side configuration, the device located at an angle less than 180 degrees relative to a beam path of the first x-ray beam to receive a second x-ray beam resulted from an interaction between the first x-ray beam and the target; anda detector aligned with the device, the detector located at an angle less than 180 degrees relative to the beam path of the first x-ray beam to receive the second x-ray beam from the device after the second x-ray beam has exited through one or more of the openings of the device;wherein the apparatus further comprises a processing unit configured to:determine two or more peaks based on data received from the detector, the two or more peaks selected from the group consisting of a first k-alpha peak, a first k-beta peak, a first Compton scatter peak, a second k-alpha peak, a second k-beta peak, and a second Compton scatter peak, andcalculate a parameter indicating a cancer characteristic by computing a ratio between two of the two or more peaks; andwherein the openings are arranged in a plurality of rows, and wherein each of the plurality of rows has multiple ones of the openings. 26. A method to image a target, comprising:directing a first x-ray beam generated from an x-ray source towards the target, wherein a second x-ray beam is generated by an interaction of the first x-ray beam with the target;using a device with an array of openings to collimate the second x-ray beam wherein at least two of the openings are next to each other in a side-by-side configuration;after the second x-ray beam has been collimated by the device, detecting the second x-ray beam using a detector that is placed at an angle less than 180 degrees relative to a path of the first x-ray beam; andobtaining quantum property for the target using the detected second x-ray beam;wherein the method further comprises:determining two or more peaks based on data received from the detector, the two or more peaks selected from the group consisting of a first k-alpha peak, a first k-beta peak, a first Compton scatter peak, a second k-alpha peak, a second k-beta peak, and a second Compton scatter peak, andcalculating a parameter indicating cancer characteristic, wherein the act of calculating the parameter comprises computing a ratio between two of the two or more peaks. 27. The method of claim 26, wherein the act of obtaining the quantum property comprises obtaining first energy spectrum data for a first part of the target and second energy spectrum data for a second part of the target. 28. The method of claim 26, wherein the second x-ray beam comprises a first portion and a second portion, wherein the act of using the device to collimate the second x-ray beam comprises collimating the second x-ray beam so that the first portion of the second x-ray beam is allowed to travel towards a first part of the detector, and the second portion of the second x-ray beam is allowed to travel towards a second part of the detector. 29. The method of claim 26, further comprising determine examination data based on signals generated by the detector, the examination data representing fluid build up, fluid retention, fluid wash out, rate of fluid build up, or rate of fluid wash out. 30. The method of claim 26, further reconstructing a volumetric image of the target using signals generated by the detector. 31. The method of claim 26, wherein the angle is a value that is anywhere between 30 degrees and 170 degrees. 32. The method of claim 26, wherein the second x-ray beam comprises a photo absorption component and a Compton scatter component. 33. The method of claim 26, further comprising collimating the first x-ray beam before the first x-ray beam reaches the target. 34. The method of claim 26, wherein the detector comprises a photo spectrum sensitive detector. 35. The method of claim 26, wherein the first x-ray beam comprises a cone beam. 36. The method of claim 26, wherein the first x-ray beam comprises a first fan beam, the first beam interacting with a first part, a second part, and a third part of the target to produce a first portion, a second portion, and a third portion, respectively, of the second x-ray beam;wherein the first part, the second part, and the third part of the target are in a first plane. 37. The method of claim 36, further comprising directing a second fan beam towards the target for examining parts of the target that are in a second plane. 38. The method of claim 26, wherein the first x-ray beam comprises a first pencil beam, the first pencil beam interacting with a first part of the target to produce a first portion of the second x-ray beam, and interacting with a second part of the target to produce a second portion of the second x-ray beam;wherein the first part and the second part of the target are along a first line. 39. The method of claim 38, further comprising directing a second pencil beam towards the target for examining parts of the target along a second line. 40. The method of claim 39, further comprising directing a third pencil beam towards the target for examining parts of the target along a third line. 41. The method of claim 26, further comprising using the quantum property to determine temporal information regarding the target. 42. The method of claim 26, wherein the quantum property comprises energy spectrum data. 43. The method of claim 26, wherein the parameter comprises information regarding a density of agent attached to a cancer site, information regarding angiogenesis, information regarding hypoxia, information regarding pathological functional marker(s), information regarding an identifier of cancer malignancy, information regarding an identifier of cancer growth, or information regarding an identifier of cancer growth rate. 44. The method of claim 26, wherein the second x-ray beam has a first beam portion and a second beam portion travelling in different respective directions, and wherein the device is used to allow the first beam portion to go through one or more of the openings while blocking the second beam portion. 45. The method of claim 26, wherein the angle is approximately 90 degrees. 46. The method of claim 26, wherein the angle is anywhere from 30 degrees to 100 degrees. 47. An apparatus to examine a target, comprising:an x-ray source configured to deliver a first x-ray beam towards the target;a device having an array of openings, wherein at least two of the openings are next to each other in a side-by-side configuration, the device located at an angle less than 180 degrees relative to a beam path of the first x-ray beam to receive a second x-ray beam resulted from an interaction between the first x-ray beam and the target;a detector aligned with the device, the detector located at an angle less than 180 degrees relative to the beam path of the first x-ray beam to receive the second x-ray beam from the device after the second x-ray beam has exited through one or more of the openings of the device; anda processing unit configured to determine two or more peaks based on data from the detector, the two or more peaks selected from the group consisting of a first k-alpha peak, a first k-beta peak, a first Compton scatter peak, a second k-alpha peak, a second k-beta peak, and a second Compton scatter peak, and wherein the processing unit is configured to determine a parameter indicating a cancer characteristic by calculating a ratio between two of the two or more peaks. 48. The apparatus of claim 47, wherein the angle is approximately 90 degrees. 49. A method to image a target, comprising:directing a first x-ray beam generated from an x-ray source towards the target, wherein a second x-ray beam is generated by an interaction of the first x-ray beam with the target;using a device with an array of openings to collimate the second x-ray beam wherein at least two of the openings are next to each other in a side-by-side configuration;after the second x-ray beam has been collimated by the device, detecting the second x-ray beam using a detector that is placed at an angle less than 180 degrees relative to a path of the first x-ray beam;determining two or more peaks based on data from the detector, the two or more peaks selected from the group consisting of a first k-alpha peak, a first k-beta peak, a first Compton scatter peak, a second k-alpha peak, a second k-beta peak, and a second Compton scatter peak; anddetermining a parameter indicating a cancer characteristic by calculating a ratio between two of the two or more peaks. 50. The method of claim 49, wherein the angle is approximately 90 degrees. |
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039765423 | description | A vent passage system 11 for conducting fission products from otherwise sealed containers 12 for fissionable material communicates with a fission product receiving system 13 and the pressure in both systems is maintained substantially less than the coolant gas pressure. The receiving system has a lower pressure than the vent passage system so that fission products will pass from the containers through the passages to the receiving system. At least one bleed passage 14 per fuel element is provided communicating from the cooling system of the reactor to the vent passage system for bleeding coolant gas into the vent passage system. The general construction of the reactor is of the type illustrated in U.S. Pat. No. 3,533,911, issued Oct. 30, 1970, the disclosure of which is incorporated by reference. A grid structure 16 comprises a plurality of vertical plates arranged to form receptacles 17 for receiving the upper ends of a plurality of fuel elements 18. The cross sections of the receptacles 17 and the upper ends of the fuel elements 18 may be rectangular, or may be circular as shown in the drawings, or may be any other convenient shape. The grid structure 16 is provided with annular bevelled surfaces 20 surrounding the lower ends of the receptacles 17. The fuel elements 18 each include an upper section consisting of an outer cylindrical wall 19 defining the top part of a tubular housing and which is received in an associated receptacle 17. Three webs 21 project inwardly at 120.degree. intervals from the cylindrical wall 19. The webs 21 are shown in FIG. 1 with one web in section to more clearly illustrate its constructional detail. The webs terminate in a central collar 22. The collar 22 is hollow and is releasably attached to the lower end of a support sleeve 23 by a bayonet type coupling. A part of this coupling is shown in FIG. 1 consisting of an outwardly extending projection 24 (one of three) from the support sleeve 23 and an inwardly extending insert 26 (one of three) through the wall of the collar 22. When the weight on the sleeve is relieved and the sleeve 23 is rotated with respect to the fuel element collar 22, the projections 24 are moved out from underneath the inserts 26 so that the fuel element 18 is released from the support sleeve 23. The support sleeve may be used to lower the fuel elements 18 from the receptacles 17 in the grid structure 16 into a refueling machine (not shown) and to move fresh fuel elements from the refueling machine up into position in the reactor core in the grid receptacles. An example of such a technique is described in the above-mentioned patent. In order to support the support sleeve 23, and accordingly, support the fuel element 18, the support sleeve is carried through a suitable penetration in the reactor vessel, not illustrated, to a position exterior of the reactor vessel. The broken out section at the top of FIG. 1 illustrates the exterior connections of the support sleeve 23. The support sleeve 23 is threaded at its upper end and a nut 27 is attached thereto. The nut 27 is tightened against a thrust washer 28, and this load, in turn, is transmitted through a thrust tube 29 which is coaxial with the support sleeve 23. The thrust tube 29 extends through the penetration in the unillustrated reactor vessel back down toward the grid structure 16. The lower end of the thrust tube 29 is provided with a series of slots 31 forming a collet section the lower end of which terminates in a bevel 32. The grid structure 16 includes a plurality of grid inserts disposed in suitably machined recesses 33 at the upper ends of the receptacles 17. Each grid insert consists of three inwardly extending webs 34 which extend inwardly at 120.degree. intervals from the cylindrical outer wall 36 and which terminate in a central collar 37. The outer wall 36 of the grid insert seats in the recess 33 and may be welded therein. The upper end of the collar 37 is provided with a bevelled surface 39 against which the collet section of the thrust tube 29 bears. When the nut 27 is tightened against the washer 28, thrust is transmitted through the thrust tube 29 to the grid insert at the bevelled surface 39. Accordingly, the grid structure 16 supports the fuel elements 18 even though the connections between the supports and the fuel elements (at the nut 27) are brought out to positions exteriorly of the reactor vessel. The lower portion of the tubular housing of each fuel element 18 is formed by an outer hexagonal wall 41 which is of slightly larger outside dimension than the wall 19. An annular bevelled surface 40 links the outer surface of the wall 19 and the outer surface of the wall 41. The bevelled surface 40 mates against the bevelled surface 20 on the grid structure 16 when the nut 27 is tightened. The fuel containers 12 are supported within the hexagonal wall 41 and extend along most of the length of the wall. Each of the fuel containers or pins 12 includes an outer cladding suitably sealed at its top by means of a plug 43. A plurality of transverse beams 42 are provided near the top of the fuel containers 12 extending across the interior of the housing and intercepting each other to form a network, as shown in FIG. 2. These beams are secured to the plugs 43 at the upper ends of the fuel pins to hold the fuel pins in the illustrated positions. The sealed interiors of the fuel containers 12 communicate with passages 45 in the beams 42 through suitable passages 43a in the plugs 43. The passages 43a and 45 form part of the vent passage system 11 and manifold the containers 12. A single tube 44 extends upwardly at one side of the fuel element and communicates with the passages 45 in the beams 42. The tube 44 forms a further part of the vent passage system 11 and connects the passages 45 to an internal fission product trap, described below. The upper section 19 of the tubular housing of the fuel element 18 is provided with a section 46 which is hollow and which completely surrounds the fuel element. The tube 44 connects the passages 45 to a passage 47 extending vertically in the outer wall 41 of the lower portion of the fuel element 18. An L-shaped passage 48 connects the upper end of the passage 47 with the hollow interior of the section 46. A vertical tube 49 carries the fission products to the top of the hollow interior of the section 46. There, the products are discharged from the tube 49 into a region filled with porous material 51. The hollow section 46 contains a fission product trap 54 of suitable material such as an activated charcoal adsorber. The fission product trap 54 extends from the porous material 51 down to a region near the closed lower end of the section 46 where further porous material 56 is provided. During normal operating conditions, the release of gaseous and volatile fission products from the fuel contained within the fuel containers 12 first diffuse through the fuel itself into any pores or interstices in the fuel. Then, by gaseous diffusion, the fission products move through any holes and clearances in the fuel and between the fuel and the external cladding into the manifold system. Some fission product traps may also be provided within the containers 12 themselves for additional delay if needed. The gas then moves upwardly through the tube 44, the passages 47 and 48, and the tube 49 into the porous material 51. The fission products then move through the fission product trap 54 and gas passing therethrough after a delay sufficient to alloy decay of the more active fission products, is collected in the porous material 56 at the bottom thereof. In order to remove gas from the porous material 56 beneath the fission product trap 54, an L-shaped passage 57 is provided which terminates in the bevelled shoulder 40 extending around the outside of the top of the lower portion of the fuel element 18. The shoulder 40 abuts the corresponding bevelled surface 20 on the lower part of the grid structure 16 surrounding the lower periphery of the receptacle 17. When the nut 27 is tightened, the fuel element is drawn up so that the shoulder 40 engages the grid structure for secure positioning of the fuel element. The mouth of the passage 57 is surrounded by a projection or boss 61 which projects outward from a recessed section 62 in the shoulder 40. The projection 61 does not extend far enough to engage the bevelled surface 20, leaving a clearance for purposes explained below. A passage 64 is provided in the grid structure 16 leading from the bevelled surface 20 and aligned with the mouth of the passage 57 to conduct gas therefrom to the receiving system 13. The passages 43a and 45, the tube 44, the passages 47 and 48, and tube 49, the fission product trap 54 and the passages 57 and 64 constitute the vent passage system 11 which manifolds gaseous fission products from the sealed fuel containers 12. The passage 64 interconnects with other passages, not shown, in the grid structure 16 to interconnect the vent passage systems of the other fuel elements in the core and conduct all fission product gases to a single receiving system. In oder to ensure that any break away flow from the internal trap 54 resulting from ambient pressure fluctuations is swept out of the passages leading to the receiving system 13 sufficiently rapidly to avoid substantial local decay and deposit, a bleed-in of coolant gas is utilized. Referring to FIG. 3, the coolant is added by providing a bleed passage 14, namely the clearance between the surface 20 and the projection 61. The bleed-in may be from any appropriate region of the reactor, depending on the inlet pressure desired. In the illustrated embodiment, sufficient space exists between the adjacent fuel elements to permit gas to enter the region. Provision is made in the fuel element 18 itself for controlling coolant flow. Coolant flow in the fuel element is downward (after passing through the grid insert 36) through the passageway defined by the wall 19 which constitutes the upper section of the tubular housing and then through the passageway defined by the wall 41 which constitutes the lower section of the tubular housing and which flares downward from a circular cross section to the hexagonal cross section shown in FIG. 2. In the lower section, the flowing coolant passes over the clad fuel pins 12 to remove heat therefrom. To minimize drag, the webs 34 and 21 are of streamlined cross section, as are the beams 42. The upper end of the wall 19 of the tubular housing is provided with a portion of varying interior cross section with respect to the direction of flow of coolant, such portion being indicated generally at 86 and constituting a constriction defined by a smooth convex annular surface. A movable plug 87 is supported coaxially within the housing adjacent the constriction 86. The upper portion of the plug has a convex shape which cooperates with the shape of the varying cross section constriction 86 to form an annular orifice through which coolant flowing downward through the tubular housing passageway must pass. The plug is shown in solid lines in the maximum flow or open position in which it presents a minimum resistance to coolant flow. The free flow area in the annular orifice is equal to the minimum free flow cross sectional area within the fuel element 18. This free flow area is maintained from the initial point of reduction to a location just downstream of the movable portion of the plug where the area increases to form a diffuser section. To decrease the coolant flow through a particular fuel element, the plug 87 is moved upstream into the constricted inlet passageway region. The pressure drop, which is normally associated with the decrease in orifice area, is augmented by the fact that the movable plug is in a high drag configuration in the minimum flow or "closed" position. The closed position of the plug is shown in the dotted lines of FIG. 1. In order to move the plug 87 from the open position to the closed position and back again, a rotary actuator rod 88 is provided which extends to the exterior of the pressure vessel through the same penetration opening through which the support sleeve 23 extends. The actuator rod includes a geared section 89 on its upper end to enable turning of the actuator rod by a suitable drive mechanism. The actuator rod terminates within the collar 22 of the fuel element and is threaded onto the upper end 91 of a threaded drive screw 92. The drive screw 92 is maintained coaxially within the collar 22 by an annular collar 93 extending therefrom. A retaining ring 94 sits on the collar 93 and is held in place by an externally threaded plug 96 which is threaded to appropriate internal threads in the interior of the collar 22. The lower end of the screw 92 is supported by a thrust bearing 97 which seats in the cavity 98 in a support member or central baffle 99. The baffle 99 is threaded onto an externally threaded section 101 at the lower end of the collar 22 and is located coaxial with the passageway through the housing. The baffle is of streamlined generally conical design to minimize drag and perturbation of coolant flow and forms an inverted teardrop shape with the plug 87 when the plug is in the open position. An internally threaded sleeve 102 surrounds the screw 92 in threaded engagement therewith and moves exially along the screw in response to the rotation thereof. The sleeve 102 is connected to the plug 87 by a series of arms 103 formed as a part of the plug which extend from the sleeve through suitable slots 104 in the collar 22. Thus, when the actuator rod 88 is turned, the screw 92 rotates in place and moves the plug 87 axially of the fuel element passageway, thereby varying the opening of the annular orifice. When in the fully open position, the movable plug 87 mates with the upper end of the baffle 99 at an interface 105 forming the inverted teardrop shape. The orifice actuator rod may be rotated manually above the reactor or remotely by a positioning motor. The orificing system may be designed to be "fail safe". That is, in the event of a part failure, the orifice may either remain in its original position or be returned to the open position as a result of both flow induced and gravitational forces. Another feature is that the flow can never be completely cut off even in the closed position, since there will be a finite free flow area when the plug is in that position. The fuel elements include a self-contained orificing system which is easily adjusted and which offers minimum resistance to coolant flow in the open position. Various modifications of the invention in addition to those shown and described herein will become apparent to those skilled in the art from the foregoing description and accompanying drawings. |
summary | ||
description | The following relates to the nuclear reactor arts, nuclear power generation arts, nuclear reactor control arts, nuclear reactor electrical power distribution arts, and related arts. In nuclear reactor designs of the integral pressurized water reactor (integral PWR) type, a nuclear reactor core is immersed in primary coolant water at or near the bottom of a pressure vessel. In a typical design, the primary coolant is maintained in a subcooled liquid phase in a cylindrical pressure vessel that is mounted generally upright (that is, with its cylinder axis oriented vertically). A hollow cylindrical central riser is disposed concentrically inside the pressure vessel. Primary coolant flows upward through the reactor core where it is heated, rises through the central riser, discharges from the top of the central riser, and reverses direction to flow downward back toward the reactor core through a downcomer annulus. The nuclear reactor core is built up from multiple fuel assemblies. Each fuel assembly includes a number of fuel rods. Control rods comprising neutron absorbing material are inserted into and lifted out of the reactor core to control core reactivity. The control rods are supported and guided through control rod guide tubes which are in turn supported by guide tube frames. In the integral PWR design, at least one steam generator is located inside the pressure vessel, typically in the downcomer annulus, and the pressurizer is located at the top of the pressure vessel, with a steam space at the top most point of the pressure vessel. Alternatively an external pressurizer can be used to control reactor pressure. A set of control rods is arranged as a control rod assembly that includes the control rods connected at their upper ends with a spider, and a connecting rod extending upward from the spider. The control rod assembly is raised or lowered to move the control rods out of or into the reactor core using a control rod drive mechanism (CRDM). In a typical CRDM configuration, an electrically driven motor selectively rotates a roller nut assembly or other threaded element that engages a lead screw that in turn connects with the connecting rod of the control rod assembly. In some assemblies, such as those described in U.S. Pat. No. 4,597,934, a magnetic jack may be used to control movement of one or more control rods. Control rods are typically also configured to “SCRAM”, by which it is meant that the control rods can be quickly released in an emergency so as to fall into the reactor core under force of gravity and quickly terminate the power-generating nuclear chain reaction. Toward this end the roller nut assembly may be configured to be separable so as to release the control rod assembly and lead screw which then fall toward the core as a translating unit. In another configuration, the connection of the lead screw with the connecting rod is latched and SCRAM is performed by releasing the latch so that the control rod assembly falls toward the core while the lead screw remains engaged with the roller nut. See Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2010/0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety; and Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, Intl Pub. WO 2010/144563 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. The CRDMs are complex precision devices which require electrical power to drive the motor, and may also require hydraulic, pneumatic, or another source of power to overcome the passive SCRAM release mechanism (e.g., to hold the separable roller nut in the engaged position, or to maintain latching of the connecting rod latch) unless this is also electrically driven. In existing commercial nuclear power reactors, the CRDMs are located externally, i.e. outside of the pressure vessel, typically above the vessel in PWR designs, or below the reactor in boiling water reactor (BWR) designs. An external CRDM has the advantage of accessibility for maintenance and can be powered through external electrical and hydraulic connectors. However, the requisite mechanical penetrations into the pressure vessel present safety concerns. Additionally, in compact integral PWR designs, especially those employing an internal pressurizer, it may be difficult to configure the reactor design to allow for overhead external placement of the CRDMs. Accordingly, internal CRDM designs have been developed. See U.S. Pub. No. 2010/0316177 A1 and Intl Pub. WO 2010/144563 A1 which are both incorporated herein by reference in their entireties. However, placing the CRDMs internally to the reactor vessel requires structural support and complicates delivery of electrical and hydraulic power. Electrical conductors that are usable inside the pressure vessel are generally not flexible and are not readily engaged or disengaged, making installation and servicing of internal CRDM units challenging. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In one illustrative embodiment, an apparatus is disclosed comprising a plurality of control rod drive mechanisms (CRDMs) each configured to raise or lower a control rod assembly and a distribution plate configured to be mounted in a nuclear reactor pressure vessel and including a plurality of connection sites at which the CRDMS are mounted, the distribution plate including electrical power distribution lines disposed on or in the distribution plate for distributing electrical power to the CRDMs mounted on the distribution plate. A method is also disclosed comprising installing a CRDM in a nuclear reactor by operations which include attaching the CRDM to a top plate of a standoff and connecting a mineral insulated cable between the CRDM and an electrical connector disposed in or on a bottom plate of the standoff to form a CRDM/standoff assembly and mounting the bottom plate of the CRDM/standoff assembly to a distribution plate wherein the mounting connects an electrical power line disposed on or in the distribution plate with the electrical connector disposed in or on the bottom plate of the standoff. In another illustrative embodiment, an apparatus is disclosed comprising a nuclear reactor including a core comprising a fissile material disposed in a pressure vessel, a mechanical reactor component disposed inside the pressure vessel and having a mounting flange with a power connector, and a power distribution plate disposed inside the pressure vessel and having a connection site configured to mate with the flange of the mechanical reactor component, the connection site including a power connector configured to mate with the power connector of the flange of the mechanical reactor component when the flange of the mechanical reactor component is mated with the connection site, power lines on or in the power distribution plate being arranged to deliver power to the power connector of the connection site, wherein the flange of the mechanical reactor component is mated with the connection site of the power distribution plate. FIG. 1 illustrates an integral Pressurized Water Reactor (integral PWR) generally designated by the numeral 10. A reactor vessel 11 is generally cylindrical and contains a reactor core 1, a steam generator 2, and a pressurizer 3. Although a pressurized water reactor (PWR) is depicted, a boiling water reactor (BWR) or other type of nuclear reactor is also contemplated. Moreover, while the disclosed rapid installation and servicing techniques are described with reference to illustrative internal CRDM units, these techniques are readily adapted for use with other internal nuclear reactor components such as internal reactor coolant pumps. In the illustrative PWR, above the core 1 are the reactor upper internals 12 of integral PWR 10, shown in inset. The upper internals 12 are supported by a mid flange 14, which in the illustrative embodiment also supports internal canned reactor coolant pumps (RCPs) 16. More generally, the RCPs may be external pumps or have other configurations, and the upper internals may be supported otherwise than by the illustrative mid flange 14. The upper internals include control rod guide frames 18 to house and guide the control rod assemblies for controlling the reactor. Control Rod Drive Mechanisms (CRDMs) 20 raise and lower the control rods to control the reactor. In accordance with one embodiment, a CRDM distribution plate 22 supports the CRDMs and provides power and hydraulics to the CRDMs. A riser transition 24 directs coolant flow upward. Control rods are withdrawn from the core by CRDMs to provide enough positive reactivity to achieve criticality. The control rod guide tubes provide space for the rods and interconnecting spider to be raised upward away from the reactor core. The CRDMs 20 require electric power for the motors which move the rods, as well as for auxiliary electrical components such as rod position indicators and rod bottom sensors. In some designs, the force to latch the connecting rod to the lead screw, or to maintain engagement of the separable roller nut, is hydraulic, necessitating a hydraulic connection to the CRDM. To ensure passive safety, a positive force is usually required to prevent SCRAM, such that removal of the positive force initiates a SCRAM. The illustrative CRDM 20 is an internal CRDM, that is, is located inside the reactor vessel, and so the connection between the CRDM 20 and the distribution plate 22 is difficult to access. Servicing of a CRDM during a plant shutdown should preferably be rapid in order to minimize the length of the shutdown. To facilitate replacing a CRDM in the field, a standoff assembly connected to the distribution plate 22 to provide precise vertical placement of the CRDM 20 is also configured to provide electrical power and hydraulics to the CRDM 20 via connectors that require no action to effectuate the connection other than placement of the standoff assembly onto the distribution plate 22. After placement, the standoff is secured to the distribution plate by bolts or other fasteners. Additionally or alternatively, it is contemplated to rely upon the weight of the standoff assembly and CRDM to hold the assembly in place, or to use welds to secure the assembly. The illustrative distribution plate 22 is a single plate that contains the electrical and hydraulic lines and also is strong enough to provide support to the CRDMs and upper internals without reinforcement. In another embodiment, the distribution plate 22 may comprise a stack of two or more plates, for example a mid-hanger plate which provides structural strength and rigidity and an upper plate that contains electrical and/or hydraulic lines to the CRDMs via the standoff assembly. The motor/roller nut assembly of the CRDM is generally located in the middle of the lead screw's travel path. When the control rod is fully inserted into the core, the roller nut is holding the top of the lead screw, and, when the rod is at the top of the core, the roller nut is holding the bottom of the lead screw and most of the length of the lead screw extends upward above the motor/roller nut assembly. Hence the distribution plate 22 that supports the CRDM is positioned “below” the CRDM units and a relatively short distance above the reactor core. FIG. 2 illustrates the distribution plate 22 with one standoff assembly 24 mounted for illustration, though it should be understood that all openings 26 would have a standoff assembly (and accompanying CRDM) mounted in place during operation of the reactor. Each opening 26 allows a lead screw of a control rod to pass through and the periphery of the opening provides a connection site for a standoff assembly that supports the CRDM. The lead screw passes down through the CRDM, through the standoff assembly, and then through the opening 26. The distribution plate 22 has, either internally embedded within the plate or mounted to it, electrical power lines (e.g., electrical conductors) and hydraulic power lines to supply the CRDM with power and hydraulics. The illustrative openings 26 are asymmetric or keyed so that the CRDM can only be mounted in one orientation. As illustrated, there are 69 openings arranged in nine rows to form a grid, but more or fewer could be used depending on the number of control rods in the reactor. The distribution plate is circular to fit the interior of the reactor, with openings 28 to allow for flow through the plate. In some designs, not all openings may have CRDMs mounted to them or have associated fuel assemblies. One possible arrangement for the hydraulic and/or electrical power lines is shown in FIG. 3. The electrical power lines, shown as dashed lines 30, runs straight between the rows of openings 26 in the distribution plate 22. Because of the limited flexibility of typical cables compatible with the high temperature and caustic environment inside the pressure vessel, the power lines within the distribution plate 22 for delivering electrical and/or hydraulic power to the CRDMs should be straight or have gradual, large-radius turns To accommodate both electrical and hydraulic power lines, in one embodiment the hydraulic power lines (not shown) follow a similar pattern to that of the electrical lines 30. In another embodiment the hydraulic power lines follow a similar path, except that the pattern of hydraulic lines is rotated 90° from the electrical path. The hydraulic power lines and electrical power lines, if internal to the plate, are separated by depth in the plate. Alternatively, one or other can be disposed on a top or bottom surface of the plate 22, or they can be disposed on opposite top and bottom surfaces of the plate. FIG. 4 illustrates a small cutaway view of one connection site of the distribution plate 22 for connecting a CRDM to the distribution plate. The connection site includes an opening 26 for passing the lead screw of a single CRDM. Located around the opening 26 are apertures 40 to accept bolts (more generally, other securing or fastening features may be used) and electrical connectors 42 for delivering electrical power to the CRDM. The illustrative CRDM employs hydraulic power to operate the SCRAM mechanism, and accordingly there is also a hydraulic connector 44 to accept a hydraulic line connection. The opening 26 and its associated features 40, 42, 44 create a connection site to accept the CRDM/standoff assembly. Internal to the plate may be junction boxes to electrically connect the connection sites to the electrical power lines 30 running in between rows of connection sites. Similarly, the hydraulic connector 44 may connect to a common hydraulic line 32 running through the distribution plate perpendicular to the electrical power lines 30 and separated by depth. FIG. 5 illustrates a standoff 24 that suitably mates to opening 26 in the distribution plate 22. The standoff assembly has a cylindrical midsection with plates 45, 46 of larger cross-sectional area on either end of the midsection. The circular top plate 45 mates to and supports a CRDM 20. The square bottom plate 46 mates to the distribution plate 22. Although the illustrative bottom plate 46 is square, it may alternatively be round or have another shape. When the CRDM 20 and the top plate 45 of the standoff 24 are secured together they form a unitary CRDM/standoff assembly in which the bottom plate 46 is a flange for connecting the assembly to the distribution plate 22. Two bolt lead-ins 50 on diagonally opposite sides of the lower plate 46 mate to the apertures 40 of the distribution plate. The bolt lead-ins, being mainly for positioning the CRDM standoff, are the first component on the standoff to make contact with the distribution plate when the CRDM is being installed, ensuring proper alignment. Two electrical power connectors 52 on diagonally opposite corners of the bottom plate 46 mate to corresponding electrical power connectors 42 of the distribution plate 22. A hydraulic line connector 54 on the bottom plate 46 mates to the corresponding hydraulic power connector 44 of the distribution plate 22. A central bore 56 of the standoff 26 allows the lead screw to pass through. The connectors 42, 44 inside the distribution plate 22 optionally have compliance features, such as springs, belleville washers or the like, to ensure positive contact, and the opposing bolts that attach at lead-ins 50 serve as tensioning devices to ensure proper seating of both the CRDM electrical connectors and hydraulic connectors. Flow slots 58 permit primary coolant to flow through the standoff. FIG. 6 illustrates a perspective view focusing on the top plate 45 of the standoff 24. The top plate 45 of the standoff mates to the CRDM and is attached via bolt holes 62. Bolt holes 62 may be either threaded or unthreaded. The CRDM and standoff can be attached to each other and electrical connections 52 and hydraulic connection 54 made before the CRDM is installed so as to form a CRDM/standoff assembly having flange 46 for connecting the assembly with the connection site of the distribution plate 22. The bottom plate 46 of the standoff 24 is secured to the connection site via bolts passing through holes 50 and secured by nuts, threads in the bolt holes 40, or the like. FIG. 7 illustrates another suitable standoff 70, with a generally square upper mounting plate 71 for the CRDM. The upper mounting plate 71 for the CRDM includes a notch 76 to enable electrical access to the bottom of the CRDM, bolt holes 77 to attach the CRDM, and notches 78 at the corners of the plate to permit primary coolant flow. The lower mounting plate 72, which connects to the distribution plate 22, includes three electrical power connectors 73, a hydraulic power connector 74, and flow slots 75 to permit coolant flow. The standoff 70 may have more or fewer electrical connections depending on whether CRDM components share an electrical connection or have their own connection. FIG. 8 shows standoff 24 connected to a CRDM 20 to form a CRDM/standoff assembly that can be mounted to the distribution plate. CRDM electrical cabling 80 extends upward to conduct electrical power received at the electrical connectors 52 to the motor or other electrical component(s) of the CRDM 20. Similarly, a CRDM hydraulic line 82 extends upward to conduct hydraulic power received at hydraulic connector 54 to the hydraulic piston or other hydraulic component(s) of the CRDM 20 to maintain latching—removal of the hydraulic power instigates a SCRAM. The entire assembly including the CRDM and the standoff is then installed as a unit on a distribution plate, simplifying the installation process of a CRDM in the field. In one embodiment, the electrical cables 80 are mineral insulated cables (MI cables) which generally include one, two, three, or more copper conductors wrapped in a mineral insulation such as Magnesium Oxide which is in turn sheathed in a metal. The mineral insulation could also be aluminum oxide, ceramic, or another electrically insulating material that is robust in the nuclear reactor environment. MI cables are often sheathed in alloys containing copper, but copper would corrode and have a negative effect on reactor chemistry. Some contemplated sheathing metals include various steel alloys containing nickel and/or chromium, or a copper sheath with a protective nickel cladding. The electrical lines 30 in the distribution plate 22 (see FIG. 3) are also suitably MI cables, although other types of cabling can be used inside the distribution plate 22 if they are isolated by embedding in the plate. MI cables advantageously do not include plastic or other organic material and accordingly are well suited for use in the caustic high temperature environment inside the pressure vessel. The relatively rigid nature of the MI cables is also advantageous in that it helps ensure the integrity of the pre-assembled CRDM/standoff assembly during transport and installation. However, the rigidity of the MI cables limits their bending radius to relatively large radius turns, so that the MI cables inside the distribution plate 22 should be arranged as straight lines with only large-radius turns, e.g. as shown in FIG. 3. The large area of the distribution plate 22, which spans the inner diameter of the pressure vessel, facilitates a suitable arrangement of the MI cables inside the plate 22. Additionally, some types of MI cables are susceptible to degradation if the mineral insulation is exposed to water. Accordingly, the ends of the MI cables, e.g. at the coupling with the connector 52 in the standoff and the coupling of the power lines 30 with the electrical connectors 42 in the distribution plate 22, should be sealed against exposure to the primary coolant water. However, advantageously, the connectors 42, 52 themselves can be immersed in water. This makes installation, to be further described, readily performed even with the reactor core immersed in primary coolant. FIG. 9 is an overhead view of a standoff assembly with installed CRDM. This view would be looking down from the upper internals into the core when the CRDM and standoff assembly are mounted in the reactor. Connecting rod 90 is contained within lead screw 92 which is raised and lowered by the CRDM. Bolt holes 50 are visible at diagonally opposite corners. Cables can be seen running to electrical connectors 52 at the other pair of corners. A portion of the vertically extending CRDM hydraulic line 82 can be seen in end view. FIG. 10 shows a suitable configuration for the mating electrical connectors 42, 52 of the distribution plate and CRDM/standoff assembly flange 46, respectively. The female electrical connector 52 of the standoff assembly 24 lowers onto and covers the male electrical connector 42 of the distribution plate. The connectors 42, 52 preferably include glands or other features to prevent ingress of water to the mineral insulation of the MI cables 30, 80 at the junctions of these cables with the respective connectors 42, 52. In this way, the connectors 42, 52 can be mated underwater without exposing the metal insulation, so as to facilitate installing the CRDM/standoff assembly at the connection site of the distribution plate 22 while keeping the reactor core and the distribution plate 22 submerged in primary coolant. To ensure a good electrical connection, the connection between connectors 42, 52 can be purged to evacuate any trapped water. Alternatively, the electrical connectors could be mated and not purged, albeit typically with some increased resistance due to wet connectors. FIG. 11 shows a suitable hydraulic interface from standoff assembly 24 to distribution plate 22. An electrical connector 52 as already described with reference to FIG. 10 is also shown. The female hydraulic connector 54 of the standoff assembly mates to the male hydraulic connector 110. The female hydraulic connector 54 is a socket that is machined directly into the bottom of the lower plate 72 of the standoff assembly 24. The top of the hydraulic connector 54 has a nipple to allow the hydraulic line 82 to be connected to the standoff assembly 22. The hydraulic line then runs up the CRDM to a piston assembly (not shown) which latches the lead screw. The hydraulic connectors 54, 110 optionally have compliance features, such as springs, belleville washers or the like, to ensure positive contact. A continuous flow of primary coolant is used as hydraulic fluid to maintain the CRDM latched during operation, so some leakage from the hydraulic connector into the pressure vessel is acceptable. In view of this, in some embodiments the mating of the hydraulic power connector of the CRDM 20 with the corresponding hydraulic power connector of the connection site of the distribution plate 22 forms a leaky hydraulic connection. Accordingly, a sufficient sealing force for the hydraulic connection is provided by the weight of the CRDM/standoff assembly and/or the force imparted by the hold-down bolts that pass through the bolt lead-ins 50 of the standoff assembly and bolt holes 40 of the distribution plate. FIG. 12 diagrammatically illustrates a method of connecting a CRDM to a standoff to form a preassembled CRDM/standoff assembly and then connecting the CRDM/standoff assembly to the distribution plate. In step S1210, the method starts. In step S1220, the CRDM is bolted to the standoff assembly by a plurality of bolts. In step S1230, the hydraulic cable is connected to the hydraulic connector of the standoff plate and the electrical cable(s) are connected the electrical connection(s). In step S1240, the standoff plate, with CRDM bolted on top of it, is lowered onto the distribution plate, with the bolt holes 50 making contact first to ensure proper alignment of the standoff assembly and CRDM. In step S1250, the hold-down bolts are installed and torqued to attach the standoff assembly to the distribution plate and to ensure positive contact in the hydraulic and electrical connectors. At step S1260, the electrical connectors are optionally purged. At step S1270, the method ends. FIG. 13 illustrates a method of removing a CRDM from a distribution plate. In step S1310, the method starts. In step S1320, the hold-down bolts are removed. In step S1330, the CRDM and connected standoff assembly are lifted away from the distribution plate. In step S1340, the CRDM is optionally removed from the standoff assembly for repair or replacement. In step S1350, the method ends. The disclosed approaches advantageously improve the installation and servicing of powered internal mechanical reactor components (e.g., the illustrative CRDM/standoff assembly) by replacing conventional in-field installation procedures including on-site routing and installation of power lines (e.g. MI cables or hydraulic lines) and connection of each power line with the powered internal mechanical reactor component with a simple “plug-and-play” installation in which the power lines are integrated with the support plate and power connections are automatically made when the powered internal mechanical reactor component is mounted onto its support plate. The disclosed approaches leverage the fact that most powered internal mechanical reactor components are conventionally mounted on a support plate in order to provide sufficient structural support and to enable efficient removal for servicing (e.g., a welded mount complicates removal for servicing). By modifying the support plate to also serve as a power distribution plate with built-in connectors that mate with mating connectors of the powered internal mechanical reactor component during mounting of the latter, most of the installation complexity is shifted away from the power plant and to the reactor manufacturing site(s). The example of FIGS. 1-13 is merely illustrative, and numerous variations are contemplated. For example, the CRDM/standoff assembly can be replaced by a CRDM with an integral mounting flange, that is, the standoff can be integrally formed with the CRDM as a unitary element (variant not shown). With reference to FIGS. 14 and 15, as another illustrative example the disclosed approaches are applied to internal reactor coolant pumps (RCPs) 1400, such as are disclosed in Thome et al., U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. For placement of the internal RCPs 1400 in the cold leg (i.e. the downcomer annulus), the RCPs 1400 are envisioned to be mounted on an annular pump plate 1402 disposed in the downcomer annulus. The pump plate 1402 serves as structural support for the RCPs 1400 and also as a pressure divider to separate the upper suction volume and the lower discharge volume. In the illustrative embodiment there are eight connection sites with six of these shown in FIG. 14 as containing RCPs 1400, and the remaining two being unused to illustrate the connection sites. The pump plate 1402 is modified to include MI cables 1404, 1405 disposed in or on the pump plate 1402. The annular shape of the pump plate 1402 precludes long straight runs of MI cable; however, the illustrative MI cables 1404, 1405 are oriented circumferentially with a large bend radius comparable with (half of) the inner diameter of the pressure vessel 11. Bolt apertures 1440 and electrical connectors 1442 are analogous to bolt apertures 40 and electrical connectors 42 of the illustrative CRDM embodiment, respectively. The opening 26 of the connection site of distribution plate 22 translates in the pump plate 1402 to be a generally circular opening 1426 (optionally keyed by a suitable keying feature, not shown) through which the RCPs 1400 pump primary coolant downward. As yet another contemplated modification, it will be appreciated that the female connector can be located in the supporting power distribution plate while the male connector can be located in the flange, standoff or other mounting feature of the internal mechanical reactor component. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. |
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055815880 | description | DETAILED DESCRIPTION OF THE INVENTION The present invention is a technique for solving the problem of achieving low corrosion potentials in the high-flux, in-core region (or in other regions which may have very high oxidant supply rates from high concentrations and/or high fluid flow rates/convection). The technique entails the formation of an electrically insulating protective coating that is doped with a noble metal on SCC-susceptible surfaces of metal components of a water-cooled nuclear reactor. The insulated protective coating is designed to alter the balance between the rate of supply of oxidants to the surface and the rate of recombination on the surface by limiting the supply kinetics (by restricting the mass transport of reactants through a porous, insulated layer), while at the same time providing for the catalytic reduction of the oxidants within the layer by a noble metal contained within the coating. This invention is related to allowed co-pending patent application Ser. No. 08/226,153 filed on Apr. 11, 1994 [Attorney Docket No. GENE-24-BR-05538], which is herein incorporated by reference. The technique of the present invention is based on the following fundamental considerations. The first consideration is that corrosion potentials are created only at metal-water interfaces. Thus, while on a metal coating the corrosion potential is formed at the interface of the metal coating with the bulk water, on a porous insulated coating, the corrosion potential is formed at the interface of the substrate metal and the water with which it is in contact (i.e., the water in the pores). The influence of corrosion potential on stress corrosion cracking results from the difference in corrosion potential at the generally high potential crack mouth/free surface versus the always low potential (e.g., -0.5 V.sub.SHE) within the crack/crevice tip. This potential difference causes electron flow in the metal and ionic flow in the solution, which induces an increase in the anion concentration in the crack, as in a classical crevice. FIG. 8 is a schematic of electrochemical processes which generally lead to elevated corrosion potentials on the outside (mouth) of a crack and low corrosion potentials in the inside (tip) of the crack. The potential difference .DELTA..o slashed..sub.c causes anions A.sup.- (e.g., Cl.sup.-) to concentrate in the crack, but only if there is both an ionic path and an electron path. FIGS. 9A to 9E provide a schematic comparison of the corrosion potentials .o slashed..sub.c which form under high radiation flux: (A) on an uncoated (e.g., stainless steel) component (high .o slashed..sub.c); (B) on a component coated with a catalytic metal coating where the rate of supply of reactants to the surface is not too rapid (low .o slashed..sub.c); (C) on a component coated with a catalytic metal coating where the rate of supply of reactants to the surface approaches or exceeds the recombination kinetics for H.sub.2 and O.sub.2 (moderate .o slashed..sub.c); (D) on a component coated with an insulated protective coating (at a low corrosion potential provided that oxidant concentrations do not get too high, see FIG. 11); and (E) on a component coated with an insulated protective coating that is doped with a noble metal (always at a low corrosion potential). Thus, to influence stress corrosion cracking, the elevated crack mouth corrosion potential must form on a surface that is in electrical contact with the component of interest. If an insulating coating (see FIGS. 9 and 10) were applied to a metal component and some porosity or cracking in the coating is assumed to exist, the corrosion potential would be formed only at the metal component-water interface. Thus, a crevice would be formed by the coating, but since it is electrically insulating, the crevice cannot represent an "electrochemical" crevice, but only a "restricted mass transport" geometry. The critical ingredient in "electrochemical" crevices is the presence of a conducting material in simultaneous contact with regions of high potential (e.g., a crack mouth) and regions of low potential (e.g., a crack tip). Thus, it would not help to have a component covered by an insulating layer, which layer is in turn covered by a metal layer (or interconnected metal particles) within which exists a crevice or crack. Under these conditions, the aggressive crevice chemistry could form in the outer metal layer, which in turn would be in contact with the component. Therefore, the amount of noble metal that is used as a dopant should be limited so that a conductive pathway cannot be formed through the thickness of the coating, such as by forming a series of interconnected noble metal particles. While the maximum amount of noble metal depends on many factors, including the particle size of the noble metals used to form the coating (in the case where the noble metal is added as a powder), the insulating material used, the existence of a removable phase as discussed elsewhere herein, the morphology of the coating and others. However, based on results reported with Au--Cu alloys, the amount of the noble metal in the coating should be about 20 atomic percent, or less, based on the concepts of percolation theory, see for example, K Sieradzki, "Atomistic and Micromechanical Aspects of Environment-Induced Cracking of Metals", Proceedings of the First International Conference on Environment Induced Cracking of Metals, NACE, 1989. The second consideration is that if the insulating coating is impermeable to water, then obviously there can be neither a corrosion potential formed on the underlying metal, nor concern for stress corrosion cracking. Any pores or fine cracks in an insulating layer provide highly restricted mass transport and thus are equivalent to a very thick near-surface boundary layer of stagnant water. Since oxidants are always being consumed at metal surfaces, this very restricted mass transport (reduced rate of oxidant supply) causes the arrival rate of oxidants through the insulated coating to the substrate to decrease below the rate of their consumption. Under these mass transport limiting circumstances, the corrosion potential rapidly decreases to values .ltoreq.-0.5 V.sub.SHE, even for high bulk oxidant concentrations, and even in the absence of stoichiometric excess hydrogen (or any hydrogen). Numerous observations consistent with this have been made, including low potentials on stainless steel surfaces at low oxygen levels (e.g., 1 to 10 ppb), as well as in (just inside) crevices/cracks, even at very high bulk oxygen levels. Thus, corrosion potentials .ltoreq.-0.5 V.sub.SHE can be achieved even at high bulk oxidant concentrations and, not only in the absence of stoichiometric excess hydrogen, but also in the absence of any hydrogen. This may prove to be a critical invention for BWR plants which are unable (because of cost or because of the high N.sup.16 radiation levels from hydrogen addition) to add sufficient hydrogen to guarantee stoichiometric excess hydrogen conditions at all locations in their plant. While the present invention provides protection from SCC in the absence of stoichiometric excess hydrogen (or any hydrogen), it should be noted that in order for the noble metal to catalyze oxidants in the crevices (restricted mass transport regions), that a reductant species such as hydrogen must be present, although the reductant need not be in stoichiometric quantities. Where the reductant species is present, the presence of the noble metal would allow the use of insulating coatings of the present invention to be used at oxidant concentrations that would be too large (i.e. that would produce a corrosion potential that is too large (e.g., greater than the critical potential) for insulating coatings that do not contain a noble metal dopant, such as disclosed in the reference described above. While various non-conducting materials could be used, zirconia (ZrO.sub.2) is a good initial choice because it can be thermally sprayed and is very stable in high-temperature water, both structurally (e.g., it is not prone to spalling and is not susceptible to environmentally assisted cracking) and chemically (e.g., it does not dissolve or react). Zirconia can be obtained in various particle sizes, so that there is flexibility in adjusting the thermal spray parameters. Alumina is also an option. The dissolution rate of alumina in 288.degree. C. water is higher than that for zirconia, but is still very low. Various other metal oxides, carbides, nitrides or carbides may also be suitable, so long as they are mechanically and chemically stable in a high temperature water environment, including not being subject to dissolution in high temperature water and not being subject to spalling under reactor operating conditions. The noble metals that may be used include any metals that are not subject to dissolution in high temperature water, and that will act as a catalyst for the reduction of oxidizing species such as oxygen and hydrogen peroxide that exist in high temperature water. Based on results with noble metal coatings in high temperature water described herein and well-known catalytic characteristics, it is believed that the metals iridium, palladium, platinum, osmium, rhodium or ruthenium will provide suitable noble metal dopants, and that the use of the noble metals palladium or platinum will be preferred, largely for cost considerations. Also, the noble metal may comprise an alloy of noble metals. The noble metal may be doped into the zirconia by any one of a number of known methods, including thermal spraying using a powder feed where the feed powder is a mixture of the insulating material and the noble metal, and others. FIG. 10 is a schematic illustration of an insulated protective coating doped with a noble metal of the present invention, depicted as particles 4 of zirconia powder and particles 6 of a noble metal powder which have been thermally sprayed onto a metal component surface 2. Due to the insulating nature of zirconia, there is no electrical connection between external (high oxidant) water and the metal component substrate. Thus, the insulated protective coating containing the noble metal prevents an electrochemical crevice cell from being formed (see FIG. 8), while restricting mass transport of oxidants to the underlying metal substrate (see FIGS. 2 and 7) to sufficiently low rates such that the corrosion potential of the metal component is always low (i.e., -0.5 V.sub.SHE). At the same time, the coating is also catalyzing the reduction of the oxidants in the crevices, which also further serves to reduce the corrosion potential, or provide tolerance to higher bulk oxidant concentrations. Preliminary experimental data (shown in FIG. 11 ) were obtained in 288.degree. C. pure water on a cylindrical stainless steel electrode coated with yttria-stabilized zirconia (YSZ) by air plasma spraying. A Cu/Cu.sub.2 O membrane reference electrode was used to measure the corrosion potentials of the stainless steel autoclave, a platinum wire and the YSZ-coated stainless steel specimen. At oxygen concentrations up to .apprxeq.1 ppm (during BWR operation, the equivalent oxygen concentration (O.sub.2 +0.5.times.H.sub.2 O.sub.2) is about 100 to 600 ppb), the corrosion potential of the YSZ-coated specimen remained at .ltoreq.-0.5 V.sub.SHE despite the high potentials registered on the stainless steel autoclave (+0.20 V.sub.SHE) and the platinum electrode (+0.275 V.sub.SHE). This is consistent with numerous observations of low potentials on stainless steel surfaces at low oxygen levels (e.g., 1 to 10 ppb) as well as inside crevices/cracks, even at very high oxygen levels. Similar observations were obtained in hydrogen peroxide, where low potentials were observed on the YSZ-coated specimen at concentrations above 1 ppm (see FIG. 12). By contrast, uncoated stainless steel exhibited a high corrosion potential of .apprxeq.+0.150 V.sub.SHE. Low potentials were also observed on the YSZ-coated specimen in water containing 1 ppm O.sub.2 when the specimen was rotated at 500 rpm, corresponding to 0.7 m/sec linear flow rate. This is not surprising, since the higher flow rates merely act to reduce the thickness of the stagnant boundary layer of liquid, a layer whose thickness is small relative to the zirconia coating. The success in maintaining low corrosion potentials under these conditions shows that the electrically insulating zirconia layer greatly reduces mass transport to the underlying metal surface such that, even in the absence of catalytic agents such as palladium, the cathodic (oxygen reduction) reaction is mass transport limited just as in uncoated specimens in solutions of very low dissolved oxygen content. Further corroboration exists in the corrosion potential measurements on Zircaloy in 288.degree. C. water, which apparently are always lower than -0.5 V.sub.SHE, even in aerated solutions. The relatively highly electrically insulating nature of the zirconia film causes the corrosion potential to be formed at the metal surface where the oxidant concentration is very low due to its restricted transport through the zirconia film. Additional experimental data is presented in FIGS. 13 and 14. A coating made of yttria-stabilized zirconia powder was deposited in three different thicknesses (3, 5 and 10 mils) on the fresh metal surface of Type 304 stainless steel (0.25 inches in diameter and 1 inch long) by air plasma spraying. The corrosion potentials of the zirconia-coated electrodes, a pure zirconium electrode and uncoated Type 304 stainless steel were measured against a Cu/Cu.sub.2 O/ZrO.sub.2 reference electrode in 288.degree. C. water containing various amounts of oxygen. After the corrosion potential measurement, test specimens were immersed in 288.degree. C. water containing various water chemistry conditions for 3 months at open circuit. In the initial tests, YSZ-coated stainless steel electrodes were mounted in the autoclave along with a zirconium electrode, an uncoated Type 304 stainless steel electrode and the reference electrode. All specimens were immersed in pure 288.degree. C. water at a flow rate of 200 cc/min for 2 days. The corrosion potential was measured sequentially with incremental addition of oxygen, as shown in FIG. 13. At given oxygen levels up to 200-300 ppb, the YSZ-coated electrodes showed low potentials (<-0.5 V.sub.SHE) essentially equivalent to those of the pure zirconium electrode, compared to the Type 304 stainless steel corrosion potential values measured at the same level of oxygen. Further increase of the oxygen concentration increased the corrosion potential of the YSZ-coated electrodes. After the system was left in 288.degree. C. water containing various water chemistry conditions for 3 months, the corrosion potential was again measured by increasing the oxygen concentration (see FIG. 14). This data indicates that the corrosion potential behavior of the YSZ-coated electrodes was retained for extended periods. From the foregoing data, it is apparent that the application of a YSZ coating on the surface of Type 304 stainless steel appears is advantageous in maintaining a low corrosion potential (<-0.5 V.sub.SHE) at high oxygen levels (up to about 300 ppb), even in the absence of hydrogen, by reducing mass transfer of oxygen to the metal surface and thereby mitigating SCC of the structural material. Since the oxygen concentration during operation of a BWR is about 200 ppb, SCC in BWR structural components could be mitigated by the application of a YSZ coating or any other electrically insulating protective coating on the surfaces of the structural material. The present invention is particularly suited for use in water-cooled nuclear reactors that contain high-temperature water, however, the invention may also be utilized in any other systems that utilize high-temperature water where SCC is a consideration, such as conventional turbines and generators. The foregoing method has been disclosed for the purpose of illustration. Variations and modifications of the disclosed method will be readily apparent to practitioners skilled in the art of water chemistry. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter. |
summary | ||
062787586 | summary | FIELD OF INVENTION The present invention relates to a nuclear reactor fuel assembly, and more specifically to a support grid structure for supporting fuel rods thereof. BACKGROUND OF THE INVENTION A typical structure of a fuel assembly for a pressurized water reactor is shown in FIG. 5. Describing it briefly, an upper and lower nozzle 1, 3 having a plurality of coolant flow holes machined therein are connected with a plurality of hollow guide tubes 5 which are parallel to one another. Support grids 7 having grid cells positioned in a rectangular arrangement which individually receive the hollow guide tubes 5 therethrough and which are fixed to the guide tubes 5 support the fuel rods 9 by placing them individually through the remainder of the grid cells. A fuel assembly 10 is thus constructed and in order to more clearly show the structure, a portion of the fuel assembly 10 from which the surroundings about one support grid 7 are removed is shown in a partial perspective view in FIG. 6. As readily understandable from FIG. 6, the disposition of the fuel rods 9 is in a square arrangement with equal numbers of columns and rows and it is constructed showing some of the fuel rods 9 at specified locations being replaced with the hollow guide tubes 5. The support grid 7 defining grid cells which individually receive, as described before, the fuel rods 9 and the hollow guide tubes 5 disposed in a square arrangement, is essentially constructed by combining two kinds of thin metal straps as shown in FIGS. 7a and 7b, namely, straps 20, 30, with each other in a perpendicular relationship. The support grid 7 defines grid cells located in a 14.times.14 arrangement, and the straps 20, 30 each essentially have an identical configuration and differ from each other in the positions at which slits 21, 31 for receiving another strap corresponding thereto are formed. In other words, slits 21 in the strap 20 are positioned on an upper side (downstream in the coolant flow), while slits 31 in the strap 30 are positioned on a lower side (upstream in the coolant flow). Mixing vanes 23, 25, 33, 35 are integrally formed at a downstream edge of the straps 20, 30 in alignment with the slits 21, 31. Furthermore, tabs 27, 29, 37, 39 for welding the straps 20, 30 which are assembled by using slits 21, 31 are provided. The positional relationship among the slits 21, 31, the mixing vanes 23, 25, 33, 35 and the tabs 27, 29, 37, 39 that are described above is schematically shown in an enlarged manner in FIGS. 5a and 8b. The above described mixing vanes 23, 25 and 33, 35 are bent and slanted in opposite directions, respectively, as shown in FIG. 9, when the straps 20, 30 are assembled. A partial top view of the support grid 7 corresponding to FIG. 9 is shown in FIG. 10. As seen from FIG. 10, the outer extremities of the mixing vanes 23, 25, 33, 35 are close to the fuel rods 9 depicted by dash-and-two-dot lines, but not close enough to come into contact with the fuel rods 9. In these structures, since a coolant stream which flows through the support grid 7 impinges on the mixing vanes 23, 25, 33, 55, the coolant stream is agitated to be stirred and make the temperature distribution therein uniform. SUMMARY OF THE INVENTION However, in the conventional support grid 7, as clearly shown in FIGS. 8 and 10 in particular, welding apertures 26, 36 are formed by an inner side of the mixing vanes 23, 25, 33, 35, respectively. Since the coolant stream passes through these welding apertures 26, 36 without impingement or interference, no agitation of the coolant passing through the welding apertures can be expected and the agitating and mixing function of all the mixing vanes is therefore not sufficient to achieve the desired effects. Accordingly, an object of the present invention is to provide a support grid with mixing vanes for a fuel assembly which do not increase a pressure drop in the coolant flow and which further provides improved agitating and mixing functions. In order to accomplish the object described above, according to the present invention, in a support grid of a nuclear reactor fuel assembly wherein the support grid has a plurality of first straps which are made of thin metal band plate, and a plurality of second straps which are made of thin metal band plate, the first and second straps are each provided with slits extending widthwise for receiving the other of the straps and assembled so as to receive opposite straps in the slits of each strap to cross each other and thereby form an eggcrate structure, mixing vanes integrally formed on a side edge of the first and second straps adjacent to a crossed area and in alignment with a slit, each of the mixing vanes being slanted so as to be adjacent to a nuclear reactor fuel rod to be placed through a grid cell of the eggcrate structure, the slanted portion of the mixing vane being shaped so as to maximize an area of the slanted portion of the mixing vane projected onto a plane perpendicular to coolant flow direction. Preferably, the mixing vanes are shaped such that they have a welding aperture formed in a base portion at the side of the slit, and a bend line extending parallel to a longitudinal axis of the strap is located closer to a distal end side than the welding aperture and a curved outer edge of the mixing vanes is located at the distal end side of the bend line. Furthermore, in place of the above, the shape of the mixing vane can be formed such that the bend line at which the slanted surface of the mixing vane begins is slanted with respect to the longitudinal axis of the strap so as to avoid the welding aperture and the length of the mixing vane of the first strap is larger than the length of the mixing vane of the second strap. |
063046326 | description | DETAILED DESCRIPTION OF THE INVENTION According to a general characteristic of the invention, the law of displacement is a continuous curve exhibiting point symmetry with respect to the point whose temporal coordinate is equal to one-half of the duration of photography, and whose spatial derivative of the temporal variable exhibits two (preferably linear) portions symmetrical with respect to an axis of symmetry passing through the center of the area of displacement of the grid. In addition, the grid is displaced according to the law of displacement at a high rate of displacement in the vicinity of the starting position and of the ending position. According to one mode of implementation of the invention, the high rate of displacement is between about three times and about ten times the value of the ratio between the area of displacement and the duration of photography. In other words, this high rate of displacement is between about three times and about ten times the value of a linear rate of displacement of the grid between the starting position and the photography position. In a preferred mode of implementation of the invention, the continuous curve is formed of two portions symmetrical with respect to the point whose temporal coordinate is equal to one-half of the said duration of photography, each of these portions representing a profile of evolution of the variable "position," a function of the square root of the variable "time." In FIG. 1, the reference F designates the focal point of an X-ray tube emitting a beam RX of X-rays in the direction of an object to be radiographed OBJ. The radiographic images are received on a receiver RI, comprising for example a CCD sensor. The receiver RI is connected to processing means MT structured about a microprocessor and the radiographic images may be visualized on a display screen ECR. Between the object OBJ to be radiographed and the receiver RI, there is arranged an anti-diffusion grid GR movable in translation substantially perpendicular to the radiation emitted, i.e., in the direction XX of FIG. 1. This grid is composed of a plurality of plates LM, all directed toward the focal point F. These plates, typically spaced on the order of 0.3 mm, make it possible to absorb the radiation diffused by the object and to allow only the direct radiation to pass. In order to avoid visualization of the plates LM on the images obtained the grid GR is displaced in rectilinear translation in its plane, i.e., in the direction XX, according to a profile, predetermined at the time that each image is taken, between a starting position and an ending position. If the "period" of the grid is designated as the distance separating the edge of one plate from the edge of the immediately adjacent plate, i.e., a distance equal to the thickness of the plate plus the distance between two adjacent plates, it has been observed that one of the principal reasons producing visualization of the plates on the images obtained is the fact that the number of grid periods that pass between the X-radiation and each pixel of the image receiver is not a whole number. In other words, the part of a grid period that does not pass between the radiation RX and a pixel of the image receiver renders the corresponding plate of the grid visible on the image obtained. In addition, it has been observed that the act of displacing the grid at a high rate of displacement in the vicinity of the starting position and of the ending position makes it possible to reduce the visualization of traces of the grid on the radiographed image, because this contributes to a reduction of the time of exposure of incomplete periods of the grid situated at the ends of the latter. However, it is not necessary to provide a high rate of displacement in the center of the area of displacement because, in this zone, complete grid periods pass between the X-radiation and a considered pixel of the image receiver. In other words, because of the periodicity of the grid, the intensity of X-radiation reaching the image receiver is the temporal integral over the period of exposure of the incident energy multiplied by a coefficient of attenuation. It is this temporal integral that makes it possible to render the incomplete periods of the grid visible on the image, and to eliminate the traces of plates corresponding to complete periods of the grid that have been displaced between the radiation RX and the pixels of the sensor RI. In a general way, the displacement profile of the grid between the starting position X0 and the ending position XM during the exposure time TP of each image (TP=T1-T0) is a continuous curve exhibiting point symmetry with respect to the point whose temporal coordinate is equal to TP/2, and whose spatial derivative dt/dx of the temporal variable has two portions symmetrical with respect to an axis of symmetry passing through the center of the area of displacement of the grid. The rate of displacement V0 in the vicinity of the starting position and of the ending position must be high, for example between about three times and about ten times the value of the ratio (XM-X0/TP) between the area of displacement and the duration of photography, i.e., three to ten times higher than the value of a linear rate of displacement. The example described in FIGS. 2 and 3 shows that this continuous curve CB is formed of two portions symmetrical with respect to the point SP of temporal coordinate TP/2. Each of these portions CB1 and CB2 represents a profile of evolution of the variable "position" (X), which is a function of the square root of the variable "time" (t). More precisely, the equation of the portion CB1 is given by the formula (1) below: EQU X(t)=Ao+bct-TC+L 0for t.ltoreq.TP/2 (1) while the equation of the portion CB2 is given by the formula (2) below: EQU X(t)=Al-b-ct+TC+L 1for t=TP/2 (2) In these formulas, Ao, Al, b, c, TC0, and TC1 are constants making it possible to adjust the position of the grid to the value X0 for the instant T0 and to the value XM for the instant T1, and making it possible to join the two portions CB1 and CB2 at the point SP. So as to obtain the high rate V0 at the instant T0 of start of photography, a preliminary displacement of the grid is provided between the origin and the position X0 according to a displacement curve CB0 having a parabolic shape. In addition, after the time T1 of end of imaging, i.e., when the grid has reached the position XM, the latter returns to the zero position by a linear decline (terminal portion CB3). The spatial derivative of the temporal variable of the curve CB illustrated in FIG. 2 is represented in FIG. 3. It is made up of two linear portions CP1 and CP2 symmetrical with respect to the axis AS passing through the midpoint (XM-X0)/2 of the area of displacement. Such a curve profile makes it possible to reduce the visibility of the plates of the grid on the images obtained and thus to improve their quality to facilitate in particular the detection of micro-calcifications, and this regardless of the duration of exposure. In addition, the invention is free of any alternating movement, which makes it less sensitive to the mechanical parameters. This improvement in the quality of images necessitates no modification of processing software or of image acquisition. Additionally, displacement of the grid is generally obtained by a stepping motor which, by nature, generates mechanical oscillations during displacement. When the frequency of the oscillations corresponds to the frequency of spacing of the plates of the grid, screen peaks are obtained which are reflected in an increased visibility of the plates on the images. It was observed that the displacement profile according to an embodiment of the invention minimized this undesirable effect. Lastly, although the invention permits distinct improvement in the quality of images with a profile of spatial derivative dx/dt made up of portions which are not necessarily linear, the visibility of the plates is reduced still further if the spatial derivative dx/dt has such linear portions. Various modifications in structure and/or steps and/or function may be made by one skilled in the art without departing from the scope of the invention. |
description | This patent application is a divisional of and claims priority to U.S. patent application Ser. No. 15/451,719, filed Mar. 7, 2017, which claims priority to U.S. Provisional Patent Application No. 62/305,272, filed Mar. 8, 2016, both of which are herein incorporated by reference in their entirety. The present disclosure generally relates to a fission product getter device and a method of fabricating a fission product getter device. A fission product getter apparatus is disclosed, in accordance with one or more illustrative embodiments of the present disclosure. In one illustrative embodiment, the fission product getter apparatus includes a getter body including a volume of getter material and having a void structure. In another illustrative embodiment, the getter material is reactive with a nuclear fission product contained within a fluid flow from a nuclear fission reactor. In another illustrative embodiment, the getter body has a determined volume parameter sufficient to maintain flow of the fluid through the void structure of the getter body for a selected period of time. In another illustrative embodiment, the determined volume parameter of the getter body has the determined volume parameter and provides a void volume within the getter body sufficient to maintain expansion of the getter body below a selected expansion threshold over a selected period of time. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is not intended to be in any way limiting. Other aspects, features, and advantages of the devices and/or processes and/or other subject matter described herein will become apparent in the teachings set forth herein. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. The present disclosure is directed to various embodiments of a getter element for removing one or more fission products from a gas and/or liquid flow, such as fission products formed during a reaction process within a nuclear fuel of a nuclear reactor. The getter element includes one or more internal passages that facilitate a continuous throughput of the fluid (liquid and/or gas) flow, and also includes getter material that chemically reacts with a target fission product to remove the fission product from the flow. The disclosed technology is suitable for implementation in a variety of nuclear reactors including without limitation fast nuclear reactors, breeder reactors, breed and burn reactors, and/or in some cases traveling wave reactors. The present disclosure is further directed to various methods of forming the getter element. FIG. 1 illustrates an example nuclear fission reactor 130 with a fast nuclear reactor core 132. The fast nuclear reactor core 132 is disposed in a reactor vessel 140 surrounded by a guard vessel 136. In one implementation, the fast nuclear reactor core 132 includes a nuclear fission igniter (not shown) that provides neutrons for the fission reaction of fissile nuclear fuel. The nuclear fission reactor 130 includes a number of fuel assemblies (e.g., a fuel assembly 138 in View B), and each fuel assembly further includes multiple fuel elements, which are also referred to herein as fuel pins. In one implementation of the disclosed technology, the individual fuel pins each further include a mechanism for collecting one or more fission products from an input stream, as described below with reference to Views B, C, and D. The fast nuclear reactor core 132 typically contains a coolant, such as a pool of coolant (such as liquid sodium) or loops through which coolant may flow throughout the nuclear fission reactor 139. In some reactors, there exists a reservoir of coolant in headspace 148 above the fast nuclear reactor core 132. Heat exchangers (not shown) may rest near or in contact with the reservoir of coolant to aid in transporting heat away from the fast nuclear reactor core 132. Referring to View A, the nuclear fission reactor 130 includes a number of fuel assemblies shown in greater detail in View B (e.g., the fuel assembly 138). Each fuel assembly further includes multiple fuel pins, such as a fuel pin 120 (shown in View C). View B illustrates an array 142 of nuclear fuel assembly devices suitable for use within the fast nuclear reactor core 132. Each assembly includes multiple fuel pins (e.g., a fuel pin 120). Although other device shapes and array configurations are contemplated, the example nuclear fuel assembly devices of FIG. 1 each include a solid hexagonal tube surrounding. Non-hexagonal tubes may also be used on some implementations. Components of an individual fuel assembly device 138 within the array 142 are shown in further detail in Views C and D. As shown in View C, the nuclear fuel assembly device 138 surrounds a plurality of elongated fuel elements, such as the fuel pin 120. When nuclear fission occurs within a fuel pin, fission products are created that can contribute to a building pressure within the pin. In some reactors, fuel pins are designed to include a large plenum area to accommodate this pressure at high burn-ups. Other reactors may include fuel pins designed to vent gases to relieve pressure, such as venting to allow the fission products to flow into contact with a coolant reservoir in the headspace 148. Since some fission products may be volatile, this venting can pose a risk. Both venting and non-venting fuel pin designs can benefit from the herein-disclosed technology, which generally provides tools and techniques for removing of one or more fission products from a fluid flow within the nuclear fission reactor 130. Components of the fuel pin 120 are shown in greater detail in View D, described below. In one implementation, the tubular structure of each of the individual fuel assembly devices, such the nuclear fuel assembly device 138, allows coolant to flow past the fuel pins through interstitial gaps between adjacent tube walls. Each tube also allows individual assembly orificing, provides structural support for the fuel bundle, and transmits handling loads from a handling socket to an inlet nozzle. Fuel pins typically consist of multiple nuclear fuel rods (such as uranium, plutonium or thorium) surrounded by cladding (and sometimes an additional barrier and/or liner) to separate the radiative material and the coolant stream. Individual pins of the nuclear fuel assembly devices 138 in the fast nuclear reactor core 132 can contain fissile nuclear fuel and/or fertile nuclear fuel depending on the original nuclear fuel rod material inserted into the pin and the state of breeding within the pin. An example fuel pin 120 is shown in greater detail in View D. The fuel pin 120 includes fuel 122, a getter element 100, and an optional plenum area 124. The getter element 100 stores a material (not shown) that is chemically reactive with a fission product 110 included in an input stream 108 received from the fuel 122. For example, the input stream 110 includes one or more fission products created during nuclear fission of the fuel 122. The getter element 100 includes at least one internal fluid flow path that facilitates continuous transmission of a gas and/or liquid through the getter element 100. The fluid flow path may be, for example, one or more elongated channels, interconnected pores, microfluidic structures, etc. The fluid flow path through the getter element 100 provides a surface area internal to the getter element 100 that may chemically react with the fission product 110 to remove the fission product 110 from the input stream 108, and thereby create an output stream 112 with a lower density of the fission product 110 than the input stream 108. In various implementations, the fission product 110 may be volatile or non-volatile. Although plenum area 124 is shown above the getter element 100, and the getter element 100 above the fuel 122, it is to be appreciated that these components may be placed relative to each other in any suitable order and manner. The various examples of fuel pins (e.g., the fuel pin 120) described herein may represent venting or non-venting fuel pins. In a venting fuel pin, the plenum area 124 is, at times, in fluid communication with the headspace 148 in a nuclear reactor, or other appropriate gas reservoir. For example, the fuel pin may include various vents or openings that facilitate fluid communication between the plenum 124 and the headspace 148. As used herein, the term “getter element” (as in the example getter element 100) is meant to refer to any structure including a “getter material” capable of chemically reacting with a fission product and thereby removing a quantity of the fission product from an input stream. Getter material may be incorporated within or formed into a “getter body.” For example, a getter body may be a free-standing structure, collection of particulates (e.g., powder), small capsules or pellets. The getter body may include the getter material alone or the getter material in addition to one or more other non-getter materials that do not react with the fission product 110. In some cases, the getter element includes a getter body and also includes a container for holding the getter body. The getter element 100 includes one or more channels for placing the input stream 108 in fluid communication with a getter body. In one implementation, the getter body has characteristics designed to maximize surface area of contact between a getter body and the input stream 108 passing through the getter element 100. For example, the getter body may include pores or other channels that increase its total surface area. In additional and/or alternative implementations, the getter element 100 includes a container with one more diffusing elements for directing the input stream 108 into contact with the getter body. Fluid space within or throughout, the getter element 100 and the getter body allows the input stream 108 to contact the getter material and chemically react to remove the fission product 110 from the input stream. In some implementations, the getter element 100 is not included in a fuel pin (e.g., the fuel pin 120), as shown. Rather, the getter element 100 is positioned elsewhere within the nuclear fission reactor 130 at a position that is accessible to targeted fission products. For example, the getter element 100 may be positioned above the reactor core within the reactor vessel and/or in headspace 148 of the nuclear fission reactor 130 to receive and react with fission product(s) fluid exiting the fuel subassemblies. Notably, certain structures of the example nuclear fission reactor 130 have been omitted from FIG. 1, such as coolant circulation loops, coolant pumps, heat exchangers, reactor coolant system, etc., in order to simplify the drawing. Accordingly, it should be understood that the example nuclear fission reactor 130 may include additional structures not shown in FIG. 1. FIG. 2 illustrates an example getter body 200 formed by a volume including getter material 204. The getter body 200 includes at least one through-channel (e.g., a through-channel 220) for transmitting a flow of an input stream 208 through the getter body 200. The through-channel(s) may extend along a longitudinal length of the getter body 200 (e.g., in the direction of the input stream 208, as shown) and/or may extend in one or more other directions so as to facilitate transport of gas from one side of the getter body 200 to another opposite side. The through-channel(s) of the getter body 200 assume a variety of different forms in the various implementations disclosed herein. Suitable forms include without limitation interconnected voids or pores, engineered pathways, and/or separations between discrete particles (e.g., in implementations where the getter body is a loose powder as described further with respect to FIG. 3, below). The input stream 208 may be gaseous, liquid, or a combination thereof, and further includes a fission product 210, which may be gaseous, liquid, solid, dissolved, suspended, or a combination thereof. In one implementation, the getter material 204 includes one or more materials that are chemically reactive with the fission product 210. In this regard, as the input stream 208 containing the fission product 210 contacts the getter body 200, the getter material 204 chemically reacts with the fission product 210 to form a byproduct that is retained within the getter body 200 while the remainder of the input stream 208 exits or moves past or through the getter body 200 as an output stream 212. Thus, the output stream 212 contains less fission product 210 than the input stream 208. The process of removing the fission product 210 from the input stream 208 via chemical reaction with a getter material is also referred to herein as an “uptake” (e.g., the getter body 200 “uptakes” the fission product 210). In one implementation, the getter body is specifically engineered to provide for uptake of substantially all of a select fission product produced within a fuel pin over a period of time, such as over the effective lifetime of the fuel pin. As used herein, uptake of “substantially all” of a select fission product refers to uptake of at least 95% and some cases more than 95% of the select fission product. The through-channel(s) of the getter body 200 serve multiple purposes. First, the existence of these channel(s) helps to relieve pressure in areas of the getter body 200 and/or corresponding fuel pin by allowing certain content (e.g., inert gases) to escape. Second, the existence of these through-channel(s) provides areas for the getter body 200 to expand into, such as at high burnup rates, thereby decreasing a likelihood of potential damage to associated areas of an associated fuel pin and or fuel subassembly. In one implementation, the through-channel(s) have a sufficient volume to maintain a through-flow above a preselected flow level despite expansion of the getter material within a predetermined range of thermal expansion. Third, the existence of these through-channel(s) increases available surface area that can react with the fission product 208. In one implementation, the surface area of these through-channel(s) is specifically designed to facilitate uptake of a specific calculated quantity of the fission product 208, such as substantially all of the fission product 208 expected to be produced by an associated fuel pin over a given interval of time. The fission product 210 of the input stream 208 may be volatile or non-volatile fission product. Example volatile fission products include without limitation: cesium (Cs) or a Cs-based compound (e.g., Cs2, CsBr, Cs2I2, CsI, etc.), rubidium (Rb) or a Rb-based compound (Rb, Rb2, RbI, RbBr, etc.), strontium (Sr) or a Sr-based compound (Sr, etc.), and iodine (and it compounds). Example non-volatile fission products include without limitation Zirconium, Molybdenum, Neodymium, etc. The getter material 204 includes any material known in the art as chemically reactive with the fission product 210. Although a variety of materials may be suitable getter materials, some implementations of the disclosed technology include metal oxides within the getter material, such as one or more of zirconium oxide (e.g., ZrO2), titanium oxide (e.g., TiO2), molybdenum oxide (e.g., MoO2, MoO3), niobium oxide (NbO2, Nb2O5), tantalum oxide (e.g., Ta2O5), etc. Because the getter materials considered do not show equivalent reactivity with all fission products of concern, the getter material may also be composed of a mixture of components, with the mixture composition tailored to maximize reaction between the getter material and one or more targeted fission products (e.g., 75%-Ta2O3/25%-Nb2O3 mixtures). Although in some embodiments it may be beneficial to have these disparate components intermixed, in others it may be beneficial to have discrete layers to selectively remove targeted fission products from the fluid at preferential stages to prevent potential detrimental interactions with subsequent layers of the getter material. The getter material may also include one or more non-reactive components such as binders, structural stabilizers, etc. In one implementation, the getter material 204 includes one or more materials that react with cesium (Cs) or a Cs-based compound. In the same or another embodiment, the getter material 204 includes at least one material that reacts with Rubidium (Rb) or a Rb-based compound. In the same or another embodiment, the getter material 204 includes at least one material that reacts with Iodine or an iodine-based compound. In FIG. 2, the getter body 200 is shown to be a cylindrically-shaped solid structure including a void structure 206 (e.g., pores). In one implementation, the void structure 206 includes randomly or regularly distributed pores forming an open pore structure. The distributed pores may be selectively engineered in size, shape, inter-connectivity, structural stability, distribution schema, etc. There exist a variety of suitable processes for forming the void structure 206 and/or other channels within the getter body including without limitation sacrificial templating, additive manufacturing, template replication, and direct foaming. These methods are discussed in greater detail below. In one embodiment, the void structure 206 of the getter body 200 is formed via a sacrificial templating process. For example, the void structure 206 may be formed by mixing the getter material 204 with a void-forming material. Voids are formed by removing (e.g., burning off or dissolving) the void-forming material. As a result of the removal of the void-forming material, voids (e.g., pores or cells) are formed throughout the volume of getter material 204 of the getter body 200. One example of implementation of a sacrificial templating procedure is described by Andre R. Studart et al. in Processing Routes to Macroporous Ceramics: Review, J. Am. Ceram. Soc. 89 [6] 1771-1789 (2006), which is incorporated herein by reference in the entirety. A variety of treatments may be suitable for processing the void material removal from the getter material and form the void structure, including any one or more of dissolving, heat treatment (e.g., during sintering or during a dedicated burn-off cycle), etc. Further details of example sacrificial templating processes are discussed in greater detail with respect to FIGS. 14-20. In another embodiment, the void structure 206 of the getter body 200 are formed via an additive manufacturing process. For example, the getter body 200 may be fabricated via a three-dimensional printing process. In this regard, the void structure 206 of the getter body 200 may be directly engineered and the formation of which may be directly controlled via the manufacturing process. Selective laser sintering, used to three-dimensionally print materials, may be additionally or alternatively appropriate and is generally described in U.S. Pat. No. 4,863,538, filed on Oct. 17, 1986, which is incorporated herein by reference in the entirety. Further details of example additive manufacturing processes are described with respect to FIG. 20-23, below. In another embodiment, the void structure 206 of the getter body 200 is formed via a template replication process. For example, pores may be formed through impregnation of a void structure (e.g., cellular or porous structure) with a getter material suspension (or precursor solution), resulting in a volume of porous getter material exhibiting the same (or nearly the same) morphology as the original porous material. One example of a replica procedure is described by Andre R. Studart et al. in Processing Routes to Macroporous Ceramics: Review, J. Am. Ceram. Soc. 89 [6] 1771-1789 (2006), which is incorporated above by reference in the entirety. Basic procedural steps in template replication are described with respect to FIG. 24. In another embodiment, the void structure 206 of the getter body 200 is formed via a direct foaming process. For example, the void structure 206 of the getter body 200 may be formed through incorporation of a gas (e.g., air) into a suspension or liquid form of the getter material (or a precursor of the getter material), which serves to establish a foam structure within the suspension or liquid. The material then undergoes a setting or solidifying step, which serves to lock in the void structure 206 formed within the foam. One appropriate example of a direct foaming procedure is described by Andre R. Studart et al. in Processing Routes to Macroporous Ceramics: Review, J. Am. Ceram. Soc. 89 [6] 1771-1789 (2006), which is incorporated above by reference in the entirety. Basic procedural steps in direct foaming are described with respect to FIG. 25. In yet other embodiments, the void structure 206 of the getter body 200 is formed by other physical methodologies (e.g., mechanical grinding, etching laser ablation, etc.), or chemical methodologies such as chemical etching. Notably, any two or more of the above-described techniques (e.g., sacrificial templating, additive manufacturing, template replication, direct foaming, chemical/physical etching, grinding, ablation, etc.) may also be used in combination to create the void structure 206. For example, a sacrificial templating process may be initially used to create small voids in the getter body 200 and a machining process may thereafter be used to create larger voids, such as near a fission gas inlet of the getter body 200. FIGS. 3A-3B illustrate simplified schematic views of an example getter element 300 including a getter body 302 configured to rest between and/or attach to capping elements 314a, 314b at either end. Although other structures are also contemplated (e.g., with respect to FIGS. 4-5), the getter body 302 of FIG. 3 is a free-standing, solid element including a void structure 306. In operation, the capping elements 314a, 314b serve to provide mechanical support to the getter body 302 and to further facilitate venting of an input stream 308 through the getter element 300. In some implementations, the main getter body 302 is not a solid free standing structure. For example, the main getter body 302 may be in particulate form (e.g., powder) or be a collection of elements (e.g., solid pellets or small capsules further storing particulates). In these implementations, the capping elements 314a, 314b may be used in combination with a container or supporting shell for containing and further supporting the main getter body 302. The capping elements 314a, 314b are made from heat-stable materials that resist deformation when subjected to the high temperatures and neutron fluxes of a nuclear reactor core. Ideal candidate materials may also be non-reactive with fission products (e.g., a fission product 310) included in the input stream 308. Example suitable materials for the capping elements 314a, 314b include, for example steels, refractory metals/alloys, or structural ceramics. In FIGS. 3A and 3B, the one or more capping elements 314a, 314b are formed from a porous material. For example, the one or more capping elements 314a, 314b may include a porous metal plate 313 (e.g., porous metal disk as shown in FIG. 2B). Other porous structures, such as vents, mesh-like material, etc., are also contemplated. In one implementation, the capping elements 314a, 314b are solid structures that include a plurality of through-holes, such as drilled holes. The holes may be a variety of sizes and distributions depending on specific implementation details such as the desired flow rate, specific getter material, targeted fission product(s), etc. FIGS. 4A and 4B illustrate simplified schematic views of another example getter element 400 including a getter body 402 configured to rest between and/or attach to capping elements 414a and 414b. The getter body 402 has a void structure 406 and includes a getter material 404 for reacting with a fission product 410 included within an input stream 408, thereby reducing a concentration of the fission product 410 in an output stream 412 as compared to the input stream 408. Unlike the porous structure of the capping elements in FIGS. 3A, 3B, the capping elements 414a, 414b are vented metal plates 415 (e.g., vented metal disk). Suitable construction materials and other details of the capping elements 414a, 414b may be the same or similar to that described above with respect to FIGS. 3A, 3B. FIG. 5A illustrates a simplified schematic view of an example support structure 500 suitable for use in a getter element. The support structure 500 includes a container portion 521 attached to endcaps 514a, 514b. In operation, the support structure 500 may provide mechanical support to a getter body and facilitate venting of an input stream 508 through the getter element and/or getter body. The support structure 500 may support a solid, free-standing getter body (e.g., as in the main getter body portion 402 of FIG. 4A); alternatively, the support structure 500 may support a getter body that is in particulate form (e.g., powder) or otherwise represented as a collection of free-standing elements (e.g., solid pellets or small capsules further storing particulates). The container portion 521 may be formed from any material that provides thermal and chemical and structural stability in the presence of fluid flow, neutron irradiation, and fission products of a selected nuclear reactor environment. In one embodiment, the container portion 521 is formed from steel. Other suitable container materials could include refractory metals or alloys, as well as structural ceramics. Although not shown in FIG. 5, the container portion 521 may include a plurality of openings about its circumference to allow for fluid and/or gas to flow through the sides of the container portion 521 as well as through vents 515 or porous openings in the endcaps 514a, 514b. In various implementations, any suitable number, size, location, and/or distribution of vents 515 in the capping element 514a may be used as appropriate for design and/or safety considerations. FIG. 5B illustrates a simplified schematic view of another example support structure 502 for positioning a getter body (e.g., as in the getter body 200 of FIG. 2 or 300 of FIGS. 3A and 3B) within a fuel pin 510. The support structure 502 includes porous endcaps 514a, 514B and a central body 516 with number of peripheral openings (e.g., an opening 518) in a cylindrical sidewall 520 to allow for intake of a fluid flow into a center of the support structure 502 and within the getter body (not shown). In one implementation, a width W1 of the support structure 502 is slightly less than a width W2 of the fuel pin 510 so as to allow a fluid flow to bypass the endcap 514a and to enter the support structure 502 through one or more of the openings (e.g., the openings 518) in the cylindrical sidewall 520. FIG. 6 illustrates a schematic view of a portion of another example support structure 600 suitable for use in a getter element. The support structure 600 includes a container portion 617 and a diffuser assembly 609. The diffuser assembly 609 further includes a diffuse capping portion 614 and a diffuse channel portion 622 (e.g., an elongated central channel). In operation, a getter body (not shown) is stored within the container portion 617. For example, the getter body may surround or partially surround the diffuse channel portion 622. The diffuser assembly 609 helps to bring gas or liquid of an input stream 608 into fluid communication with the getter material of the getter body. For example, the diffuse capping portion 614 and the diffuse channel portion 622 includes openings (e.g., pores, vents, etc.) that provide fluid flow paths into and/or through the getter body. The diffuse channel portion 622 is shown in FIG. 6 as a singular central channel with a number of holes allowing gas to freely flow between areas internal to the diffuse channel portion 622 and areas external to the diffuse channel portion 622. However, it is to be appreciated that the support structure 600 may include a plurality of channels in lieu of or in addition to the diffuse channel portion 622. For example, the diffuser assembly 609 may include other channels distributed throughout other regions of the container portion 617. In some cases, the diffuser assembly 609 includes a gas transmission channel that surrounds the getter body, such as a porous annular channel surrounding the getter body. Due to the uptake of a fission product 610 into the getter body, the getter body, over time, accumulates the fission product 610. Accumulation of fission product 610 within the getter body may result in the reduction of fission gas flow through the container portion 617 and/or throughout the getter body within the container portion 617. In some instances, where accumulation is severe, one or more porous structures of the getter body may be become blocked. In such cases, the diffuser assembly 609 may help to maintain a flow of the input stream 608 through a getter material (not shown) regardless of this blockage. In addition, the diffuser assembly 609 may serve to ensure fluid flow through fluid flow paths within the getter body in the event of volumetric expansion of the getter material of the getter body. The diffuse capping portion 614 may take on a variety of forms, such as that of a porous metal or ceramic plate or a vented plate with a set of vent holes. The diffuse channel portion 622 is also porous and may be, for example, a porous metal rod or a metal rod with a set of vent holes. Some non-metal materials (e.g., ceramics) may also be suitable for forming all or various components of the diffuser assembly 609. FIG. 7 illustrates an end-on view of a portion of another example getter body 702 with void structures 706. In one implementation, the getter body 702 is sized and shaped to rest within one of the corresponding support structures 500 or 600 of FIGS. 5 and 6, respectively. In operation, the illustrated end of the getter body 702 may receive an input stream including a fission product. When the input stream contacts the getter body 702, getter material 704 in the getter body 702 chemically reacts with one or more fission products in the input stream, removing those fission product(s) from the stream. In FIG. 7, the getter body 702 is a free-standing solid structure. For example, the getter body 702 may be a porous sintered metallic or ceramic structure. Although other arrangements are contemplated, the void structures 706 of the getter body 702 are arranged such that the size varies as a function of position within the getter body 702. For example, size of the void structures 706 may generally decrease as a function of radial distance from the center of the getter body 702. For example, the distribution of void structures 706 may be influenced by the size and/or weights of void-forming structures utilized during fabrication of the getter body 702. In this regard, void-forming structures (e.g., such as those described below with respect to FIGS. 13-20) may, when mixed with the getter material 704, act to self-sort and form a distribution (e.g., gradient distribution) via a settling and/or agitation process. The void structures 706 of the getter body 702 may be distributed throughout the getter body 702 in any pattern or distribution. In some implementations, the void structures 706 include pores in greater size near a fission gas inlet and pores smaller in size near a fission gas outlet. FIG. 8 illustrates an end-on view of a portion of another example getter body 802 that is cylindrical in shape and includes multiple concentric regions 804a, 804b of getter material separated from one another by transmission pathways (e.g., an annular-shaped void 806). The illustrated arrangement may help to maximize a surface area of contact between the getter material of the getter body 802 and an input stream (not shown) that is directed through the getter body 800. In one implementation, the concentric regions 804a, 804b of getter material are solid structures, such as sintered metallic or ceramic structures. In another implementation, the getter body 802 is formed via a powder that fills each of a number of porous concentric shells of a getter container. A variety of other structures are also contemplated (some of which are described with respect to the following figures). FIG. 9 illustrates a cross-sectional view of a fuel pin 920 of a nuclear reactor equipped with an example getter element 900. The getter element 900 is shown disposed within fuel pin 920 and positioned to receive an input stream 908 (e.g., fission gas) from nuclear fuel 922 of the fuel pin 920. For example, the getter element 900 is disposed (alone or in combination with other getter elements) at a location upstream of the nuclear fuel 922 and origination point of the input stream 908, but downstream of a fission gas plenum 924. In another implementation, the getter element 900 is positioned within the fission gas plenum 924 (e.g., with or without space of the plenum on either or both ends of the getter element 900). Capping elements 914a and 914b provide barriers between the getter element 900 and the immediately adjacent structures. In one implementation, the separation caps 914a, 914b are porous endcaps (e.g., plates with pores or vents). In another implementation, the separation caps 914a and 914b are valves that open under pressure generated by the input stream 908. The getter element 900 includes a getter body (not shown) including a getter material that reacts with one or more volatile or non-volatile fission products 910, resulting in an output stream 912. The output stream 912 exiting the getter element 900 has a lower volatile fission product content level than the input stream 908 entering the getter element 900. In one embodiment, the output stream 912 is vented from the fuel pin 920, such as through one or more pin vents of the fission plenum 924. In some implementations, the getter material reacts with one or more volatile fission products 910 in the input stream 908 such as cesium, rubidium, strontium, etc. In additional or alternative implementations, the getter material of the getter body reacts with one or more non-volatile fission products. FIG. 10 illustrates a cross-sectional view of a fuel pin 1020 of a nuclear reactor equipped with two example getter elements 1000a, 1000b arranged in series between nuclear fuel 1022 and a fission plenum 1024. In operation, fission gas from the fuel 1022 is passed via an input stream 1008 through the getter elements 1000a, 1000b in series. Within each of the getter elements 1000a, 1000b, one or more fission products 1010 within the input stream 1008 undergo chemical reactions with getter material, thereby cleaning or partially cleaning the input stream 1008 to reduce a concentration of the fission product 1010 in an output stream 1012. Separation caps 1014a, 1014b, 1014c are barriers that are either porous or capable of selectively opening, such as under pressure of the input stream 1008. In one implementation, the first getter element 1000a includes a first getter material for targeting the uptake of a first fission product, while the second getter element 1000b includes a second getter material for targeting the uptake of a second fission product. For example, the first material of the first getter element 1000a may include a getter material targeted for uptaking a first element or compound, while the second material of the second getter element 100b may include a getter material targeted for uptaking another compound including the first element and/or another different element. In one exemplary implementation, one of the two getter elements 1000a and 1000b includes a getter material for uptake of cesium, such as niobium or titanium oxides, while the other one of the two getter elements 1000a and 1000b includes a different getter material for uptake of iodine, such as silver, copper, or barium. It is noted herein that fuel pin 1020 of FIG. 10 is not limited to two getter elements or the materials listed above, which are provided merely for illustrative purposes. Other implementations may include fewer or greater than two getter elements. It is noted herein that the shape of the one or more getter elements (e.g., 1000a, 1000b) of the present disclosure is not limited to the cylindrical shape depicted in FIGS. 1-10. The one or more getter elements 1000 of the present disclosure may take on any general geometrical shape. In other implementations, the one or more getter elements take on a variety of shapes including without limitation hexagonal prism shapes, parallelepiped shapes, triangular prism shapes, helical shapes, conical shapes or the like. In one embodiment, the one or more getter elements 1000 contained within the fuel pins 1020 are structured so as to substantially conform to the internal shape of the fuel pins 1020. In this regard, the one or more getter elements 1000 may take on any shape known in the art based on the shape of the fuel pins 1020. It is noted that the getter element(s) (e.g., 1000a, 1000b) of the present disclosure may be adapted to operate in any nuclear reaction environment. The nuclear fuel contained within the fuel pin 1020 may include any fissile and/or fertile nuclear fuel known in the art including without limitation recycled nuclear fuel, unburned nuclear fuel, and enriched nuclear fuel. In one embodiment, the fuel 1022 includes a metal nuclear fuel and is used to form a core of a metal fuel nuclear reactor along with a plurality of other fuel pins. In one embodiment, metal fuel nuclear reactor is a fast reactor. For example, the metal fuel nuclear reactor may include a breeder reactor, such as, but not limited to, a traveling wave reactor. FIG. 11 illustrates a perspective view of a nuclear reactor core 1100 including a set of fuel assemblies (e.g., a fuel assembly 1130). Each fuel assembly further includes a set of fuel pins and each fuel pin includes one or more getter elements, as discussed previously herein. The structure and arrangement of the fuel assemblies of the reactor core may take on any form known in the art. In the example arrangement of FIG. 11, the fuel assemblies are arranged in a hexagonal array. It is noted that the arrangement depicted in FIG. 11 is not a limitation on the present disclosure and is provided merely for illustrative purposes. In some implementations, the fuel assemblies are arranged according to other shapes such as, but not limited to, a cylinder, a parallelepiped, a triangular prism, a conical structure, a helical structure and the like. FIG. 12 illustrates a top view of an example fuel assembly 1200 including a set of fuel pins (e.g., a fuel pin 1120). Each of the fuel pins is equipped with one or more getter elements for cleaning a fission gas to remove one or more volatile or non-volatile fission products. In FIG. 12, the fuel pins are cylindrically-shaped and are arranged in a close packed hexagonal array; however, this arrangement may take on other forms in other implementations. For example, the fuel pins 1220 of the fuel assembly 1200 may individually be shaped hexagonally, parallelepiped, triangular, helical, conical or the like. In other embodiments, although not shown, the fuel pins 1220 of the fuel assembly 1200 may be arranged in a rectangular array, a square array, a concentric ring array and the like. FIG. 13 illustrates example operations 1300 for forming a getter body for use in cleaning a fission gas output stream of a nuclear reactor. A determining operation 1302 determines an amount of fission product contained within a fluid flow output from a nuclear fission reactor core over a selected period of time. The selected period of time can be a single or multiple fuel cycles and may be the expected lifetime of a single fuel pin or fuel assembly of the reactor. It is to be appreciated that different fuel assemblies and/or fuel pins may have different expected lifetimes or fuel cycles which can be accommodated with different expected fission product determinations for different fuel elements. The amount of fission product contained within the fluid flow further corresponds to a specific amount of nuclear fission fuel consumed during the selected period of time which can be determined using any suitable neutronic methods and/or model of the present fuel type and expected neutronic environment of the fuel element (e.g., fuel burn up) over the specified period of time. A providing operation 1304 provides a getter process mixture that includes a getter material reactive with a fission product of the fluid flow output from the nuclear fission reactor core. An amount of the getter process mixture to use in forming the getter body is determined by operations 1306 and 1308, described below. Another determining operation 1306 determines a desired yield of a reaction product to be formed via a chemical reaction between the fission product and the getter material over the selected period of time. In one implementation, the desired yield of the reaction product is an amount calculated as resulting from a reaction between the getter material and substantially all of the fission product determined in the determining operation 1302. Based on the desired yield of reaction product, another determining operation 1308 determines a volume parameter of the getter process mixture that identifies an amount or volume of the getter process mixture needed to yield the desired amount of reaction product in the selected period of time. The determining operation 1308 may also determine not only the amount of the getter process mixture but also a desired volumetric measure or density of the getter material suitable for uptake of the volume parameter of the desired yield of a reaction product. Specifically, the reaction product, when uptake occurs has a volume that may decrease the void structure or increase the density of the getter material. By determining this volume of the desired yield of the reaction product (or predetermined amount of reaction product), a volumetric parameter of the getter material can be selected that matches or exceeds the determined volume of the desired yield of the reaction product to ensure that fluid flow through the getter material is maintained (which may be maintained at or above a selected flow rate or flow level) and/or volumetric swelling of the getter material stays within design boundaries. For example, the volumetric parameter of the getter material may include without limitation, any one or more of pore size, pore concentration, theoretical density of the getter material, mass ratio of getter material to sacrificial void forming structures, etc. A forming operation 1310 forms a getter body defined by the determined volume parameter. The getter body is formed by the getter process. In some implementations, the forming operation 1310 further entails placing the getter body within a container (e.g., forming a final “getter element” includes at least one channel or passageway for transmission of gas or liquid therethrough). In some implementations, the getter element includes a getter body in the form of loose powder, a plurality of pellets, particulates, etc., within a porous container. In other implementations, the getter body is formed by a number of chemical and/or physical processes that generate a solid (e.g., free-standing) structure, such as a solid structure including a number of interconnected pores or a plurality of void regions. Thus, the getter element may not always include a container. The getter body may include channels or pores of a variety of other shapes, such as elongated channels. In still other implementations, the getter body is formed by multiple different porous components (e.g., a plurality of free-standing porous pellets, porous diffusing components, etc.) FIG. 14 illustrates a series of example operations 1400 for preparing a getter process mixture and forming a getter body with a plurality of voids (e.g., pores). The example operations 1400 disclose void creation via use of sacrificial structures, which are structures that decompose (thereby forming ‘voids’ within the getter process mixture) upon thermal and/or chemical treatment. In other implementations (such as those described with respect to FIGS. 20-25 below), voids of the getter body are formed by other methodologies and/or other void-forming structures. For example, additive manufacturing, template replication, and direct foaming are all suitable methods for creating void-forming structures that do not utilize sacrificial void-forming structures. A selecting operation 1402 selects volume of getter material to be included in the getter process mixture. The getter material may include any single or combination of material known in the art suitable for chemically reacting with one or more volatile or non-volatile fission products in a nuclear reactor. In one embodiment, the getter material is provided in powder form. For example, the getter material provided in step 1402 includes, but is not limited to, a metal oxide powder. For instance, the metal oxide powder provided in step 1402 may include, but is not limited to, ZrO2, TiO2, MoO2, MoO3, NbO2, Nb2O5, Ta2O5, VO2, V2O5, and Cr2O3. Any of these and analogous materials have been shown to readily react with one or more volatile fission products including, but not limited to, Cs, CsBr, CsI, Rb, RbI, RbBr, or other Rb-compounds, Sr or Sr-based compounds, and iodine (and its compounds). In addition to one or more reactive materials such those described above, the getter material may also include one or more non-reactive components, such as binders and structural stabilizers. In one embodiment, the getter material includes a metal oxide powder with a select particle size, such as an average particle size between about 100 and 500 nm. In another embodiment, the getter material includes a metal oxide powder with an average particle size at or below 100 nm. For example, the getter material may include, but is not limited to, a volume of nanopowder having an average particle size below 100 nm. A providing operation 1404 provides a volume of void-forming structures (e.g., sacrificial void-forming structures) for combination with the getter material in the getter process mixture. In one embodiment, the void-forming structures include one or more organic materials known to undergo pyrolysis (e.g., chemical decomposition) at elevated temperature(s) in the absence of oxygen. For example, the organic materials may be selected so as to decompose at temperatures at or below an applied sintering temperature (e.g., reached during heat applied in a densifying operation 1408, described below). The organic material used to form the void-forming structures may be selected so as to breakdown at a temperature between 200 and 600° C. For instance, the void-forming structures may be formed from an organic material that decomposes at temperature below approximately 500° C. (e.g., 330-410° C.). In one embodiment, the sacrificial void-forming structures are formed from a synthetic organic material. For example, the void-forming structures may be formed from any synthetic organic material known in the art, such as, but not limited to, polyethylene (PE), polymethylmethacrylate (PMMA), polyvinyl chloride (PVC), polystyrene (PS), nylon, naphthalene and the like. In another embodiment, the void-forming structures are formed from a natural organic material. For example, the void-forming structures may be formed from any natural organic material known in the art, such as, but not limited to, gelatine, cellulose, starch, wax and the like. In still another embodiment, the void-forming structures break down upon chemical treatment. For example, the void-forming structures may be formed from one or more water soluble ionic compounds. In one embodiment, the void-forming structures include one or more salts. For instance, the salt-based void-forming structures may include, but are not limited to, NaCl, KCl, LiCl and the like. In another embodiment, the void-forming structures include one or more metal or ceramic compounds that react with one or more acidic leaching agents. Sacrificial templating using chemical treatment is discussed generally in H. Wang, I. Y. Sung, X. D. Li, and D. Kim, “Fabrication of Porous SiC Ceramics with Special Morphologies by Sacrificing Template Method,” J. Porous Mater., 11 [4] 265-71 (2004), which is incorporated herein by reference in the entirety. Sacrificial templating using chemical treatment is also discussed generally in H. Kim, C. da Rosa, M. Boaro, J. M. Vohs, and R. J. Gorte, “Fabrication of Highly Porous Yttria-Stabilized Zirconia by Acid Leaching nickel from a Nickel-Yttria-Stabilized Zirconia Cermet,” J. Am. Ceram. Soc., 85 [6] 1473-6 (2002), which is incorporated herein by reference in the entirety. Sacrificial templating using chemical treatment is also discussed generally in N. Miyagawa and N. Shinohara, “Fabrication of Porous Alumina Ceramics with Uni-Directionally-Arranged Continuous Pores Using a Magnetic Field,” J. Ceram. Soc. Jpn., 107 [7] 673-7 (1999), which is incorporated herein by reference in the entirety. In one example sacrificial templating method, a solid template is impregnated with a suspension including the getter material. The structure is solidified through one or more techniques known in the art (e.g., as explained in the above-referenced publications), and the template structure is removed, such as by acidic leaching. For example, a coral may be impregnated with hot wax, the wax may be cooled, and the coral can be leached out using a strong acidic solution. In another embodiment, the void-forming structures include one or more solids that undergo sublimation. For example, the sacrificial void-forming structures may include any solid that readily sublimes, such as, but not limited to, naphthalene. In this regard, the one or more solid sacrificial void-forming structures may sublime out of the getter process mixture to generate a porous structure. The void-forming structures are capable of producing a void structure with a volume sufficient to maintain a selected fission gas flow through the getter body. For example, the void-forming structures may create pores with a size distribution between 10 and 300 μm. The void-forming structures may have, but are not limited to, an average size of approximately 100 micrometer, 150 micrometer, 50 micrometer, 30 micrometer, etc., as appropriate for the determined void size. It is noted herein that the size range listed above is not a limitation on the present disclosure and is provided merely for illustrative purposes. The selected size and/or concentration of void-forming structures may depend on the desired size of voids of void-structure and the desired density of the resultant getter body. Moreover, the size of the void-forming structures may be selected so as to account for expected volumetric growth of the reactive material in the void structure. The void-forming structures provided in the providing operation 1404 may take on any shape known in the art, including without limitation those example shapes illustrated in FIGS. 15A-15D of the present disclosure. While much of the present disclosure focuses on solid void-forming structures, this is not a limitation on the present disclosure. Rather, it is noted herein that void-structures may also be liquid or gas form. For example, the void-forming structures may include water and oils that evaporate or sublimate out of the getter body to create void regions. In still other implementations, the void-forming structures are gaseous in form, such as gases injected into a liquid structure including the getter material (e.g., as in direct forming, a technique described below). A forming operation 1406 forms a getter process mixture that includes both the volume of getter material and the volume of the void-forming structures. For example, the getter material and void-forming structures may be mixed in any selected proportion to achieve a desired void-structure in a resulting getter body. In one implementation, a mass ratio of getter material to void-forming structures may include, but is not limited to, a ratio between 1:1 to 3:1. For example, the getter material may be a nanopowder, and the mass ratio of the nanopowder to spherical PE void-forming structures may include, but is not limited to, one or more of the following: 1:1; 1:25:1; 1.5:1; 1.75:1; 2:1; 2.25:1; 2.50:1; 2.75:1 or 3.0:1. In one embodiment, forming operation 1406 mixes the getter material and the void-forming structures via a wet mixing process. For example, the void-forming structures may be mixed with a solution to form a component mixture solution which in some cases may be a suspension mixture (e.g., including particles large enough to settle). Among other components, the solution may include, for example, water or alcohol (e.g., ethanol). The forming operation 1406 may, in some implementations, entail addition of a binder agent to the mixture including the void-forming structures and getter material to aid in cohesion of the getter material and/or the forming of voids from the void-forming structures. The binder agent may include any binder agent known in the art of powder processing. For example, the binder agent may include, but is not limited to, polyethylene glycol (PEG). For instance, the mixture of step 1406 may include, but is not limited to, 1-10% binder agent by mass (e.g., 5% PEG by mass). Binder agents may be useful in both wet and dry mixing processes. In one wet mixing process, a surfactant is added to a suspension including the getter material, void-forming structures, and a solution. The surfactant serves to aid in the dispersion of the getter material (e.g., if in powder form). In one embodiment, the surfactant is added to the solution prior to addition of the getter material and/or the sacrificial void-forming structures. The amount of surfactant added to the suspension may include, but is not limited to, 0.05 to 2% by mass (e.g., 0.1% by mass). The surfactant may include any surfactant known in the art such as, but not limited to, polyoxethlyene (20) sorbitan monooleate. In another example wet mixing process, the getter process mixture is a suspension (e.g., getter material, sacrificial void-forming structures, and solution) and is treated with an ultrasonic bath. For example, the ultrasonic bath may be applied after addition of a binder agent and/or surfactant (e.g., as described above). The ultrasonic bath may help break up clumps of getter material powder and facilitate uniform mixing of the getter material and sacrificial void-forming structures in the solution. Additional or alternative filtering of particulate matter may be used including agitation, mesh filters, etc. In any of the above-described embodiments including a suspension, the forming operation 1406 may further include one or more operations for drying the suspension. For example, a furnace or oven may be used to dry the suspension. In contrast to the above-described wet-mixing and drying techniques, the forming operation 1406 may also be a dry mixing process. For example, a dry mixture including the getter material and the void-forming structures may be mixed using any mixing device known in the art, such as, but not limited to, a mixer, tumbler or the like. It is noted herein that a binder agent may also be employed in a dry mixing process. In one such implementation, a binding agent (e.g., PEG) is added to dry getter material powder and the void-forming structures in a select proportion (e.g., 1-10% binder agent by mass). A densification operation 1408 densifies the getter process mixture. In one embodiment, the densification operations 1408 includes pressing the getter process mixture at a selected pressure to form a consolidated pellet. Although the applied pressure may vary from one implementation to another, the applied pressure is—in general—sufficient to form a self-supporting consolidated volume. In one implementation, the densification operation 1408 applies a pressure in the range of 200 to 1300 MPa (e.g., 750 MPa) to the getter process mixture. The getter material and sacrificial void-forming structures may be consolidated using any densification device and/or technique known in the art. For example, the getter material and sacrificial void-forming structures may be pressed into a pellet using any pellet die known in the art of pellet processing. The density of the consolidated volume (e.g., the compressed getter process mixture) may be controlled by the die pressure applied to the getter process mixture and/or by the amount of void-forming structure included in the getter process mixture. In some implementations, the densification operation 1408 entails sintering. Sintering may, for example, include heating the getter process mixture to a selected temperature for a selected time. In one implementation, the getter process mixture is heated to a temperature between about 1000 and 1500° C. and held at that temperature between 1 and 24 hours. For example, the getter process mixture may be heated to a temperature of 1350° C. and held at that temperature for 4 hours. By way of another example, the consolidate volume may be heated to a temperature of 1100° C. and held at that temperature for 8 hours. The sintering of ceramic materials is generally discussed in Borg, R. J., & Dienes, G. J., An Introduction to Solid State Diffusion. San Diego: Academic Press Inc. (1988), which is incorporated herein by reference in the entirety. In some implementations, sintering of the getter process mixture can cause a thermal breakdown of the void-forming structures. Specifically, the void-forming structures may break down (e.g., undergo pyrolysis) and exit the getter body, leaving behind a solid getter body. In some implementations, sintering is carried out in an atmosphere to enhance pyrolysis of the void-forming structures. For example, the sintering step may be carried out in the presence of an atmosphere containing oxygen (e.g., air). In some implementations that utilize sintering, the densification operation 1408 further entails applying a pre-heat treatment to the getter process mixture prior to sintering to help initiate and/or fully facilitate thermal breakdown of the void-forming structures. For example, the pre-heat treatment heats the getter process mixture to an intermediate temperature lower than a sintering temperature for a select period of time so as to fully burn out the void-forming structures prior to sintering. For instance, the getter process mixture may be heated to an intermediate temperature between 400 and 800° C. and held at that intermediate temperature for 1 to 10 hours. In one specific implementation, the consolidated volume is heated to an intermediate temperature of 500° C. for 4 hours. In implementations that utilize heat treatment in the densification operation 1408, the temperature of the consolidated volume may be controlled at a selected ramp rate. For example, a ramp rate is selected for use during the void-forming structure burn-off process to ensure that the consolidated volume does not break apart prior to sintering. In one implementation, the temperature of the consolidated volume is ramped at a rate between 0.1 and 5° C./min, such as, but not limited to, 1° C./min. Notably, some implementations of the disclosed technology do not include the densification operation 1408 (e.g., pressurization, heating, sintering.) For example, some void-forming structures may be capable of forming voids naturally, such as through settling. In still other implementations, the densification operation 1408 entails compaction without heating or sintering. Various parameters of the densification operation 1408 may be selectable to control the density of the resulting getter element. For example, the ratio of the amount (by mass) of getter material to void-forming structures may be controlled so as to control the density of the consolidated volume and, thus, the densified getter element. By way of another example, the pressure applied via the densification operation 1408 may be controlled so as to control the density of the getter process mixture and the resulting getter element. Moreover, weights and sizes of the void-forming structures may be selected to form a distribution of voids describable by a particular size or shape gradient. For example, the distribution may form naturally via settling or agitation of void-forming structures of different sizes or shapes. In another embodiment, multiple layers of void-forming structures with different sizes and/or shapes are systematically created in the getter process mixture. FIGS. 15A-15D illustrate example shapes of sacrificial void-forming structures that decompose when subjected to thermal and/or chemical treatment. The sacrificial void-forming structures of FIGS. 15A-15D are merely illustrative and non-limiting examples of structures that may be used to create “voids” in a getter body formed from a getter process mixture. Specifically, FIG. 15A illustrates an example volume of sacrificial void-forming structures 1500 that are spherical in shape (e.g., a sacrificial void-forming structure 1502). In other implementations, the sacrificial void-forming structures are shaped differently, such as ellipsoids, oblate spheroids, prolate spheroids, etc. For example, FIG. 15B illustrates the quantity 1502 of ellipsoid-shaped sacrificial void-forming structures. FIG. 15C illustrates an example volume 1504 of oblate-spheroid-shaped sacrificial void-forming structures, and FIG. 15D illustrates an example volume 1506 of prolate-spheroid-shaped sacrificial void-forming structures. It is noted herein that spheres formed from PE having a size distribution in the range of 50 and 200 μm display adequate thermal decomposition at temperatures between 330 and 410° C. suitable for use as void-forming structures of the present disclosure. FIG. 16 illustrates a conceptual view of a portion of a consolidated volume 1600 of a getter process mixture, such as that formed during the densification operation 1408 described with respect to FIG. 14. The consolidated volume is a pressurized volume including a getter material 1602 and void-forming structures 1604 that provide at least one through-channel 1606 that permits transport of a fluid flow through the volume 1600. FIG. 17 illustrates graph 1700 depicting the percentage of theoretical density (TD) of a getter process mixture achieved as a function of applied die pressure. As shown in the graph 1700, density, as expressed in terms of percent of TD, increases with increasing die pressure. In one implementation, density of a getter element is selected to balance fission product uptake in the getter element with the ability to maintain sufficient flow through the getter element. In one embodiment, the density of the fabricated getter element is between 25 and 45% TD. For example, the density of the fabricated getter element may have a density between 35 and 40% TD. In another implementation, the fabricated getter element has a density between 50-70% TD. In still another implementation, the fabricated getter element has a density between 60 and 70% TD. FIG. 18A illustrates an example getter body 1800 including sacrificial void-forming structures 1806 intermixed with a getter material 1804. FIG. 18B illustrates the example getter body 1800 after undergoing thermal or chemical treatment that decomposes the sacrificial void-forming structures, leaving behind voids (e.g., a void 1808). In some implementations, the getter body 1800 of FIG. 18B is subjected to high pressures and heat to transform the getter body 1800 into a sintered pellet or other structure. FIGS. 19A and 19B illustrate scanning electron microscopy (SEM) images of the void structure of a getter element formed using spherical PE void-forming structures, in accordance within one or more embodiments of the present disclosure. More specifically, FIG. 19A illustrates a radial cross-section of the fabricated getter element and depicts a number of pores that form the overall void structure of the getter element. In one embodiment, the average pore size of the illustrated void structure is between 50 and 200 μm. For example, the void structure may have, but is not limited to, an average pore size of approximately 100-120 μm. It is to be appreciated that the void forming structures may have any appropriate size and/or shape (or even various sizes and/or shapes) as may be suitable. For example, the void forming structures may include structures having a diameter greater than 200 μm. FIG. 19B illustrates a zoomed-in view of a single pore of the void structure and depicts the grain structure of the sintered getter material. It is noted herein that the getter element associated with the SEM images of FIGS. 19A and 19B may have a density within the ranges provided above. FIG. 20 illustrates a series of example additive fabrication operations 2000 for forming a getter element. Unlike the getter body forming processes described above (e.g., operations 1400 described with respect to FIG. 14), the additive fabrication process operations 2000 form a getter body without using any sacrificial void-forming structures. For example, the additive fabrication operations 2000 may entail 3D printing to create voids, such as via a selective laser sintering process. A providing operation 2002 provides a getter material. In one embodiment, the getter material in provided in particulate form. For example, getter material may be a metal oxide powder (e.g., ZrO2, TiO2, MoO2, MoO3, NbO2, Nb2O5, Ta2O5, VO2, V2O5, and Cr2O3). In one embodiment, the average particle size of the getter material is between 100 and 500 nm. In another embodiment, the average particle is at or below 100 nm. It should be understood that a wide range of particle sizes, including those in excess of 500 nm, may be suitable for use in different implementations depending on the getter material and manufacturing processes employed. An additive formation operation 2004 uses an additive manufacturing operation (e.g., 3D printing) to synthesize a free-standing three-dimensional object from the getter material. Collectively, the free-standing structures form a getter body (e.g., as shown and further described with respect to FIGS. 22-24, below). One example suitable additive manufacturing process is selective laser sintering. Selective laser sintering uses a laser to sinter powdered material by aiming and firing the laser at points in space defined by a 3D model, thereby binding material together to create a solid structure. Selective laser sintering is generally described in U.S. Pat. No. 4,863,538, filed on Oct. 17, 1986, which is incorporated above by reference in the entirety. Elements manufactured via the example additive fabrication operations 2000 may include any micro and/or macro-structure(s) capable of maintaining the fission product uptake for the given application, which may maintain sufficient flow through a getter element. A few example getter body structures are provided in FIGS. 21-23. FIG. 21 illustrates an example getter body 2102 formed via an additive fabrication (e.g., 3D printing) process. Various elements (e.g., an element 2104) of the getter body 2102 may assume different shapes and sizes in different implementation. In one implementation the elements of the getter body 2102 are not attached to one another, but rest freely within a container (e.g., a cylindrical container to create the illustrated distribution). In another implementation, the different elements are interconnected. Spaces between the different elements of the getter body 2102 create transmission channels 2106 through which fluid of an input stream 2108 may pass and come into contact with the active surface of the transmission channels within the getter body 2102. FIG. 22 illustrates another example getter body 2202 formed via an additive fabrication (e.g., 3D printing) process. The getter body 2202 is a single, free-standing structure including rows and columns of pores 2206 to maximize surface area of contact between the getter body 2202 and an input stream 2208. FIG. 23 illustrates yet another example getter body 2302 formed via an additive fabrication process. The getter body 2302 includes a number of individual elements (e.g., an element 2304) each added to an interconnected structure via an additive fabrication process. Spaces between the different elements of the getter body 2302 create transmission channels 2306 through which gas or liquid of an input stream 2308 may pass and come into contact with surface area of the getter body 2302. The getter body 2302 includes different elements including different getter materials, as indicated by the shading in FIG. 23. For example, a first element 2304 printed to a first portion of the getter element includes a first getter material, while a second element 2305 printed to a second portion of the getter element includes a second getter material. In this regard, two or more types of materials may be provided in the single getter body 2302, facilitating a targeted uptake of two or more types of fission products or more products from an input stream 2308. For example, the first element 2304 may include a getter material targeted for uptaking cesium, while the second element 2305 includes a getter material targeted for uptaking another element or another compound of cesium. Other getter bodies formed by similar processes may include greater than two getter elements for uptake of greater than two target fission products. FIG. 24 illustrates example operations 2400 for forming a getter element via a sacrificial templating process. A providing operation 2402 provides a suspension including a getter material reactive with a targeted fission product of a nuclear reaction. An impregnation operation 2404 impregnates a porous template structure with the suspension. A solidifying operation 2406 solidifies the suspension, and a removing operation 2408 removes the template structure, leaving behind a solidified getter element with a porous structure mimicking the porous template. For example, the removal operation 2408 may entail a thermal or chemical treatment that causes the porous template structure to decompose. FIG. 25 illustrates example operations 2500 for forming a getter element via a direct foaming operation. A providing operation 2502 provides a suspension including a getter material reactive with a target fission product. An introduction operation 2504 introduces gas into the suspension to form a foam that includes the getter material. A first solidifying operation 2506 solidifies the suspension, and a second solidifying operation 2508 solidifies the foam to form a solidified getter element having a getter body that includes a void structure created by the gas. The herein described components, operations, devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken as limiting. Furthermore, it should be understood that process operations described herein may be performed in any order, adding and omitting as desired, unless explicitly claimed otherwise or a specific order is inherently necessitated by the claim language. The above specification, examples, and data provide a complete description of the structure and use of exemplary embodiments of the disclosed technology. Since many embodiments of the disclosed technology can be made without departing from the spirit and scope of the disclosed technology, the disclosed technology resides in the claims hereinafter appended. Furthermore, structural features of the different embodiments may be combined in yet another embodiment without departing from the recited claims. |
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