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047132140	abstract	Device for purifying liquid metal coolant for a fast neutron nuclear reactor, comprising a pump (6) for circulating the liquid metal, the metal then being purified and heated after purification, as well as a filter (24) of metallic fibers. It consists of an assembly of annular chambers with a vertical axis, closed at their lower part by a base plate (10) and defined and separated from each other by an assembly of coaxial cylindrical metallic shells. The device comprises, from the exterior inwards, a degassing chamber (12), a chamber enclosing an economizer-exchanger (14, 16), a thermal insulation wall (18), a cooling chamber (20) and a purifying chamber (22) in the central part of which is a filter cartridge (24). In the central part of the cartridge (24) is a channel (25) for collecting the purified liquid metal in communication with a basin (44) for collecting the purified liquid metal rising above the device. The invention applies, in particular, to fast neutron nuclear reactors of an integrated type, cooled with liquid sodium.
	claims	1. An X-ray detector for detecting X rays, comprising:a semiconductor for generating electric charges therein upon X-ray incidence;electrodes formed on opposite sides of said semiconductor for application of a predetermined bias voltage;a first carrier selection layer formed on said semiconductor at a side of a positive one of said electrodes, that is at a side of an electrode having a higher potential, for restricting an injection of holes; anda second carrier selection layer formed on said semiconductor at a side of a negative one of said electrodes, that is at a side of an electrode having a lower potential, for restricting an injection of electrons;wherein said semiconductor is amorphous selenium (a-Se) doped with a predetermined quantity of an alkali metal or alkaline earth metal in a quantity ranging from 0.01 to 10 ppm. 2. An X-ray detector as defined in claim 1, wherein one of said electrodes formed at an X-ray incidence side is a positive one of said electrodes to which the bias voltage is applied to increase potential. 3. An X-ray detector as defined in claim 2, wherein the quantity of the alkali metal or alkaline earth metal doped is in a range of 0.05 to 2 ppm. 4. An X-ray detector as defined in claim 2, wherein said alkali metal or alkaline earth metal is one of Li, Na, K, and Ca. 5. An X-ray detector as defined in claim 2, wherein said semi-conductor is formed on a TFT substrate having thin film transistor switches, charge storing capacitors and one of said electrodes. 6. An X-ray detector as defined in claim 1, wherein the quantity of the alkali metal or alkaline earth metal doped is in a range of 0.05 to 2 ppm. 7. An X-ray detector as defined in claim 1. wherein said alkali metal or alkaline earth metal is one of Li, Na, K, and Ca. 8. An X-ray detector as defined in claim 1, wherein said semi-conductor is formed on a TFT substrate having thin film transistor switches, charge storing capacitors and one of said electrodes. 9. An X-ray detector for detecting X rays, comprising:a semiconductor for generating electric charges therein upon X-ray incidence;electrodes formed on opposite sides of said semiconductor for application of a predetermined bias voltage;a first carrier selection layer formed on said semiconductor at a side of a positive one of said electrodes, that is at a side of an electrode having a higher potential, for restricting an injection of holes;wherein said semiconductor is amorphous selenium (a-Se) doped with a predetermined quantity of an alkali metal or alkaline earth metal in a quantity ranging from 0.01 to 10 ppm.
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048636720	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to absorber rods for nuclear reactors with spherical fuel elements which are exhausted after a single passage through the core and more particularly absorber rods inserted directly into the pile in order to affect the prevailing neutron flux in the reactor by absorber material located in an annular gap between two concentrically arranged cylindrical rod elements. 2. Description of the Related Technology Absorber rods are used in nuclear reactors to control the reactor output, the startup and shutdown processes, to equalize burnup, and to shut down the reactor. For this reason, they contain a neutron absorbing substance, i.e., an absorber material. The absorber material reduces the neutron flux and thus reactivity of the reactor depending on the immersion depth of the rod into the reactor filled with fuel elements based on its neutron capture cross section. The neutron flux in the reactor attains its maximum flow density as a function of the burnup state of the fuel elements at different heights of the reactor. If the radiation intensity, i.e., the radioactivity of the fuel elements varies, the maximum neutron flux also changes. In nuclear reactors having piles of spherical fuel elements, in contrast to nuclear reactors with block or rod shaped fuel elements, it is possible to replace fuel elements continuously, without interrupting the operation of the reactor, and thereby to affect the burnup state of the fuel elements in nuclear reactors having piles of spherical fuel element . The reactor may be adjusted so that the fuel elements are used up after a single passage and replaced by new ones. This operational principle is also called the OTTO principle (OTTO=once through then out). In reactors operated by the OTTO principle, the maximum of the neutron flux is located in an intermediate space between the pebble pile and the reactor cover. The absorber rods are inserted through this space into the reactor. For reactor specific reasons the absorber rods project in their rest position into the reactor space above the pebble pile and are therefore constantly exposed to the reactor atmosphere. In the case of nuclear reactors according to the OTTO principle, the absorber rods in this area are exposed to additional neutron irradiation stresses. Additional mechanical stresses appear upon rod insertion into the pile as there are no guide installations in nuclear reactors with piles of spherical fuel elements. These additional stresses result from forces against the fuel elements which oppose the insertion of the absorber rod. This resistance of the pile to insertion increases with the depth of the insertion. Absorber rods are only supported in a guide area in armored tubes in a fashion similar to a cantilever beam. Depending on its free length and section modulus the free end immersing into the peeble pile may be deflected from its immersion axis. Accordingly, the absorber rod is exposed to a lateral force producing a bending moment in addition to the force acting in a direction opposing its penetration. It is therefore necessary to take these types of operational mechanical stresses into account in the design of absorber rods for a nuclear reactor They must be correlated with the already present stressing of the rods. Absorber rods are stressed thermally upon their immersion in the reactor in two respects. The radiation heat emitted by the fuel elements leads to a heating of the rod and heat is generated in the absorber material of the absorber rod as the result of neutron absorption. An unacceptable increase in the temperature of the absorber rod due to these heat sources, i.e., an increase in temperature to a value at which the rod would lose its minimum mechanical strength, must be safely excluded. The same is true for the case in which the absorber rod would lose its necessary elasticity and ductility due to neutron embrittlement. Exposure to neutron radiation is, as set forth above, dependent on the layout and the mode of operation of the reactor, i.e., the position of the maximum neutron flux density in the reactor. Mechanical stresses are functions of geometrical parameters, such as the rod cross section, core diameter or core height and thermal stresses are determined by the fuel element inventory. SUMMARY OF THE INVENTION In view of relationships set forth above, it is an object of the invention to provide absorber rods which may be manufactured simply and cost effectively; the design configuration of which enables long term use free of incidents in a reactor having a spherical fuel element pile. The absorber rod, according to the invention, includes concentrically arranged pairs of cylindrical rod elements. The inner rod element performs the support function, i.e., absorbing and transmitting the forces and moments resulting from the movement of the rod upon the insertion of the rod into the pebble pile. The inner rod is dimensioned to escape damage by deformation or fracture. The outer element serves as a protective shield for the inner rod element against excessive thermal and radioactive stresses. The absorber material serves to shield the inner rod element from radioactive stress. The inner cylindrical rod element is advantageously designed as a support element to absorb mechanical stresses. It is protected by the absorber material against the constantly acting neutron radiation. The outer rod element may in this fashion have a function limited to maintaining the absorber element in its position and shielding the inner supporting tube. All of the outer mechanical forces and moments are introduced by the rod tip attached by welding and absorbed and transmitted by the inner tube. The tip of the rod is generally an integral head piece welded to the adjacent inner and outer rod elements. The inner and outer rod elements are joined to the tip by respective annular weld joints axially offset relative to each other. For thermal and neutron physical reasons the weld is appropriately located as far as possible from the tip of the rod, preferably at a distance corresponding to the diameter of the rod, from the frontal plane of the rod tip. The head piece may exhibit a central bore with a cylindrical piece, which in turn supports a holding device for the head piece. In the event of a fracture of the absorber tip the holding device functions to retain the tip, thereby preventing its irreversible immersion in the pebble pile. In this manner even damaged absorber rods may be retrieved completely from the fuel element pile, without any potentially interfering residues remaining therein. As mentioned above, gas flows through the absorber rods for cooling. It is advantageous to use part of the flow of cooling gas passing through the fuel element pile. This partial flow is separated above the fuel element pile where relatively low cooling gas temperatures are prevailing and guided through axially placed cooling gas slots distributed over the circumference of the outer rod element. By the appropriate choice of the size, number and axial positions of the slots on the absorber rod, adequately low material temperatures may be obtained in both the inner and the outer rod elements. It is possible in this manner to use the absorber rods which are designed according to the invention in high temperature reactors with cooling gas outlet temperatures of up to 750.degree. . It is further possible to use them in nuclear reactors combined with gas turbine machines. In nuclear reactor plants of this type having closed gas loops, such as so-called single loop installations, the gas temperature in the reactor core may attain values of 750.degree. to 950.degree. C. At least two inlet parts are provided in the outer rod element in axially different locations for introduction of the cooling gas. The inlets operate as a function of the position of the rod relative to the pebble pile. In principle, the part of the rod located in the pile is always cooled. The two inlet parts exhibit a large number of axially arranged inlet slots distributed over the circumference of the outer rod elements, thereby insuring uniform flow and cooling throughout the zone or area of the rod below the inlets. The cooling gas outlet is provided in the tip of the rod and are in the form of slots which have the advantage over bore holes in that they reduce the loss of pressure and improve cooling. As the result of the aforedescribed cooling, both a radial and an axial temperature gradient are created in the absorber rod. In order to prevent the additional loading of the absorber rod, by stresses generated by restricted thermal expansion, the rod elements without a support function are mounted on one end only and are slidingly guided in a defined slot. To further enhance cooling, the inner rod element may be in the form of a ribbed tube. In order to prevent the occurrance of notch effects and to minimize thermal stress peaks, care must be taken in the design of the single piece head portion or rod tip to provide a rounded structure terminating the annular gap between the outer and the inner connecting cylinders to which the outer and inner rod elements are joined. This measure prevents permanent damage such as cracking by radial deflections of the outer rod element transmitted to the outer cylindrical connecting piece of the head piece. The absorber material, which may be absorber rings of boron carbide, is annular in shape and manufactured with uniform dimensions (diameter and height) for simplification and the reduction of costs. Spacers or rings are inserted between the inner and outer rod elements in order to keep the absorber material away from the joint location (weld) of the outer and inner rod element to the rod tip. This prevents undesirable carburization and carbide formation in the rod elements during welding and in operation. Furthermore, in this manner temperature and stress peaks are kept away from the weld joints. The spacers also serve to maintain annular gaps between the inner and the outer rod elements and the absorber material in order to prevent harmful reactions between the rod elements and the absorber material upon swelling of the absorber material under radiation. An axial gap between the inner and the outer rod element is not occupied to equalize absorber material axial swell under radiation thereby avoiding additional axial stresses in the absorber rod. The absorber rod geometry is a smooth, cylindrical welded structure with a constant outer diameter. The outer surface of the rod is provided with an abrasion resistant layer, which is conveniently applied by means of flame spraying to prevent abrasion and the resulting variation in diameter, particularly in the area of bearing locations and barking devices. Coating layers produced in this manner are characterized by above average adhesion and a high density. Coating with chromium carbide has been found to be especially advantageous. Both the weakening of the outer rod element and abrasion on reactor components and fuel elements are prevented by maintenance of a smooth rod surface. The absorber rod has a screwed-on coupling on its upper end to establish the connection with the associated rod drive. The coupling exhibits a spring elastic claw coupling and may be released by remote control. The coupling has two parts, a claw body with coaxially arranged circumferential claws and a cylindrical counterpart with a collar engaged by the claws. The two coupling halves are easily separated and axially displaced by a release device, which may be a circular, axially displaceable part pressuring the claws from its clamping position. Bores are provided on the word circumference, to be engaged by a holding tool for securing the rod during installation and dismantling. This and further advantageous embodiments and improvements of the invention are set forth in the claims. A greatly different layout of a shutdown rod for nuclear reactors with a pile of spherical fuel elements is shown in DE-2 066 109. It consists of a rod with two concentric cylindrical rod elements connected to a common tip. The compressive stress on the contact surface of the rod tip and fuel elements is reduced by an increased support surface. The solution consists of a recess or hollow in the rod tip adapted to the contour of the fuel element similar to recess 46. No further characteristics relevant to the present invention are disclosed. An embodiment of the invention is presented below with reference to the drawings, wherein the invention is shown in more detail, together with advantageous configurations and improvements.
046541861	description	DETAILED DESCRIPTION The core of the nuclear reactor is crossed in a known manner by a primary fluid which circulates in a primary loop; this fluid absorbs energy as it passes through the core and gives up its energy during its passage through the steam generator to a secondary fluid circulating in a secondary loop. The primary circuit comprises a cold branch and a hot branch, the cold branch being of course placed between the steam generator and the core and the hot branch between the core and the steam generator, in the direction of flow of the fluid. The temperature of the primary fluid is measured at two points in a conventional manner, one of the points being situated on the cold branch and the other point being situated on the hot branch. In FIG. 1, the cold branch temperature signal is designated by 1 and the hot branch temperature signal by 1'. An operator 2 computes, in a conventional manner, the value of the enthalpy at the point of temperature measurement in the cold branch and the point of temperature measurement in the hot branch. The enthalpy can, for example, be determined by a second degree polynomial in T where T, is the measured temperature. To obtain improved precision, the computation can also be carried out using a third degree polynomial. Two signals 3 and 3' are thus obtained for enthalpy in the cold branch and for enthalpy in the hot branch. Two time shift operators 4 and 4' permit the signals 3 and 3' respectively to be delayed. The transfer functions employed are respectively .epsilon..sup.-.tau.op and .epsilon..sup.-.tau.'op where .tau..sub.o and .tau.'.sub.o represent, respectively, the total average time of transit of a molecule of primary fluid between the points of temperature measurement in the cold branch and in the hot branch, and the total average time of transit of a molecule of primary fluid between the points of temperature measurement in the hot branch and in the cold branch (p being the LAPLACE variable). For increased precision, it may be taken into account that the above-mentioned time of transit can be different for two different molecules of water; in particular, the reactor configuration is such that, generally speaking, water molecules have very different speeds at the outlet of the core. It would therefore be possible, using an integrator, to take these different times of transit into account instead of considering only the average time, as shown in FIG. 1. Signals 5 and 5' are obtained at the output of the time shift operators 4 and 4' and are then entered into the registers 6 and 6'. The register 6 produces the difference between the hot branch enthalpy signal 3' and the cold branch enthalpy delayed signal 5. the output signal of the register 6 is designated by 7. The register 6' produces the difference between the hot branch enthalpy delayed signal 5' and the cold branch enthalpy signal 3. The output signal of the register 6' is designated by 7'. Signals 7 and 7' are entered respectively into multipliers 8 and 8' where they are multiplied by the primary flow rate signal 9, the latter being measured in a completely conventional manner. To increase the precision, the primary flow rate signal 9 is filtered at 10 to take account of the variation in the average time of transit of a molecule of fluid through the core. The transfer function of the filter is in this case ##EQU1## where .tau..sub.1 represents the average time of transit of a molecule of primary fluid through the core. At the output of the multiplier 8 a primary thermal power signal 11 is obtained, and at the output of the multiplier 8' a signal 11' is obtained which represents the thermal power absorbed by the steam generator. The dynamics of the primary fluid temperature measurements in the cold branch and in the hot branch are compensated by two identical phase lead correctors 12 and 12' into which the signals 11 and 11' are entered, respectively. The output signals of these correctors are shown as 13 and 13'. The transfer function of these correctors is ##EQU2## where .tau..sub.2 is the time constant of the temperature measurements (measurement corrector) and where .tau..sub.3 is a reduction filter of the transient gain of the measurement corrector. Signals 13 and 13' are then entered into the comparators 14 and 14' where they are compared respectively to a neutron power signal and to a signal representing the thermal power produced by the steam generator. The neutron power signal is obtained in a conventional manner by means of neutron flux measurement chambers which are situated outside the core. The neutron power signal is shown as 15. Signal 15 is corrected as a function of the temperature variations by the use of a correction coefficient K1 between the measurement of neutron flux and the temperature of the annular space in which the neutron power chambers are conventionally situated, it being possible for this temperature to be taken as similar to the cold branch temperature. In FIG. 1, of course, .theta. represents a nominal temperature. In order to respect the signal phase as well as possible in transient operation, the temperature correction is shifted in time by the term .epsilon..sup.-.tau.15p (.tau..sub.15 is the time of transit between the cold branch measurement point and the core entry). To this term may be added a low-pass filter ##EQU3## to allow for the time required by the flux measurement chambers to respond to a variation in the temperature of the cold branch. 16 refers to the neutron power signal which has been corrected for temperature. Signal 16 is made dynamically equivalent to signal 13 by means of a point model 17 of heat transfer between the nuclear flux and the thermal flux of the primary fluid (.tau..sub.4 represents the time constant of heat transfer); the output signal of the model 17 is then entered into a point model of heat transfer of primary fluid in the core corresponding to the time of transit .tau..sub.1 /2 of a molecule of primary fluid from the center of the core to the outlet of the core. The output signal of the model 18 is then delayed by a time shift operator 19 expressing the time of transit .tau..sub.5 of a molecule of primary fluid from the outlet of the core to the point of temperature measurement in the hot branch. The neutron power signal which has been made dynamically equivalent to the signal of primary thermal power is shown as 20. Signal 20 is compared to signal 13 in the comparator 14. The signal 21 which is produced by this comparator is used for correcting the neutron power signal 16 by means of a corrector 22. This corrector 22 comprises an integrator with an integration constant .tau..sub.6 and gain k.sub.2..tau..sub.7 and .tau..sub.8 are respectively, phase lead and phase delay time constants, .tau..sub.8 being smaller than .tau..sub.7. The signal 23 produced by the corrector 22 is added to the signal 16 in the register 24. At the output of the register 24 a normalized signal of neutron power is obtained, and at the output of the model 17 a normalized signal of thermal power is obtained. The device according to the invention thus permits a fast and precise signal of the primary power of the reactor to be obtained by means of only two temperature-measuring sensors and conventional chambers for measuring neutron power. The device according to the invention further permits a fast and precise signal of the secondary power of the reactor to be obtained by means of the same two temperature sensors, as will be described hereafter. To produce a fast and precise signal of the secondary power of the reactor, signal 13' is compared, at 14', to a signal representing the thermal power produced by the steam generator of the cooling loop under consideration; this signal, which is derived from a simplified secondary balance, lacks precision but has the great advantage of being a fast-response signal. This signal 16' is produced in a computer 25 from four signals, namely the temperature and the flow rate of the steam generator feed water, together with the pressure and the flow rate of steam produced by the steam generator. The signal 16' is made dynamically equivalent to the signal 13' by means of a point model 17' of heat transfer between the secondary fluid and the primary fluid (.tau..sub.q is the time constant of heat transfer). The model 17' is followed by a point model 18' of heat transfer of the primary fluid in the steam generator corresponding to the time of transit .tau.Hd 10/2 of a molecule of primary fluid from the center of the steam generator to the outlet of the steam generator. A time shift operator 19' enables the time of transit .tau..sub.11 of a molecule of primary fluid from the outlet of the steam generator to the point of temperature measurement in the cold branch to be taken into account. At the output of the operator 19' a signal 20' is obtained which represents the thermal power produced by the steam generator, this signal being made dynamically equivalent to the signal 13' representing the thermal power absorbed by the steam generator. These two signals 20' and 13' are compared in the comparator 14'. The signal 21' produced by the comparator 14' is entered in a corrector 22'. The signal 23' produced by this corrector is used to correct the signal 16' to which it is added in a register 24'. In this case the corrector 22' is an integrator whose integration constant is .tau..sub.12 and gain k.sub.3. .tau..sub.13 and .tau..sub.14 are time constants of the phase advance and phase lag of the corrector (.tau..sub.14 is smaller than .tau..sub.13). A fast and precise signal of normalized secondary power is obtained at the output of the register 24'. The device according to the invention thus makes it possible to obtain at any time a fast and precise signal of primary power (neutron power and thermal power transmitted at the center of gravity of the core) and of secondary power, using two temperature sensors in the cold branch and the hot branch, neutron power measurement chambers and a simplified secondary balance. This device is of particular advantage during periods of transient operation. It permits instant detection of any variation in the primary or secondary power, permitting possible failures which have caused these variations to be remedied very quickly. The device according to the invention contributes to a proper protection of the core, particularly in the high performance reactors which are constructed at present. The correctors 22 and 22' may be designed differently so as to optimize the response of the correction signal. Furthermore, the registers 24 and 34 could be replaced by multipliers to preserve the measurement zero. The example described relates to a single loop, but the invention can of course apply to a reactor with several loops.
	claims	1. A power source for providing electrical energy, comprising: a particle-emitting source radiating particles having kinetic energy; an ionizable liquid receiving said particles from said particle-emitting source, said ionizable liquid ionizing into ions and electrons when said particles collide with said liquid; and said ions and electrons are separately collectable; whereby a current is generated by separate collection of said ions and electrons. 2. A power source for providing electrical energy as set forth in claim 1 , further comprising: claim 1 a cathode, said cathode receiving said ions. 3. A power source for providing electrical energy as set forth in claim 1 , further comprising: claim 1 an anode, said anode receiving said electrons. 4. A power source for providing electrical energy as set forth in claim 2 , wherein said cathode is zirconium. claim 2 5. A power source for providing, electrical energy as set forth in claim 3 , wherein said anode is iridium. claim 3 6. A power source for providing electrical energy as set forth in claim 1 , further comprising: claim 1 a zirconium cathode, said cathode receiving said ions; and an iridium anode, said anode receiving said electrons. 7. A power source for providing electrical energy as set forth in claim 6 , further comprising: claim 6 said ionizable liquid in communication with said cathode and said anode; said ions migrating to said cathode and said electrons migrating to said anode. 8. A power source for providing electrical energy as set forth in claim 7 , wherein said particle-emitting source comprises: claim 7 a radioactive element. 9. A power source for providing electrical energy as set forth in claim 8 , wherein said radioactive element comprises: claim 8 an alpha-particle (xcex1-particle) emitting element. 10. A power source for providing electrical energy as set forth in claim 9 , wherein said alpha-particle (xcex1-particle) emitting element comprises: claim 9 curium-244. 11. A power source for providing electrical energy as set forth in claim 7 , wherein said ionizable liquid comprises: claim 7 an element of the periodic table. 12. A power source for providing electrical energy as set forth in claim 7 , wherein said liquid element comprises: claim 7 a semimetal. 13. A power source for providing electrical energy as set forth in claim 12 , wherein said liquid element comprises: claim 12 gallium. 14. A power source for providing electrical energy, comprising: a cathode; an anode; a particle-emitting source radiating particles having kinetic energy; an ionizable liquid in communication with said cathode and said anode and receiving said particles from said particle-emitting source, said ionizable liquid ionizing into ions and electrons when said particles collide with said liquid; said ions migrating to said cathode and said electrons migrating to said anode; whereby a current is generated across said cathode and anode. 15. A power source for providing electrical energy as set forth in claim 14 , further comprising: claim 14 said ions and electrons migrating under the influence of an electric field. 16. A power source for providing electrical energy as set forth in claim 14 , wherein said cathode is zirconium. claim 14 17. A power source for providing electrical energy as set forth in claim 14 , wherein said anode is iridium. claim 14 18. A power source for providing electrical energy as set forth in claim 14 , wherein said particle-emitting source comprises: claim 14 a radioactive element. 19. A power source for providing electrical energy as set forth in claim 18 , wherein said radioactive element comprises: claim 18 curium-244. 20. A power source for providing electrical energy as set forth in claim 14 , wherein said ionizable liquid comprises: claim 14 a liquid element of the periodic table. 21. A power source for providing electrical energy as set forth in claim 20 , wherein said liquid element comprises: claim 20 gallium. 22. A power source for providing electrical energy, comprising: an zirconium cathode; a iridium anode; a particle-emitting source including curium-244 radiating alpha particles having kinetic energy; and an ionizable liquid including gallium in communication with said cathode and said anode and receiving said particles from said particle-emitting source, said ionizable liquid ionizing into ions and electrons when said particles collide with said liquid; whereby said ions migrating to said cathode and said electrons migrating to said anode such that a current is generated across said cathode and anode.
	description	This application claims the benefit of priority from Korean Patent Application No. 10-2006-0010714 filed on Feb. 3, 2006 in the Korean Intellectual Property Office, the entire contents of which is incorporated herein by reference. 1. Field of the Invention Example embodiments relate to an E-beam lithography system and, more particularly, to an E-beam lithography system that may synchronously irradiate surfaces of a plurality of photomasks. 2. Description of the Related Art As semiconductor technology continues to develop, advancements in semiconductor devices, especially memory devices, are improving. That is, high speed, low power consumption, high capacity, miniaturized memory devices are being developed. As semiconductor memory devices are improved, technologies for improving integration are becoming even more important. Improving the integration of semiconductor devices is being achieved through the development of new and/or improved circuit design techniques, materials, and various process techniques. Patterning techniques are important in improving the integration of semiconductor devices. Through the patterning technique, the pattern of unit devices, such as transistors, for example, may be finely formed on a wafer. Patterning techniques generally include a photolithography technique and an etching technique. The photolithography technique may include fabricating a photomask and transferring a pattern to a wafer using the photomask. A photomask fabricating technique refers to a technique for forming a pattern to be transferred to a wafer on a photomask. If a pattern is formed on the photomask, the pattern should be formed to have the correct shape and a uniform size. Further, the substrate and the pattern should not have defects. The substrate may be a glass substrate, for example. Another important consideration is that an ideal layer-to-layer overlay alignment tolerance of the photomasks should be zero. If a semiconductor chip is produced on a wafer, various patterns and unit devices may be formed using several photomasks. As a result, a finished semiconductor chip may operate correctly only if the patterns formed on the wafer using the several photomasks are precisely aligned. If even one layer is placed incorrectly and/or is misaligned, a finished unit device may not function properly, which may cause the finished semiconductor including the unit device to be inoperable and/or malfunction. Therefore, it is important for photomasks used in fabricating a semiconductor chip to have precise layer-to-layer overlay alignment. Further, in the case of a fine pattern, duplicate exposure may be performed on a wafer using a plurality of photomasks to form one layer of a pattern, thereby further increasing the importance of layer-to-layer overlay alignment of photomasks. In order to align the patterns formed on the photomasks, factors including registration and orthogonality may be used. Registration refers to an alignment factor indicating how precisely the pitch, size, and interval of the patterns in the X and/or Y direction match ideal and/or computer data. Orthogonality refers to the angles of the X and/or Y directions of the patterns. Various patterns may have several shapes created by combining lines in the X-direction and lines in the Y-direction and/or patterns in the X-direction and patterns in the Y-direction. As such, orthogonality refers to a measure representing whether the lines in the X-direction are precisely perpendicular to the lines in the Y-direction and/or whether the patterns in the X-direction are precisely perpendicular to the patterns in the Y-direction. Herein, registration and orthogonality are referred to generically as pattern alignment. If the photomasks are overlaid, the factor representing whether the pattern alignments match each other is referred to as layer-to-layer overlay alignment and is abbreviated herein as overlay alignment. FIGS. 1A to 1D are views schematically illustrating pattern alignment and/or computer data of two photomasks, which may be fabricated using an E-beam lithography system. FIGS. 1A to 1D are used to further describe and illustrate examples of what is referred to herein as “pattern alignment” and a “difference in pattern alignment”. The drawings are somewhat exaggerated for ease of explanation. A photomask (a) and a photomask (b) may be photomasks that are separately fabricated and have the same pattern. Alternatively, the photomasks may be photomasks which are fabricated to be overlaid. Each of the photomasks may be individually fabricated. FIG. 1A is an example view showing ideal pattern alignment, which may be computer data about patterns to be formed on photomasks. The pattern of the photomask (a) and the pattern of the photomask (b) have an ideal pattern alignment in the X and Y directions meaning there is no error. Hence, if the two patterns shown in FIG. 1A are overlaid, the patterns of the photomask (a) and the pattern of the photomask (b) correspond. However, because the photomasks shown in FIG. 1A represent ideal pattern alignment, photomasks having the pattern alignment shown in FIG. 1A are seldom realizable. FIG. 1B is a view comparing an example photomask (a) having ideal pattern alignment with an example photomask (b) having a difference in pattern alignment. In FIG. 1B, the photomask (a) has ideal pattern alignment corresponding to computer data, and the photomask (b) has pattern alignment in which the registration and orthogonality are slightly different from the computer data Thus, the photomasks shown in FIG. 1B may not precisely overlay each other. If a pattern is formed using the photomasks shown in FIG. 1B, the pattern formed by the photomask (a) would not be precisely overlaid on the pattern formed by the photomask (b) and thus, a unit device would likely not be correctly produced. Thus, a finished semiconductor device including the unit device may be inoperable. Even if the finished semiconductor device including the unit device operates, the finished semiconductor device would likely have low reliability and/or durability. FIG. 1C is a view comparing pattern alignment of example photomasks having the same degree of difference. In FIG. 1C, both the photomask (a) and the photomask (b) have the same pattern alignment. If the photomasks are overlaid, the patterns formed from the photomasks may precisely correspond to each other. Thus, a semiconductor device fabricated using the photomasks of FIG. 1C may operate correctly. However, it is nearly impossible to fabricate the photomasks such that the photomasks have the same degree of difference using conventional techniques. FIG. 1D is a view illustrating two example photomasks having different degrees of difference in pattern alignment. FIG. 1D illustrates a typical example of two photomasks formed using conventional techniques. The photomask (a) and the photomask (b) illustrated in FIG. 1D have different pattern alignment. If a semiconductor device is fabricated using the photomasks, the pattern of an upper layer is not precisely overlaid on the pattern of a lower layer. As a result finished unit devices formed using the photomask (a) and the photomask (b) shown in FIG. 1D and a semiconductor device including the unit devices may malfunction and/or be inoperable. For example, assuming that the photomask (a) is a line pattern and the photomask (b) is a via hole pattern, lines and via holes would likely not overlay each other if the example photomasks illustrated in FIGS. 1B and 1D were used and thus, an electric connection would likely not be created. Thus, the formed unit devices may be inoperable. If an electric connection is somewhat achieved using the example photomasks illustrated in FIGS. 1B and 1D, the lines and the via holes are likely only partially overlaid and thus, an electric resistance of the unit devices formed using the example photomasks illustrated in FIGS. 1B and 1D would likely be undesirably large. In this case, the unit devices may not operate smoothly. Accordingly, the reliability and the durability of a finished semiconductor device including these unit devices would decrease. Further, if the example photomasks illustrated in FIG. 1D are overlaid on the example photomasks having ideal pattern alignment illustrated in FIG. 1A, a normal device may be formed. However, in this case, if the example photomasks have different pattern alignment, overlay alignment tolerance may increase. For example, the overlay alignment tolerance may be twice as much as that of the pattern alignment. Thus, reducing and/or minimizing the overlay alignment tolerance of photomasks is generally more important than reducing and/or minimizing the pattern alignment tolerance photomasks. Pattern alignment tolerance and overlay alignment tolerance may result from the motion error of a stage of conventional photomask fabricating equipment, which may move photomasks forwards, backwards, left, and right. A conventional E-beam lithography system for fabricating photomasks may be operated such that an irradiating system radiating an electron beam irradiates a desired and/or predetermined position; and a stage on which photomasks is mounted moves forwards, backwards, left, and right, thus radiating the electron beam on the surfaces of the photomasks to form patterns. Conventionally, the precision of the mechanical movement of the stage determines the pattern alignment and the overlay alignment. Using conventional devices and/or techniques, it is difficult to reduce and/or solve pattern misalignment issues, which refer to the overlay alignment tolerance of photomasks or overlaid photomasks. The pattern alignment tolerance and the overlay alignment tolerance generally do not depend on the size of a pattern to be formed, but instead have a predetermined mechanical limit. Therefore, assuming that the pattern alignment tolerance for the design rule of a pattern is the tolerance rate of the photomask, the smaller the pattern to be formed, the larger the relative tolerance of the pattern alignment and/or the overlay alignment. As described above, the relative tolerance of the pattern alignment and/or the overlay alignment affects the manufacturing process and circuit design, so that a circuit must be designed in consideration of the pattern alignment tolerance and/or the overlay alignment tolerance. When considering the pattern alignment tolerance and/or the overlay alignment tolerance, the size of each pattern is set to be larger than an ideal size, so that the margin for the process is relatively large. This is a major factor impeding the tendency toward the miniaturization of semiconductor devices. Example embodiments provide an E-beam lithography system. The E-beam lithography system may include a loading unit for loading/unloading substrates, an alignment chamber in which substrates are aligned, and a lithography chamber including a stage where more than one of the substrates are mounted and irradiated with at least one electron beam. The E-beam lithography system may also include a vacuum chamber creating a vacuum in the chambers and a transfer chamber for transferring the substrates from the loading unit or the chambers. According to an example embodiment, the stage may include an E-beam measuring unit for adjusting the electron beam and aligning the stage. The E-beam measuring unit may include a focus measuring portion, a level measuring portion, and a position measuring portion. The focus measuring portion may have multiple levels and/or a shape of multilayered stairs. The level measuring portion may have a shape of a polygon which is convex at an upper part thereof or of a hemisphere. The position measuring portion may have a shape of a polygon having a protruding portion or a cross shape. According to an example embodiment, the E-beam measuring unit may be located at a center of the stage. According to an example embodiment, the stage may include a first direction stage moving portion moving the stage in a first direction and a second direction stage moving portion moving the stage in a second direction. The stage may also include a first direction stage guide and a second direction stage guide. Still further, the stage may include a first direction laser mirror and a second direction laser mirror. According to an example embodiment, a plurality of irradiating systems may be provided in the lithography chamber. Each of the irradiating systems may include an electron gun, an electron lens, and at least one aperture. The at least one aperture may include at least one of a variable rectangular aperture or a circular aperture. The irradiating systems may share one E-beam control system controlling density, current, voltage and energy of the electron beam. According to an example embodiment, the alignment chamber may include a cassette aligning a plurality of substrates. Each of the substrates may be a photomask or a reticle. An example embodiment provides an E-beam lithography method. The E-beam lithography method may include loading substrates onto a loading unit of an E-beam lithography system, transferring the loaded substrates into a vacuum chamber, transferring the substrates fed into the vacuum chamber into an alignment chamber, aligning the substrates in the alignment chamber, transferring the aligned substrates into a lithography chamber and mounting the substrates on a stage provided in the lithography chamber, controlling a plurality of irradiating systems radiating an electron beam onto the substrates mounted on the stage, using a common E-beam control system, driving the stage and simultaneously radiating the electron beam onto the substrates, and transferring the substrates into the loading unit, after the electron beam has been radiated onto the substrates. According to an example embodiment, controlling the electron beam controls density, speed, size, and energy of the electron beam in the E-beam control system using an E-beam control unit provided on the stage. According to an example embodiment, radiating the electron beam while driving the stage includes and repeats moving the stage in a first X direction while fixing the stage in a Y direction and radiating the electron beam, moving the stage in the Y direction while fixing the stage in the first X direction, moving the stage in a second X direction while fixing the stage in the Y direction and radiating the electron beam, and moving the stage in the Y direction while fixing the stage in the second X direction. The first X direction and the second X direction may be opposite directions along an X-axis. Moving the stage in the first X direction and moving the stage in the second X may continuously perform a plurality of unit lithography methods. An example embodiment of the present invention provides an E-beam lithography method. The E-beam lithography method may include transferring substrates into an alignment chamber, aligning the substrates in the alignment chamber; transferring the aligned substrates into a lithography chamber, mounting more than one of the substrates on a stage provided in the lithography chamber; and radiating at least one electron beam onto the substrates mounted on the stage. Example embodiments are described more fully hereinafter with reference to the accompanying drawings. The example embodiments may, however, be embodied in many different forms and should not be construed as limited to the example embodiments set forth herein. Rather, these example embodiments are provided so that this disclosure will be thorough and complete. Like reference numerals refer to like elements throughout. The size or relative size of layers or areas is somewhat exaggerated in the drawings for clarity of the description. It will be understood that when an element is referred to as being “on” another element, it can be directly on the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly on” another element, there are no intervening elements present. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. The terminology used herein is for the purpose of describing particular example embodiments only and is not intended to be limiting. As used herein, the singular forms “a,” “an” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises” and/or “comprising,” or “includes” and/or “including” when used in this specification, specify the presence of stated features, regions, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, regions, integers, steps, operations, elements, components, and/or groups thereof. Furthermore, relative terms, such as “lower” or “bottom” and “upper” or “top,” may be used herein to describe one element's relationship to other elements as illustrated in the figures. It will be understood that relative terms are intended to encompass different orientations of the device in addition to the orientation depicted in the figures. For example, if the device in one of the figures is turned over, elements described as being on the “lower” side of other elements would then be oriented on “upper” sides of the other elements. The exemplary term “lower,” can therefore, encompasses both an orientation of “lower” and “upper,” depending of the particular orientation of the figure. Similarly, if the device in one of the figures is turned over, elements described as “below” or “beneath” other elements would then be oriented “above” the other elements. The exemplary terms “below” or “beneath” can, therefore, encompass both an orientation of above and below. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which the example embodiments belong. It will be further understood that terms, such as those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and the present disclosure, and will not be interpreted in an idealized or overly formal sense unless expressly so defined herein. The example embodiments are described herein with reference to idealized schematic views shown in the drawings and thus, the illustrated views may change in response to manufacturing technology and/or tolerance. That is, example embodiments are not limited to the specific shape shown in the drawings, but instead include a shape varying according to the manufacturing process. Therefore, the areas illustrated in the drawings have schematic characteristics, and the shape of the areas shown in the drawings is for illustrative purposes only, but does not limit the scope of example embodiments. Herein, a photomask refers to an article transferring a pattern to a wafer using light. The photomask may include a photomask, a reticle, a reflective photomask, and/or a photo mirror, for example. Hereinafter, example embodiments of an E-beam lithography system are described in detail with reference to the accompanying drawings. FIG. 2 is a plan view schematically showing an example embodiment of an E-beam lithography system 100. As shown in FIG. 2, the E-beam lithography system 100 may include a loading unit 110, a transfer chamber 140, an alignment chamber 130, an E-beam lithography chamber 150, a stage 160, a plurality of irradiating systems 170, and a vacuum chamber 120. The loading unit 110 may include a first robot arm Ra to load/unload photomasks M and to transfer the photomasks M, for example. The transfer chamber 140 may include a second robot arm Rb to transfer the photomasks M from the loading unit 110 and/or chambers. The alignment chamber 130 may be provided with a cassette 175 on which the photomasks M may be mounted. The alignment chamber may be used to align the photomasks M. The E-beam lithography chamber 150 may be used to radiate an electron beam onto the one or more surfaces of the photomasks M. The stage 160 may be used substantially, simultaneously mount a plurality of photomasks M in the E-beam lithography chamber 150. A plurality of irradiating systems 170 may radiate electron beams onto the plurality of photomasks M. The vacuum chamber 120 may create a vacuum in the chambers 120, 130, 140, and 150. Doors 115, 125, 135, and 145 may be provided between the chambers 120, 130, 140, and 150. The stage 160 may include an E-beam measuring unit 180 to adjust an electron beam and to align the stage 160. The E-beam measuring unit 180 may be located at the center of the stage 160, for example, and may include a focus measuring part 180a, a level measuring part 180b and a position measuring part 180c. The focus measuring part 180a may have a shape including multilayered stages. The level measuring part 180b may have a shape of a polygon, which is convex at its upper portion, or a hemi-spherical shape, for example. The position measuring part 180c may have the shape of a polygon having a protruding portion or a cross shape, for example. The E-beam measuring unit 180 is described later in more detail with reference to FIGS. 6A-8B. The stage 160 may also include a Y-direction stage 161, which moves in the Y direction; an X-direction stage 162, which is provided on the Y-direction stage 161 and moves in the X direction; and bases 163, which are provided on the X-direction stage 162. The photomasks M or the cassette 175 may be mounted on the bases 163, for example. The stage 160 is described later in more detail with reference to FIG. 3. The irradiating systems 170 may include electron guns for discharging electrons, a plurality of electron lenses for guiding the electrons discharged from the electron guns in one direction, and apertures for adjusting electron beams. The interior of each irradiating system 170 and each aperture is described later in more detail with reference to FIGS. 4A-5B. The irradiating systems 170 may share one E-beam control system (not shown). The E-beam control system may control the conditions under which the irradiating systems 170 radiate electron beams. Particularly, voltage, current, amount of electrons, drift, energy, etc., may be controlled by the E-beam control system. Because the irradiating systems 170 may share one E-beam control system, the irradiating systems 170 may be controlled to share the same conditions, which may be relatively low in tolerance. The first and second robot arms Ra and Rb may transfer a plurality of photomasks and/or the cassette 175 on which the photomasks are mounted. The cassette 175 may mount and align a plurality of photomasks M substantially simultaneously. FIG. 3 is a perspective view schematically showing an example stage 160 of an example embodiment of an E-beam lithography system 100. Referring to FIG. 3, the stage 160 of the E-beam lithography system 100 may include the Y-direction stage 161, which moves in the Y direction; the X-direction stage 162, which is provided on the Y-direction stage 161 and moves in the X direction; and the bases 163, which are provided on the X-direction stage 162. The photomasks M and/or the cassette 175 may be mounted on the bases 163. The stage 160 may be mechanically controlled by a stage controller (not shown) and/or a stage drive system (not shown), based on an electric signal provided by a processor of the E-beam lithography system 100. The X-direction stage 162 may include X-direction stage guides 166 and an X-direction laser mirror 164. Likewise, the Y-direction stage 161 may include Y-direction stage guides 167 and a Y-direction laser mirror 165. The Y-direction stage 161 may move along the Y-direction stage guides 167, and the X-direction stage 162 may move along the X-direction stage guides 166. The bases 163 may include the X-direction laser mirror 164, which detects the movement of the stage in the X direction, and the Y-direction laser mirror 165, which detects the movement of the stage in the Y direction. Each base 163 may include a ground portion that contacts a conductor transmitting an electric signal to the surfaces of the photomasks M mounted on the base and discharges static electricity. The X direction and the Y direction are not absolute directions. That is, the X direction and the Y direction may be exchanged with each other. Further, the position of the stage may be changed in various example embodiments. FIGS. 4A and 4B are perspective views schematically showing the lithography chamber 150 and the irradiating systems 170 of an example embodiment of an E-beam lithography system 100. Referring to FIG. 4A, the irradiating systems 170 may radiate electron beams onto the one or more surfaces of the photomasks M mounted on the stage 160, which may be provided in the lithography chamber 150. As shown in FIG. 4A, the E-beam measuring unit 180 may be provided at the center of the stage 160. While a plurality of photomasks M (e.g., the two photomasks illustrated in FIGS. 2 and 4A) is mounted on the stage 160, the stage may move forward, backward, left, and right. The irradiating systems 170 may be individually moved. However, while the irradiating systems 170 radiate electron beams, the position of each of the irradiating systems 170 may be fixed. FIG. 4B is a cutaway view schematically showing an example interior of each of the irradiating systems 170. Referring to FIG. 4B, each of the irradiating systems 170 of an example embodiment of an E-beam lithography system 100 may include an electron gun 170a, electron lenses 170b, and an aperture 170c. The electron gun 170a may discharge electrons. The electron lenses 170b may guide the discharged electrons, and the aperture 170c may adjust the size of an electron beam. The electron gun 170a may be subjected to electric energy and may discharge electrons through heat release or field emission. The electron lenses 170b may form a magnetic field to guide the electrons discharged from the electron gun 170a in a direction. The aperture 170c may adjust the size of the electron beam radiating from each irradiating system 170. FIGS. 5A and 5B are views illustrating operation of an aperture 170C of an example embodiment of an E-beam lithography system 100. The aperture 170C may adjust an electron beam. FIG. 5A schematically shows a variable rectangular aperture. Four blinds Ba adjust the size of an opening S1, S2 through which an electron beam passes, thereby adjusting the size of the electron beam. A variable rectangular aperture may be used during an irradiating method using a vector scanning method, for example. In the vector scanning method, the size of an electron beam may be adjusted during an irradiating operation and the stage may not move only in a fixed direction, and instead may move in various directions while radiating the electron beam. FIG. 5B schematically shows a circular aperture. By rotating a plurality of blinds Bb of the circular aperture, the size of an opening S3, S4 through which an electron beam passes may be adjusted, thereby adjusting the size of the electron beam. The circular aperture may be used during a raster scanning method, for example. In the raster scanning method, an electron beam may not be adjusted in size during an irradiating operation and the stage may be moved only in a desired and/or predetermined direction while radiating the electron beam. While only eight blinds are shown in FIG. 5B, the size of the opening S3, S4 may be adjusted using more blinds Bb according to another example embodiment. FIGS. 6A to 8B are views illustrating an example operation of adjusting an electron beam using an E-beam measuring unit 180 of an example embodiment of an E-beam lithography system 100. The E-beam measuring unit 180 may perform a measuring operation using a laser or an electron beam. An example of an E-beam measuring unit using a laser is described below. However, an electron beam may be substituted for the laser in the following description. FIGS. 6A and 6B are views illustrating an example operation of measuring a focus distance between the irradiating systems 170 and the stage 160 using the focus measuring part 180a. For example, a laser reflected after being radiated onto the multilayered stages of the focus measuring part 180a may be received and profiles may be obtained based on the received laser as shown in FIG. 6B. FIG. 6B shows example profiles obtained by reversing the intensity of the received laser. Because the magnitude of the beam spot of the reflected laser is changed depending on the focus distance, the height of the irradiating systems 170 and/or the stage 160 may be appropriately adjusted to obtain the magnitude of an appropriate beam spot, thereby adjusting the focus of an electron beam. FIGS. 7A and 7B are views illustrating an example operation of adjusting an irradiating angle using a level measuring portion 180b of an example embodiment of an E-beam measuring unit 180. Particularly, FIG. 7B shows example profiles of a laser that is radiated onto the surface of the level measuring part 180b having a hemi-spherical shape and thereafter reflected. Because the reflection and the scattering of the reflected laser are changed according to an angle at which the laser enters the hemi-spherical level measuring part 180b, the example profiles illustrated in FIG. 7B may be obtained. Referring to FIG. 7B, the middle profile shown in FIG. 7B shows the reflection and scattering profile when a laser enters in a vertical direction. The left and right profiles of FIG. 7B show the reflection and scattering profiles when lasers enter at desired and/or predetermined angles. By adjusting the vertical angle of each irradiating system 170 and/or adjusting the level of the stage 160, the profiles may form relatively uniform, symmetrical arrangements. FIG. 7B illustrates that the incident angles of the laser may tend to the left and right. However, the incident angles of the laser may also tend to the front and back. Thus, the vertical angle of each irradiating system 170 or the level of the stage 160 may be adjusted using front and rear laser profiles as well as left and right laser profiles according to an example embodiment. FIGS. 8A and 8B are views illustrating an example operation of adjusting the position of the stage 160 using an example position measuring portion 180c of an example embodiment of an E-beam measuring unit 180. If the laser, reflected after being radiated onto the position measuring part 180c having the shape of the polygon having the protruding portion or the cross shape, is received, the example profile shown in FIG. 8B may be obtained. Because the radiated laser is not reflected but instead is scattered at edges of the protruding portion of the position measuring part 180c having the polygonal or cross shape, the intensity of the received laser is low as shown in FIG. 8B in which the vertical axis representing intensity is inverted. In FIG. 8B, peaks are formed at the edges of the protruding portion. Accordingly, the stage 160 may be moved such that the peaks correspond to those of a preset profile. Thus, the stage 160 may be correctly positioned. FIG. 8B represents the profile in only the X. However, because the position measuring part 180c has a polygonal and/or cross shape, it is possible to adjust the stage 160 in both X and Y directions. FIG. 9 is a view illustrating an example embodiment of a method of radiating electron beams onto one or more surfaces of a plurality of photomasks M using an E-beam lithography system 100, for example. As shown in FIG. 9, an area R onto which an electron beam is to be radiated may be divided into a plurality of unit cells Xc. The electron beam may be radiated in stages, and each unit cell Xc may have the shape of a block having a unit distance Xu in the X direction and a unit distance Yu in the Y direction. The unit distance Xu in the X direction may be varied according to the performance of the E-beam lithography system 100 and/or the pattern data to be radiated. Particularly, the unit distance Xu may depend on the data processing capability of a processor of the E-beam lithography system 100, the capacity of memory, and/or the durability of hardware. Further, the unit distance Xu may depend on the compactness of the pattern to be irradiated using the electron beam, the variety of the shape, and/or the size of the radiated electron beam. Thus, the unit distance Xu does not refer to the distance based on the hardware, but instead, refers to a distance based on the software. That is, if a pattern to be irradiated is not compact, so that the physical quantity of radiated electron beams is relatively small, or electron beams having a relatively large size are radiated, and/or only a certain shape is radiated, the unit distance Xu may be increased accordingly. Further, the unit distance Xu may be increased by improving the hardware of the E-beam lithography system 100. That is, if the performance of the processor of the E-beam lithography system 100 is improved, the capacity of the memory is increased, and/or the durability of equipment is improved, the unit distance Xu may be increased. The unit distance Yu in the Y direction is defined as a width at which electron beams may be radiated in a single scanning operation of each irradiating system 170. While the stage 160 moves from the left side to the right side and vice versa, each irradiating system 170 reciprocates a desired and/or predetermined distance in a direction substantially perpendicular to the moving direction of the stage 160. In this case, the unit distance Yu is the distance reciprocating in one direction. If Yu is increased, the area on which electron beams are radiated at one time is increased, thus improving productivity. However, the E-beam radiating resolution may suffer. Conversely, if the unit distance Yu is reduced, the E-beam radiating resolution is increased but productivity is lowered. Thus, the unit distance Yu should be set to an appropriate value. In the example shown in FIG. 9, the same pattern may be formed on the left and right photomasks M by radiating electron beams. However, different patterns may be formed on the left and right photomasks M by radiating electron beams according to a different example embodiment. According to an example embodiment, each irradiating system 170 may include both a variable rectangular aperture and a circular aperture, which may be used to determine the shape of an electron beam. Thus, one irradiating system 170 may radiate an electron beam using the variable rectangular aperture, while another irradiating system 170 may radiate an electron beam using the circular aperture. Of course, the two irradiating systems 170 may radiate electron beams using the same aperture as well. If the two photomasks M have different patterns, different times may be spent radiating electron beams onto the photomasks M. The time spent radiating an electron beam onto one photomask M may be several times longer than the time spent radiating an electron beam onto the other photomask M. The times spent radiating electron beams onto both the photomasks M may be adjusted accordingly. For example, the time spent radiating the electron beams onto both the photomasks M may be adjusted to be equal to each other. In an E-beam lithography method according to an example embodiment, the area R onto which electron beams are radiated is divided into unit cells Xc. The electron beams are radiated onto the unit cells Xc in stages, thus making the entire E-beam radiating time constant. Further, to reduce the time spent radiating electron beams, an example embodiment of an irradiating system radiates electron beams while reciprocating in the X direction and moving in the Y direction in stages. An example embodiment of an E-beam lithography method is described below in stages with reference to FIG. 2. First, the photomasks M may be loaded onto a loading unit 110 of an example embodiment of an E-beam lithography system 100. A vacuum chamber door 115 connected to the vacuum chamber 120 may be opened, and the first robot arm Ra provided in the loading unit 110 of FIG. 2, for example, may transfer the loaded photomasks M into the vacuum chamber 120. Next, the vacuum chamber door 115 may be closed, and a vacuum pump (not shown) connected to the vacuum chamber 120 may operate to create a vacuum in the vacuum chamber 120. The vacuum level of the vacuum chamber 120 may be substantially similar to the vacuum levels of the other chambers 130, 140, and 150 shown in FIG. 2. Thereafter, the transfer chamber door 125 connected to the transfer chamber 140 may be opened, and the second robot arm Rb provided in the transfer chamber 140 may transfer the photomasks M into the alignment chamber 130. Accordingly, the alignment chamber door 135 may be opened. The photomasks M may be aligned on the cassette 175 in the alignment chamber 130. The step of aligning the photomasks M on the cassette 175 may use one or more of the four corners of the photomasks M, for example. The internal robot arm Rb may then transfer the photomasks M and/or the cassette 175, which have been aligned, into the lithography chamber 150, and may mount the photomasks M and/or the cassette 175 on the stage 160. The alignment chamber door 135 and the lithography chamber door 145 may be opened at this time. If the photomasks M and/or the cassette 175 have been mounted on the stage 160, the lithography chamber door 145 may be closed. The E-beam control system may adjust the stage 160 and the irradiating systems 170 using the E-beam measuring unit 180 to locate that stage 160 at an initial E-beam radiating position. The irradiating systems 170 may then begin radiating electron beams onto the surfaces of the photomasks M, which are mounted on the stage 160. The stage 160 may move up, down, left and right. In this case, the stage 160 does not move in only one direction while radiating electron beams, but instead, repeatedly performs the following while radiating electron beams. For example, as shown by the arrows in FIG. 9, the stage 160 moves in one X direction while being fixed in the Y direction. In such a state, electron beams are radiated. Next, the stage 160 moves by one unit distance in a Y direction while being fixed in the X direction. At this time, electron beams are not radiated. Then, the stage 160 moves in an X direction opposite that of the first step while being fixed in the Y direction and electron beams are radiated. Next, the stage 160 moves by a unit distance in the same Y direction while being fixed in the X direction. Electron beams are not radiated during the movement in the Y direction. These processes may be repeated. According to the example described above, the unit distance Yu moved in the Y direction may be kept constant; electron beams may be radiated onto a plurality of unit cells Xc in stages; and whenever the process of radiating an electron beam onto each unit cell Xc is completed, the E-beam radiating state of the irradiating systems 170 and/or the position of the stage 160 may be checked using the E-beam control system. In addition, the scope of the above example embodiments may be applied to a wafer manufacturing process for manufacturing a semiconductor device as well as a photomask. If applied to a wafer manufacturing process, a chuck for mounting a wafer and/or a wafer stage may be substituted for the cassette 175 and/or the stage 160. More chucks and/or stages may be desired and/or needed to perform a wafer manufacturing process. Further, a machining cassette suitable for the shape of a wafer may be separately manufactured and applied. Further, according to the number of wafers to be manufactured simultaneously, a plurality of irradiating systems 170, E-beam control systems and E-beam measuring units 180 may be provided. The positions of the components may be variously changed. For example, a plurality of irradiating systems 170 may be arranged side by side or in a lattice form. According to the arrangement of the irradiating systems 170, a plurality of E-beam measuring units 180 may be provided at various positions. As necessary, the E-beam control system may include a plurality of E-beam control systems. Although example embodiments have been described above for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and/or substitutions are possible, without departing from the scope and spirit of the example embodiments.
063296624	description	DETAILED DESCRIPTION OF THE INVENTION First, a radiographic intensifying screen employed for the combination of the invention is explained in detail. The radiographic intensifying screen according to the invention comprises a support and one or more phosphor layers provided thereon, and may have the same structure as a known intensifying screen. For example, a light-reflecting layer and/or a light-absorbing layer may be provided between the support and the phosphor layer. Preferably, a surface protective layer is provided on the phosphor layer. The fluorescent dye or pigment can be incorporated in any layer of the intensifying screen, but is preferably incorporated into the phosphor layer or the surface protective layer (particularly preferably into the phosphor layer) so that it can work effectively. The phosphor contained in the phosphor layer is a rare earth phosphor represented by the following formula (I): EQU M.sub.w O.sub.w X:M' (I) in which M represents at least one rare earth atom selected from the group consisting of Y, La, Gd and Lu; X represents at least one chalcogen atom selected from the group consisting of S, Se and Te, or at least one halogen atom selected from the group consisting of F, Br, Cl and I; M' represents a rare earth atom activating M; and w is 2 when X is chalcogen or w is 1 when X is halogen. In the formula (I), M' preferably is Dy, Er, Eu, Ho, Nd, Pr, Sm, Ce, Th, Tm and/or Yb, more preferably is Tb. The content of M' preferably is in the range of 0.0002 to 0.2 mol, more preferably 0.0005 to 0.05 mol, particularly preferably 0.001 to 0.02 mol, per 1 mol of M. The rare earth phosphor of the formula (I) may include a small amount of additives for improving certain properties and/or those for modifying the surface condition (such as silica and alumina). Concrete examples of the rare earth phosphors of the formula (I) are as follows: terbium activated rare earth oxysulfide phosphors [e.g., Y.sub.2 O.sub.2 S:Tb, Gd.sub.2 O.sub.2 S:Tb, La.sub.2 O.sub.2 S:Tb, (Y,Gd).sub.2 O.sub.2 S:Tb, and (Y,Gd).sub.2 O.sub.2 S:Tb, Tm]; terbium activated rare earth oxyhalide phosphors [e.g., LaOBr:Tb, LaOBr:Tb,Tm, LaOCl:Tb, LaOCl:Tb,Tm, GdOBr:Tb, and GdOCl:Tb]; and thulium activated rare earth oxyhalide phosphors [e.g., LaOBr:Tm and LaOCl:Tm]. Particularly preferred is a terbium activated gadolinium oxysulfide phosphor (Gd.sub.2 O.sub.2 S:Tb). In the phosphor of Gd.sub.2 O.sub.2 S:Tb, a portion of Gd (less than 50 atomic %) can be substituted with Y, La and/or Lu. Further, additives such as Ho may be included in an amount of less than 10 atomic % and the compound such as silica and alumina may be also incorporated to modify the surface condition of the phosphor. U.S. Pat. No. 3,725,704 gives a detailed description of Gd.sub.2 O.sub.2 S:Tb. FIG. 2 shows an emission spectrum of Gd.sub.2 O.sub.2 S:Tb which is observed when excited with X-ray (40 KVp, tungsten target tube). According to FIG. 2, Gd.sub.2 O.sub.2 S:Tb exhibits not only a main emission peak at approx. 545 nm but also plural emission peaks in the wavelength region of 380 to 500 nm. The fluorescent dye or pigment employed for the invention absorbs a portion of the luminescence of the phosphor, and then emits a light in the visible region. For example, in the case that the phosphor is Gd.sub.2 O.sub.2 S:Tb, the fluorescent dye or pigment absorbs a luminescence in the wavelength range of lower than approx. 545 nm, and then emits a light at approx. 545 nm. Preferably, the dye or pigment absorbs a luminescence in the wavelength region shorter than 500 nm, and then emits a light whose spectrum has a peak in the wavelength region of 450 to 600 nm, more preferably 490 to 600 nm, and more preferably 500 to 570 nm, under the condition that the wavelength of emission peak is longer than the wavelength of light absorption peak by at least 10 nm. The half-width of the light emitted by the fluorescent dye or pigment preferably is not more than 100 nm, more preferably not more than 80 nm, most preferably not more than 70 nm. The fluorescent dye or pigment converts a blue component (which causes the cross-over phenomenon) of the luminescence into a light near the maximum peak (which makes radiographic images) of the luminescence of the phosphor, and thereby effectively reduces the cross-over and remarkably enhances the sensitivity of the radiographic combination of the invention. It is preferred for the fluorescent dye or pigment of the invention to exhibit a high quantum yield. The quantum yield is defined by the following formula: EQU Quantum yield (%)=(number of emitted photons/number of absorbed photons).times.100. A fluorescent dye or pigment having a higher quantum yield is preferred. The quantum yield of the dye or pigment preferably is not lower than 20%, more preferably not lower than 40%, most preferably not lower than 60%. There are no specific restrictions on substances used as the fluorescent dye or pigment, and hence it may be an inorganic compound or an organic compound. In the case that the fluorescent dye or pigment is incorporated into the phosphor layer, the particle size of the dye or pigment preferably is small and is usually in the range of less than 2 .mu.m, preferably less than 1 .mu.m so as not to lower the packing density of the rare earth phosphor. Further, it is also preferred for the fluorescent dye or pigment not to agglomerate in the intensifying screen. For this reason, the fluorescent dye or pigment preferably exhibits a high quantum yield even in the form of small particles, and therefore an organic fluorescent dye or pigment is preferably used. As the fluorescent dye or pigment, known dyes and pigments are employable. Appropriate dyes and pigments are described in, for example, "Senryo Binran (Dye Handbook, in the Japanese language)" (pp. 315-1109, 1970, edited by Yukigosei-Kyokai) and "Shikizai-Kogaku Handbook (Dye Technology Handbook in the Japanese language)" (pp. 225-417, 1989, edited by Shikizai-kyokai). In "Laser dyes" (written by Mistuo Maeda, 1984, published by Academic Press), particularly preferred dyes are described. Examples of the preferred dyes include carbocyanine dyes (shown in Table 4 on pp. 26-29 of "Laser dyes"); phthalocyanine dyes (shown in Table 11 on pp. 74-75); xanthene dyes (shown in Table 12 on pp. 76-105); triarylmethane dyes (shown in Table 13 on pp. 106); acridine dyes (shown in Table 14 on pp. 107-110); condensed ring compounds (shown in Table 18 on pp. 137-149); coumarin and azacoumarin dyes (shown in Table 23 on pp. 189-238); quinolone and azaquinolone dyes (shown in Table 25 on pp. 239-246); oxazole and benzoxazole compounds (shown in Table 26 on pp. 247-261); furan and benzofuran compounds (shown in Table 29 on pp. 273-275); pyrazoline compounds (shown in Table 30 on pp. 276); phthalimido and naphthalimide compounds (shown in Table 31 on pp. 277); peteridine compounds (shown in Table 32 on pp. 282); and pyrylium, phosphorine, boradiazinium and pyridine compounds (shown in Table 33 on pp. 283). Further, diketopyrrolopyrrole compounds (described in Japanese Patent Provisional Publication No. 58-210084) and perylene compounds (described in Japanese Patent Provisional Publication No. H7-188178) are also preferred. The fluorescent dye or pigment preferably exhibits both an absorption spectrum having a maximum peak in the wavelength region of 350 to 500 nm and a fluorescence (emission) spectrum having a maximum peak in the wavelength region of 500 to 600 nm. More preferably, the maximum of the absorption spectrum is in the wavelength region of 400 to 490 nm and that of the fluorescence spectrum is in the wavelength region of 500 to 600 nm. Most preferably, the maximum of the absorption spectrum is in the wavelength region of 400 to 490 nm and that of the fluorescence spectrum is in the wavelength region of 500 to 570 nm. Therefore, carbocyanine dyes, xanthene dyes, triarylmethane dyes, acridine dyes, coumarin or azacoumarin dyes, phthalimido or naphthalimide compounds, pyrylium compounds, diketopyrrolopyrrole compounds or perylene compounds are preferred. The maximum peak of the fluorescence spectrum most preferably is in the wavelength region of 500 to 555 nm. Accordingly, carbocyanine dyes, xanthene dyes, triarylmethane dyes, coumarin dyes, phthalimido or naphthalimide compounds, diketopyrrolopyrrole compounds or perylene compounds are most preferred. The fluorescent dye or pigment preferably gives almost no (or very short) afterglow, and its fluorescence lifetime preferably is not longer than 10.sup.-2 second. It is further preferred that the fluorescent dye or pigment is hardly decomposed with heat, light or lapse of time. Concrete examples of the fluorescent dye or pigment employable for the invention are given below, but those examples by no means restrict the invention. 1 ##STR1## 2 ##STR2## 3 ##STR3## 4 ##STR4## No. R.sup.1 R.sup.2 R.sup.3 R.sup.4 X.sup.- 5 OCH.sub.3 H COOH H Cl 6 OCH.sub.3 C.sub.6 H.sub.13 COOH H Cl 7 NH.sub.2 H COOH H Cl 8 N(C.sub.2 H.sub.5).sub.2 H COOH H ClO.sub.4 9 OCH.sub.3 C.sub.6 H.sub.13 COOC.sub.2 H.sub.5 H BF.sub.6 10 ##STR5## 11 ##STR6## 12 ##STR7## 13 ##STR8## 14 ##STR9## 15 ##STR10## 16 ##STR11## 17 ##STR12## 18 ##STR13## 19 ##STR14## 20 ##STR15## 21 ##STR16## 22 ##STR17## 23 ##STR18## 24 ##STR19## 25 ##STR20## 26 ##STR21## 27 ##STR22## 28 ##STR23## 29 ##STR24## 30 ##STR25## 31 ##STR26## 32 ##STR27## 33 ##STR28## 34 ##STR29## 35 ##STR30## 36 ##STR31## 37 ##STR32## 38 ##STR33## 39 ##STR34## 40 ##STR35## 41 ##STR36## 42 ##STR37## 43 ##STR38## 44 ##STR39## 45 ##STR40## 46 ##STR41## 47 ##STR42## 48 ##STR43## 49 ##STR44## 50 ##STR45## 51 ##STR46## ##STR47## 52 R = m-CF.sub.3 53 m-Cl 54 p-Br 55 p-N(CH.sub.3).sub.2 56 p-N(C.sub.2 H.sub.5).sub.2 ##STR48## ##STR49## 57 ##STR50## 58 ##STR51## 59 ##STR52## 60 ##STR53## 61 ##STR54## 62 ##STR55## 63 ##STR56## 64 ##STR57## 65 ##STR58## 66 ##STR59## 67 ##STR60## The phosphor layer can be formed, for example, in the following manner. The above-described phosphor particles and, if the fluorescent dye or pigment is to be incoporated into the phosphor layer, the particles of the fluorescent dye or pigment are dispersed in an organic solvent together with a binder resin, to prepare a coating dispersion. The coating dispersion is then coated over the support (or over an undercoating layer such as light-reflecting layer beforehand provided on the support), and dried to form a phosphor layer. Besides the above method, the phosphor layer may be formed by the steps of coating the above dispersion on a temporary support, drying the coated dispersion to form a phosphor sheet, peeling off the sheet, and fixing the sheet onto a permanent support with an adhesive. The size of the phosphor particles is not specifically restricted, but is usually in the range of approx. 1 to 15 .mu.m, preferably approx. 2 to 10 .mu.m. Preferably, the phosphor particles are highly packed in the phosphor layer, and the volume filling content of the phosphor is usually in the range of 60 to 85%, preferably 65 to 80%, more preferably 68 to 75%. The content of the phosphor particles in the phosphor layer is usually not less than 80 wt. %, preferably not less than 90 wt. %, more preferably not less than 95 wt. %. The fluorescent dye or pigment is usually used in an amount of 0.1 to 5,000 mg, preferably 1 to 1,000 mg, more preferably 5 to 500 mg per 1 kg of the phosphor. Many known publications and references describe binder resins, organic solvents and various additives employable for the phosphor layer of the intensifying screen. The thickness of the phosphor layer can be set according to the target sensitivity, but preferably is 70 to 150 .mu.m for a front side screen or 80 to 400 .mu.m for a back side screen. The X-ray absorption efficiency depends on the amount of the phosphor particles in the phosphor layer. The phosphor layer may be a single layer or may have multi-layered structure. Preferably, the phosphor layer consists of one to three component layers, and more preferably consists of one or two component layers. For example, the phosphor layer may consist of plural component layers each of which contains the phosphor particles of different size. In that case, the size of the phosphor particles contained in each component layer may gradually increase from the bottom component layer (layer provided next to the support) to the top component layer. In this phosphor layer, the more deeply a component layer is provided, the smaller particles it contains. Otherwise, the phosphor layer may contain a mixture of phosphor particles having different sizes. Further, the size distribution of the phosphor particles may vary in the depth direction of the phosphor layer (Japanese Patent Publication No. 55-33560). For example, in the case that Gd.sub.2 O.sub.2 S:Tb phosphor (whose variation coefficient of size distribution is usually in the range of 30 to 50%) is employed, mono-dispersive particles (whose variation coefficient is not more than 30%) are preferably used. The support of the intensifying screen can be optionally selected from those employed in the conventional radiographic intensifying screens. Examples of the supports include polymer films containing a white pigment (e.g., titanium dioxide) or a black pigment (e.g., carbon black). The phosphor layer may be directly provided on the surface of the support, or the phosphor layer may be provided via an undercoating layer such as a light-reflecting layer containing light-reflecting material (e.g., titanium dioxide). For example, Examples 1 and 2 in Japanese Patent Provisional Publication No. H9-21899 disclose a light-reflecting layer preferably employable for the intensifying screen of the invention. The light-reflecting layer contains titanium dioxide (mean particle size: 0.1 to 0.5 .mu.m) in a volume filling content of 10 to 75%, and the thickness is in the range of 10 to 100 .mu.m. On the phosphor layer, the surface protective layer is preferably formed. The surface protective layer preferably exhibits scattering with a scattering length of 5 to 80 .mu.m observed at a maximum wavelength of the luminescence emitted by the phosphor. The scattering length more preferably is in the range of 10 to 70 .mu.m, most preferably 10 to 60 .mu.m. In this specification, the word "scattering length" indicates a mean distance in which a light travels straight until it is scattered, and therefore a short scattering length means that the layer highly scatters a light. On the other hand, the absorption length (which indicates a mean distance in which a light travels straight until it is absorbed) of the surface protective layer is not restricted. From the viewpoint of sensitivity of the intensifying screen, it is preferred for the protective layer not to absorb a light. However, in order to make up for shortage of the scattering, the surface protective layer may be made to slightly absorb a light. The absorption length preferably is more than 800 .mu.m, more preferably more than 1,200 .mu.m. In accordance with Kubeluka-Munk theory, the scattering length and the absorption length can be calculated from the data obtained in the following measurement. First, three or more film samples are prepared. Each film sample has a different thickness, but made of the same components as the target surface protective layer. The thickness (.mu.m) and the diffuse transmittance (%) of each sample are then measured. The diffuse transmittance (%) can be measured by means of a spectrophotometer equipped with an integrating sphere. In the below-described examples of the present specification, an automatic recording spectrophotometer (U-3210, manufactured by HITACHI, Ltd.) equipped with an integrating sphere of 150 .phi. (150-0910) is used. The diffuse transmittance must be measured at the wavelength corresponding to the maximum peak of the luminescence emitted by the phosphor contained in the phosphor layer on which the target surface protective layer is provided. From the thickness (.mu.m) and the diffuse transmittance (%) obtained in the above measurements, the scattering length is calculated in accordance with the following formula (A) which is derived from Kubeluka-Munk theory. The formula (A) can be easily derived, under the boundary condition of the diffuse transmittance (%), from the formulas 5.1.12 to 5.1.15 described in "Keikotai Handbook [in Japanese, Handbook of phosphor]", published by Ohm-sha, 1987, pp.403. Formula (A): EQU T/100=4.beta./[(1+.beta.).sup.2.multidot.exp(.alpha.d)-(1-.beta.).sup. 2.multidot.exp(-.alpha.d)] in which T represents a diffuse transmittance (%), d represents a thickness (.mu.m), and .alpha. and .beta. are defined by the formulas: .alpha.=[K(K+2S)].sup.1/2 and .beta.=[K/(K+2S)].sup.1/2, respectively. The formula (A) is applied to the measured T (diffuse transmittance) and d (thickness) of each film sample, and thereby the values of K and S are determined. The scattering length (.mu.m) and the absorption length (.mu.m) described above are values defined by 1/S and 1/K, respectively. The surface protective layer preferably contains light-scattering fine particles dispersed in a binder resin. The refractive index of the light-scattering fine particles generally is more than 1.6, preferably more than 1.9, and the particle size generally is in the range of 0.1 to 1 .mu.m. Examples of the light-scattering fine particles include fine particles of magnesium oxide, zinc oxide, zinc sulfide, titanium dioxide, niobium oxide, barium sulfate, lead carbonate, silicon dioxide, poly(methyl methacrylate), styrene, and melamine. Preferred are zinc oxide, zinc sulfide, titanium oxide and lead carbonate, because they have a high refractive index. Titanium dioxide of anatase type is particularly preferred. The binder resin employable for the surface protective layer is not specifically restricted, and examples of the binders include polyethylene terephthalate, polyethylene naphthalate, polyamide, aramid, fluorocarbon resins and polyesters. The surface protective layer can be formed by the steps of dispersing the light-scattering particles in an organic solution of the binder resin to prepare a coating liquid, coating the liquid on the phosphor layer directly or via an optionally provided auxiliary layer, and then drying the coated liquid to form a protective layer. The surface protective layer may be formed by other steps, namely, coating the above liquid on a temporary support, drying the coated liquid to form a protective sheet, peeling off the protective sheet from the temporary support, and then placing the protective sheet with an adhesive on the phosphor layer. The thickness generally is in the range of 2 to 12 .mu.m, preferably 3.5 to 10 .mu.m. The aforementioned fluorescent dye or pigment may be incorporated into the coating liquid to prepare a surface protective layer containing the fluorescent dye of pigment. With respect to the process for producing intensifying screens and materials preferably employed for the invention, reference is made to various publications (e.g., Japanese Patent Provisional Publications No. H9-21899 and No. H6-347598). The silver halide photographic material employed for the radiation image forming combination of the invention is now explained in detail. The photographic material gives a cross-over in a degree of not more than 10%, preferably not more than 8% in the radiographic process. In the present specification, the degree (value) of cross-over is determined in the following manner. First, an intensifying screen and a "both-sided emulsion type" photographic material (having an emulsion layer on both sides) are prepared. The photographic material is placed in contact with the front surface of the screen, and then a sheet of black paper is further placed on the front side (i.e., the side other than at in contact with the scree n) of the photographic material. In this arrangement, the photographic material is exposed to X-rays with different doses (which are adjusted by varying the distance between the intensifying screen and the focal spot of the X-ray generator). The exposed material is developed, and then the developed material is divided into two samples. From one sample is removed an emulsion layer which has been in contact with the intensifying screen. From another sample is removed an emulsion layer which has not been in contact with the intensifying screen. The optical density of the emulsion layer remaining on the support is measured and plotted against the corresponding dose to obtain a characteristic curve. Based on thus obtained characteristic curve for the emulsion layer remaining on each support, the average sensitivity difference (.DELTA.logE) is estimated from the liner part of the curve, and the degree of cross-over is calculated in accordance with the following formula: EQU cross-over (%)=100/[antilog(.DELTA.logE)+1]. In the present invention, in order to reduce the cross-over to 10% or below, the photographic material preferably contains a dye whose absorption spectrum exhibits both the maximum peak in the wave length region of 500 to 600 nm and the absorption coefficient at 550 nm twice or more larger than that at 450 nm. More preferably, the absorption coefficient at 550 nm is three times or more larger than that at 450 nm. In the conventional screen containing Gd.sub.2 O.sub.2 S:Tb phosphor, the incorporated dye must absorb a light not only at the wavelength of the luminescence maximum (approx. 545 nm) but also in the wavelength region of 380 to 570 nm in order to sufficiently reduce the cross-over. In the combination of the invention, however, even if the dye has a small absorption coefficient in a short wavelength region (not longer than 500 nm), the cross-over can be sufficiently reduced. Further, a light in the wavelength region of 500 to 600 nm can be effectively absorbed by the above-mentioned dye, and accordingly the cross-over can be efficiently reduced in a relatively small amount of the dye. There is no specific restriction on the process for producing a photographic material containing a dye having a relatively sharp absorption peak in the wavelength region of 500 to 600 nm, and hence various methods are employable. For example, the dye is adsorbed on silver halide fine particles so as to exhibit a sharp absorption peak, and then a coating liquid containing the particles thus treated with the dye is coated on an undercoating layer (Japanese Patent Provisional Publication No. H2-29641). Otherwise, the dye may be adsorbed on transparent inorganic material such as mica so as to exhibit a sharp absorption peak, and then incorporated into a photographic material. A particularly preferred method is described in Japanese Patent Provisional Publication No. H1-172828. In the method, solid fine particles of a dye which can be decolorized in the developing process is dispersed and fixed in the emulsion layer. The "dye which can be decolorized in the developing process" here means a dye which sufficiently absorbs a light before the developing process but whose absorbance in the visible region lowers to 0.4 or less, preferably 0.25 or less after the developing, fixing, and washing (including stabilizing) processes. With respect to the dye which can be decolorized in the developing process, a detailed description is given below. As the above dyes, known dyes or pigments such as compounds described in "Senryo Binran (Dye Handbook, in the Japanese language)" (pp. 315-1109, 1970, edited by Yukigosei-Kyoukai) or "Shikizai-Kogaku Handbook (Dye Technical Handbook, in the Japanese language)" (pp. 225-417, 1989, edited by Shikizai-kyokai) can be used. Preferred are dyes represented by the following formula (FA): EQU D-(X.sub.1).sub.y1. (FA) In the formula (FA), D represents a group derived from a compound having chromophore. X.sub.1 represents a dissociative proton connecting to D directly or via a divalent connecting group; or otherwise X.sub.1 represents a group having a dissociative proton. In the case that the dissociative proton is attached to D via a divalent connecting group, X.sub.1 includes the connecting group. In the formula (FA), y.sup.1 represents an integer of 1 to 7. The compound having chromophore (from which D is derived) can be optionally selected from various known dye compounds. Examples of the dye compounds include oxonol dyes, merocyanine dyes, cyanine dyes, arylidene dyes, azomethine dyes, triphenylmethane dyes, azo dyes, anthraquinone dyes, and indoaniline dyes. When the compound of the formula (FA) is incorporated into a photographic material, X.sub.1 (group having a dissociative proton or a dissociative proton which may have a connecting group) does not dissociate and hence the compound is substantially insoluble in water. During the process for developing the material (i.e., during the developing process), however, X.sub.1 dissociates to make a compound substantially soluble in water. Examples of the group represented by X.sub.1 include carboxylic acid groups, sulfonamide groups, arylsulfamoyl groups, sulfonylcarbamoyl group, carbonylsulfamoyl group, enol group of oxonol dye, and phenolic hydroxyl group. Examples of the compounds represented by the formula (FA) preferably employed for the invention include those of the following formulas (FA1), (FA2) and (FA3): EQU A.sub.1 =L.sub.1 -(L.sub.2 =L.sub.3).sub.m -Q (FA1) EQU A.sub.1 =L.sub.1 -(L.sub.2 =L.sub.3).sub.n -A.sub.2 (FA2) EQU A.sub.1 =(L.sub.1 -L.sub.2).sub.k =B.sub.1 (FA3). In the above formulas, each of A.sub.1 and A.sub.2 represents an acidic nucleus. B.sub.1 represents a basic nucleus and Q represents an aryl group or a heterocyclic group. Each of L.sub.1, L.sub.2 and L.sub.3 represents a methine group, "m" is 0, 1 or 2, and each of "n" and "k" is 0, 1, 2 or 3. The compound represented by the formula (FA1), (FA2) or (FA3) contains at least one group selected from the group consisting of carboxylic acid group, sulfonamide group, arylsulfamoyl group, sulfonylcarbamoyl group, carbonylsulfamoyl group, enol group of oxonol dye, and phenolic hydroxyl group. However, the compound does not contain water-soluble groups (e.g., sulfonic acid group, phosphoric acid group) other than the above groups. The acidic nucleus represented by A.sub.1 or A.sub.2 preferably is a cyclic ketomethylene compound or a compound having a methylene group placed between electron attractive groups. Examples of the cyclic ketomethylene compounds include 2-pyrazoline-5-one, rhodanine, hydantoin, thiohydantoin, 2,4-oxazolidinedione, isooxazolone, barbituric acid, thiobarbituric acid, indandione, dioxopyrazolopyridine, hydroxypyridone, pyrazolidinedione, and 2,5-dihydrofuran-2-one. Those compounds may have substituent groups. The compound having a methylene group placed between electron attractive groups can be represented by the following formula: Z.sub.1 CH.sub.2 Z.sub.2. In the formula, each of Z.sub.1 and Z.sub.2 represents --CN, --SO.sub.2 R.sub.1, --COR.sub.1, --COOR.sub.2, --CONHR.sub.2, --SO.sub.2 NHR.sub.2, --C[.dbd.C (CN).sub.2 ]R.sub.1 or --C[.dbd.C(CN).sub.2 ]NHR.sub.1. In the formula, R.sub.1 represents an alkyl group, an aryl group or a heterocyclic group, and R.sub.2 represents hydrogen atom or the groups represented by R.sub.1. The above groups may have substituent groups. Examples of the basic nuclea represented by B.sub.1 include pyridine, quinoline, indolenine, oxazole, imidazole, thiazole, benzooxazole, benzoimidazole, benzothiazole, oxazoline, naphthooxazole, and pyrrole. Those compounds may have substituent groups. Examples of the aryl groups represented by Q include phenyl and naphthyl. Those compounds may have substituent groups. Examples of the heterocyclic groups represented by Q include pyrrole, indole, furan, thiophene, imidazole, pyrazole, indolizine, quinoline, carbazole, phenothiazine, phenoxazine, indoline, thiazole, pyridine, pyridazine, thiadiazine, pyran, thiopyran, oxadiazole, benzoquinoline, thiazoazole, pyrrolothiazole, pyrrolopyridazine, tetrazole, oxazole, coumarin, and coumarone. Those compounds may have substituent groups. The methine group represented by L.sub.1, L.sub.2 or L.sub.3 may have substituent groups, and the substituent groups may be combined with each other to form a five- or six-membered ring (e.g., cyclopentene and cyclohexene). The substituent groups that the above-mentioned compounds or groups may have are not restricted unless they make the compounds of the formula (FA) [(FA1), (FA2) or (FA3)] substantially soluble in water of pH 5 to pH 7. Examples of the substituent groups include carboxylic acid group, sulfonamide group having 1-10 carbon atoms (e.g., methanesulfonamide, benzenesulfonamide, butanesulfonamide, and n-octanesulfonamide), sulfamoyl group having 0-10 carbon atoms (e.g., unsubstituted sulfamoyl, methylsulfamoyl, phenylsulfamoyl, and butylsulfamoyl), sulfonylcarbamoyl group having 2-10 carbon atoms (e.g., methanesulfonylcarbamoyl, propanesulfonylcarbamoyl, and benzenesulfonylcarbamoyl), acylsulfamoyl group having 1-10 carbon atoms (e.g., acetylsulfamoyl, propionylsulfamoyl, pivaloylsulfamoyl, and benzoylsulfamoyl), linear or cyclic alkyl group having 1-8 carbon atoms (e.g., methyl, ethyl, isopropyl, butyl, hexyl, cyclopropyl, cyclopentyl, cyclohexyl, 2-hydroxyethyl, 4-carboxybutyl, 2-methoxyethyl, benzyl, phenetyl, 4-carboxybenzyl, and 2-diethylaminoethyl), alkenyl group having 2-8 carbon atoms (e.g., vinyl, allyl), alkoxy group having 1-8 carbon atoms (e.g., methoxy, ethoxy, butoxy), halogen atom (e.g., F, Cl, Br), amino acid group having 0-10 carbon atoms (e.g., unsubstituted dimethylamino, diethylamino, and carboxyethylamino), alkoxycarbonyl group having 2-10 carbon atoms (e.g., methoxycarbonyl), amide group having 1-10 carbon atoms (e.g., acetylamino and benzamide), carbamoyl group having 1-10 carbon atoms (e.g., unsubstituted carbamoyl, methylcarbamoyl, and ethylcarbamoyl), aryl group having 6-10 carbon atoms (e.g., phenyl, naphthyl, 4-carboxyphenyl, 3-carboxyphenyl, 3,5-dicarboxyphenyl, 4-methanesulfonamidophenyl, an 4-butanesulfonamidophenyl), aryloxy group having 6-10 carbon atoms (e.g., phenoxy, 4-carboxyphenoxy, 3-methylphenoxy, and naphthoxy), alkylthio group having 1-8 carbon atoms (e.g., methylthio, ethylthio, and octylthio), arylthio group having 6-10 carbon atoms (e.g., phenylthio and naphthylthio), acyl group having 1-10 carbon atoms (e.g., acetyl, benzoyl, and propanoyl), sulfonyl group having 1-10 carbon atoms (e.g., methanesulfonyl and benzenesulfonyl), ureido group having 1-10 carbon atoms (e.g., ureido and methylureido), urethane group having 2-10 carbon atoms (e.g., methoxycarbonylamino and ethoxycarbonylamino), cyano group, hydroxyl group, nitro group, and heterocyclic group (e.g., 5-carboxybenzoxazole ring, pyridine ring, sulfolane ring, pyrrole ring, pyrrolidine ring, morpholine ring, piperazine ring, pyrimidine ring, and furan ring). The dyes of the formula (FA) preferably employed in the photographic material of the invention are described in Japanese Patent Provisional Publication No. H4-45436. Concrete examples of the compounds of the formula (FA) [(FA1), (FA2) or (FA3)] are as follows: F-1 ##STR61## F-2 ##STR62## F-3 ##STR63## F-4 ##STR64## F-5 ##STR65## F-6 ##STR66## F-7 ##STR67## F-8 ##STR68## F-9 ##STR69## F-10 ##STR70## F-11 ##STR71## F-12 ##STR72## F-13 ##STR73## F-14 ##STR74## F-15 ##STR75## F-16 ##STR76## F-17 ##STR77## F-18 ##STR78## F-19 ##STR79## F-20 ##STR80## F-21 ##STR81## F-22 ##STR82## F-23 ##STR83## ##STR84## No. R.sup.1 R.sup.2 n Q F-24 --CN ##STR85## 0 ##STR86## F-25 --CN ##STR87## 0 ##STR88## F-26 --CN ##STR89## 0 ##STR90## F-27 --CN ##STR91## 0 ##STR92## F-28 --CN ##STR93## 1 ##STR94## F-29 ##STR95## ##STR96## 1 ##STR97## F-30 ##STR98## ##STR99## 1 ##STR100## F-31 ##STR101## ##STR102## 1 ##STR103## F-32 ##STR104## ##STR105## 0 ##STR106## F-33 ##STR107## ##STR108## 0 ##STR109## F-34 --CN ##STR110## 0 ##STR111## F-35 ##STR112## F-36 ##STR113## F-37 ##STR114## F-38 ##STR115## F-39 ##STR116## F-40 ##STR117## F-41 ##STR118## F-42 ##STR119## F-43 ##STR120## F-44 ##STR121## F-45 ##STR122## F-46 ##STR123## F-47 ##STR124## F-48 ##STR125## F-49 ##STR126## F-50 ##STR127## F-51 ##STR128## F-52 ##STR129## F-53 ##STR130## F-54 ##STR131## F-55 ##STR132## In Japanese Patent Provisional Publication No. H7-152112, the compounds of the formulas (FA1), (FA2) and (FA3) are described. The compounds of (II-2) to (II-24), (III-5) to (III-18) and (IV-2) to (IV-7) in the publication are those of the formulas (FA1), (FA2) and (FA3), respectively. As the dye which can be decolorized in the developing process, cyanine dyes, pyrylium dyes and aminium dyes described in Japanese Patent Provisional Publication No. H3-138640 are also employable for the invention in the form of dispersed solid fine particles. The dyes employable for the invention can be prepared in the manner described in the following publications: WO 88/04794; EP 0274723A1, EP 276566, EP 299435; Japanese Patent Provisional Publications No. 52-92716, No. 55-155350, No. 55-155351, No. 61-205934, No. 48-68623; U.S. Pat. Nos. 2,527,583, 3,486,897, 3,746,539, 3,933,798, 4,130,429, and No. 4,040,841; Japanese Patent Provisional Publications No. H2-282244, No. H3-7931, No. H3-167546, No. H1-266536, No. H3-136038, No. H3-226736, No. H3-138640, and No. H3-211542; Japanese Patent Applications No. H6-227982, No. H6-227983, No. H6-279297, No. H7-54026, No. H7-101968, and No. H7-135118; and Japanese Patent Provisional Publications No. H2-282244, No. H7-113072 and No. H7-53946. In the invention, the dye which can be decolorized in the developing process is preferably used in the form of solid fine particles. The dye can be mechanically pulverized and dispersed by conventional means (e.g., ball mill, shaking ball mill, floating ball mill, sand mill, colloid mill, jet mill, roller mill) in the presence of a dispersing aid. As the dispersing aid, the known polymer compounds are employable, and if desired, two or more of polymer compouns may be used in combination. Further, other compounds such as anionic, nonionic or cationic surface active agents and polymers may be used in combination, but preferably the polymer compounds are used alone. The dispersing aid is generally mixed with a powder or wet-cake of the dye to prepare a slurry before the dispersing process, and then the slurry is introduced into the dispersing means. The dispersing aid may be beforehand mixed with the dye, and then the mixture is treated with heat or a solvent to prepare a powder or wet-cake. Further, the dispersing aid may be added into the dispersion liquid while the dye is being pulverized in the dispersing means. In order to stabilize the dispersion liquid, the dispersing aid may be added after the dispersing process. In any case, a solvent (e.g., water, alcohols) is generally used together with the dispersing aid. The pH value may be controlled using a proper pH adjusting agent before, during or after the dispersing process. The dye may be pulverized and dispersed by the steps of controlling the pH value to dissolve the dye in a solvent, and then varying the pH in the presence of a dispersing aid to deposit thereon fine particles of the dye. As the solvent, an organic solvent is employable. After the above steps, the organic solvent is usually removed. After the dispersing process, the prepared dispersion can be stored under stirring or under a highly viscous condition in the presence of a hydrophilic colloid (for example, in the form of jelly with gelatin) so that the fine particles may not precipitate. Preferably, antiseptics are added so as to prevent microbes from increasing during storage. The mean particle size of the dye generally is in the range of 0.005 to 10 .mu.m, preferably 0.01 to 3 .mu.m, more preferably 0.05 to 0.5 .mu.m. The dye may be incorporated into any part of the photographic material, but it is preferred to incorporate the dye into an layer which is provided on the support. For instance, the dye can be incorporated into an undercoating hydrophilic colloid layer and/or a photo-insensitive hydrophilic colloid layer. Those layers are provided between the support and the silver halide photosensitive emulsion layer. If the photosensitive emulsion layer comprises two or more photosensitive emulsion layers, the dye can be incorporated into a lower photosensitive emulsion layer, that is an emulsion layer near to the support surface. Generally, two or more undercoating hydrophilic colloid layers are formed on the support in the following manner. On the support, a hydrophilic colloid liquid is coated and then dried to form a bottom undercoating hydrophilic colloid layer. After that, on the formed layer, the liquid is again coated and then dried to form a next layer. Those procedures are repeated to form plural undercoating hydrophilic colloid layers. On thus formed undercoating hydrophilic colloid layers, a photo-insensitive hydrophilic colloid layer and a photosensitive silver halide emulsion layer are formed at the same time by a simultaneous superposing coating method. Therefore, before the photo-insensitive layer is dried, the emulsion layer is provided thereon. The undercoating layer and the photo-insensitive layer (provided between the undercoating layers and the emulsion layer) may contain the same dye or different dyes. Further, plural dyes may be incorporated into one layer. In that case, the pulverizing and dispersing method, the mean particle size and the particle shape may be also the same or different from each other. The amount of the dye depends on the target absorbance and absorption coefficient of the dispersion, but is usually in the range of 0.005 to 2 g/m.sup.2, preferably 0.03 to 2 g/m.sup.2, more preferably 0.04 to 0.15 g/m.sup.2, most preferably 0.05 to 0.12 g/m.sup.2 (per 1 m.sup.2 of the support, total amount for both faces). The dye may be incorporated into the layer(s) provided on only one side (either the front side or the back side) of the support. There are no specific restrictions on materials for the hydrophilic colloid, but usually gelatin is preferably used. The undercoating hydrophilic colloid layer containing the dye preferably contains gelatin in an amount of 0.05 to 0.3 g, more preferably 0.05 to 0.2 g per 1 m.sup.2 of one surface of the support. The photo-insensitive hydrophilic colloid layer containing the dye preferably contains gelatin in an amount of 0.05 to 0.5 g, more preferably 0.1 to 0.4 g per 1 m.sup.2 of one surface of the support. The total amount of the hydrophilic colloid contained in the layers (i.e., undercoating layer containing the dye, photo-insensitive layer containing the dye, emulsion layer and surface protective layer) preferably is in the range of 1.4 to 2.6 g per 1 m.sup.2 of one surface of the support. At least one of the undercoating layer and the photo-insensitive layer described above preferably contains a gelatin-reactive polymer latex. The term of "gelatin-reactive polymer latex" here means a polymer whose surface has a group chemically reactive with a terminal group of gelatin. In the invention, the polymer latex having an active methylene group is preferably employed, and is represented by the following formula (II): EQU --(C.sup.1).sub.x --(A.sup.1).sub.y --(B.sup.1).sub.z -- (II). In the formula (II), C.sup.1 represents a repeating unit derived from an ethylenic unsaturated monomer having an active methylene group, A.sup.1 represents a different repeating unit derived from an ethylenic unsaturated monomer whose homopolymer has a glass transition point of not higher than 35.degree. C., B.sup.1 represents a different repeating unit derived from an ethylenic unsaturated monomer, and the repeating units of C.sup.1, A.sup.1 and B.sup.1 are different each other. Each of x, y and z represents a weight percent ratio of the repeating unit C.sup.1, A.sup.1 or B.sup.1, respectively. The values of x, y and z are in the ranges of 0.5 to 40, 60 to 99.5, and 0 to 50, respectively, under the condition of x+y+z=100. The ethylenic unsaturated monomer giving the repeating unit C.sup.1 can be represented by the following formula: ##STR133## In the formula, R.sup.1 represents a hydrogen atom, an alkyl group having 1-4 carbon atoms (e.g., methyl, ethyl, n-propyl, and n-butyl) or a halogen atom (e.g., chlorine and bromine). Preferably, R.sup.1 is a hydrogen atom, methyl group, or chlorine atom. Further, X represents a monovalent group having an active methylene group (described later in detail). L is a single bond or a divalent connecting group having the following formula: EQU --(L.sup.1).sub.m1 --(L.sup.2).sub.m --. In the formula, L.sup.1 represents --CON(R.sup.2)-- (in which R.sup.2 is a hydrogen atom, an alkyl group having 1-4 carbon atoms, or a substituted alkyl group having 1-6 carbon atoms), --COO--, --NHCO--, --OCO--, or a group represented by one of the following formulas: ##STR134## In the formulas, each of R.sup.3 and R.sup.4 independently represents a hydrogen atom, a hydroxyl group, a halogen atom, a substituted or unsubstituted alkyl, an alkoxy, acyloxy or aryloxy group; and R.sup.2 represents the same as described above. L.sup.2 is a group connecting L.sup.1 and X, and is represented by the following formula: EQU --[X.sup.1 --(J.sup.1 --X.sup.2).sub.p --(J.sup.2 --X.sup.3).sub.q --(J.sup.3).sub.r --].sub.s --. In the formula, J.sup.1, J.sup.2 and J.sup.3 may be the same or different from each other, and each of them independently represents --CO--, --SO.sub.2 --, --CON(R.sup.5)-- (in which R.sup.5 is a hydrogen atom, an alkyl group having 1-6 carbon atoms or a substituted alkyl group having 1-6 carbon atoms), --O.sub.2 N(R.sup.5)-- (in which R.sup.5 represents the same as described above), --N(R.sup.5)--R.sup.6 -- (in which R.sup.5 represents the same as described above, and R.sup.6 is an alkylene group having 1-4 carbon atoms), --N(R.sup.5)--R.sup.6 --N(R.sup.7)-- (in which each of R.sup.5 and R.sup.6 represents the same as described above, and R.sup.7 is a hydrogen atom, an alkyl group having 1-6 carbon atoms or a substituted alkyl group having 1-6 carbon atoms), --O--, --S--, --N(R.sup.5)--CO--N(R.sup.7)-- (in which each of R.sup.5 and R.sup.7 represents the same as described above), --N(R.sup.5)--SO.sub.2 --N(R.sup.7)-- (in which each of R.sup.5 and R.sup.7 represents the same as described above), --COO--, --OCO--, --N(R.sup.5)--CO.sub.2 -- (in which R.sup.5 represents the same as described above), and --N(R.sup.5)CO-- (in which R.sup.5 represents the same as described above). Each of "p", "q", "r" and "s" represents 0 or 1. X.sup.1, X.sup.2 and X.sup.3 may be the same or different from each other, and each of them independently represents a substituted or unsubstituted alkylene group having 1-10 carbon atoms, an aralkylene group or a phenylene group. The alkylene group may be a straight chain or branched chain. Examples of the alkylene groups include methylene, methylmethylene, dimethylmethylene, dimethylene, trimethylene, tetramethylene, pentamethylene, hexamethylene, and decylmethylene. Examples of the aralkylene groups include benzylidene, and examples of the phenylene groups include p-phenylene group, m-phenylene group and methylphenylene group. Preferred examples of the monovalent groups of X include R.sup.8 --CO--CH.sub.2 --COO--, NC--CH.sub.2 --COO--, R.sup.8 --CO--CH.sub.2 --CO--, and R.sup.8 --CO--CH.sub.2 --CON(R.sup.5)--. In these formulas, R.sup.5 represents the same as described above, and R.sup.8 represents a substituted or unsubstituted alkyl group having 1-12 carbon atoms (e.g., methyl, ethyl, n-propyl, n-butyl, t-butyl, n-nonyl, 2-methoxyethyl, 4-phenoxybutyl, benzyl, and 2-methanesulfonamidoethyl), a substituted or unsubstituted aryl group (e.g., phenyl, p-methylphenyl, p-methoxyphenyl, and o-chlorophenyl), alkoxy group (e.g., methoxy, ethoxy, methoxyethoxy, and n-butoxy), a cycloalkyloxy group (e.g., cyclohexyloxy), an aryloxy group (e.g., phenoxy, p-methylphenoxy, o-chlorophenoxy, p-cyanophenoxy), an amino acid, and a substituted amino group (e.g., methylamino, ethylamino, dimethylamino, and butylamino). Examples of the ethylenic unsaturated monomer giving C.sup.1 are given below, but the following examples by no means restrict the invention. M-1 2-acetoacetoxyethyl methacrylate PA1 M-2 2-acetoacetoxyethyl acrylate PA1 M-3 2-acetoacetoxypropyl methacrylate PA1 M-4 2-acetoacetoxypropyl acrylate PA1 M-5 2-acetoacetamidoethyl methacrylate PA1 M-6 2-acetoacetamidoethyl acrylate PA1 M-7 2-cyanoacetoxyethyl methacrylate PA1 M-8 2-cyanoacetoxyethyl acrylate PA1 M-9 N-(2-cyanoacetoxyethyl)acrylamide PA1 M-10 2-propionylacetoxyethyl acrylate PA1 M-11 N-(2-propionylacetoxyethyl)methacrylamide PA1 M-12 N-4-(acetoacetoxybenzyl)phenylacrylamide PA1 M-13 ethylacryloyl acetate PA1 M-14 acryloylmethyl acetate PA1 M-15 N-methacryloyloxymethylacetoacetamide PA1 M-16 ethylmethacryloyl acetoacetate PA1 M-17 N-allylcyanoacetamide PA1 M-18 methylacryloyl acetoacetate PA1 M-19 N-(2-methacryloyloxymethyl)cyanaacetamide PA1 M-20 p-(2-acetoacetyl)ethylstyrene PA1 M-21 4-acetoacetyl-1-methacryloylpiperazine PA1 M-22 ethyl-.alpha.-acetoacetoxy methacrylate PA1 M-23 N-butyl-N-acryloyloxyethylacetoacetamide PA1 M-24 p-(2-acetoacetoxy)ethylstyrene PA1 II-1 ethyl acrylate/M-1/acrylic acid copolymer (85/10/5) PA1 II-2 n-butyl acrylate/M-1/methacrylic acid copolymer (85/5/10) PA1 II-3 to 7 n-butyl acrylate/M-1/acrylic acid copolymer (x/y/z) PA1 II-8 n-butyl acrylate/styrene/M-1/methacrylic acid copolymer (65/20/5/10) PA1 II-9 methyl acrylate/M-4/methacrylic acid copolymer (80/15/5) PA1 II-10 n-butyl acrylate/M-5/acrylic acid copolymer (85/10/5) PA1 II-11 n-butyl acrylate/M-7/methacrylic acid copolymer (85/10/5) PA1 II-12 2-ethylhexylacrylate/M-9 copolymer (75/25) PA1 II-13 n-butyl acrylate/M-14/potassium styrenesulfinate copolymer (75/20/5) PA1 II-14 n-hexyl acrylate/methoxyethyl acrylate/M-2 copolymer (70/20/10) PA1 II-15 2-ethylhexyl acrylate/M-15/methacrylic acid copolymer (90/5/5) PA1 II-16 n-butyl acrylate/M-1/M-17/acrylic acid copolymer (75/5/15/5) PA1 1) Example 1 in Japanese Patent Provisional Publication No. H6-332088, and Examples 1 and 2 in Japanese Patent Provisional Publication No. H7-219162; PA1 2) Emulsion containing silver chloride having {100} principal plane, described in PA1 3) Photographic material containing silver bromide iodide, silver bromide or silver bromide chloride having {111} principal plane, described in PA1 4) Monodispersed cubic particles (in which the variation coefficient of the diameter of the circle corresponding to the projected area is preferably in the range of 3 to 40%), described in PA1 a1: [total projected area of {111} tabular particles having an aspect ratio of 2-30/that of all particles].times.100=95%, PA1 a2: [average aspect ratio (mean diameter/mean thickness) of {111} tabular particles having an aspect ratio of 2-30]=12.0, PA1 a3: [mean diameter of the circle corresponding to the projected area of {111} tabular particles having an aspect ratio of 2-30]=1.20 .mu.m, PA1 a4: [mean thickness of {111} tabular particles having an aspect ratio of 2-30]=0.10 .mu.m, and variation coefficient of diameters of the circles corresponding to the projected area=15.5% (which confirms that the sizes of the prepared {111} tabular particles were monodispersed). PA1 Compound A-17 EQU C.sub.12 H.sub.25 O(CH.sub.2 CH.sub.2 O).sub.10 H PA1 Compound A-18 EQU C.sub.12 H.sub.25 O(CH.sub.2 CH.sub.2 O).sub.10 H PA1 a2: [average aspect ratio (mean diameter/mean thickness) of {111} tabular particles having an aspect ratio of 2-30]=8.0, PA1 a3: [mean diameter of the circle corresponding to the projected area of {111} tabular particles having an aspect ratio of 2-30]=0.6 .mu.m, PA1 a4: [mean thickness of {111} tabular particles having an aspect ratio of 2-30]=0.07 .mu.m, and variation coefficient of diameters of the circles corresponding to the projected area=18%. PA1 and water to adjust the volume to 2 litters, and then the value of pH was adjusted to 10.1 with sodium hydroxide. PA1 and water to adjust the volume to 55 mL. PA1 And the value of pH was adjusted to 5.2 with sodium hydroxide, and then water was added to adjust the volume to 1 litter. PA1 were dissolved in distilled water, and then the value of pH was adjusted to 4.5 with sodium hydroxide. Then, distilled water was further added to adjust the volume to 1 litter. The ethylenic unsaturated monomer giving the repeating unit A.sup.1 is a monomer whose homopolymer has a glass transition temperature of not higher than 35.degree. C. Examples of the monomers include an alkylacrylate (e.g., methyl acrylate, ethyl acrylate, n-butyl acrylate, n-hexyl acrylate, benzyl acrylate, 2-ethylhexyl acrylate, and n-dodecyl acrylate), an alkyl methacrylate (e.g., n-butyl methacrylate, n-hexyl methacrylate, 2-ethylhexyl methacrylate, and n-dodecyl methacrylate), a diene (e.g., butadiene and isoprene), and a vinylester (e.g., vinyl acetate and vinyl propionate). The above homopolymer preferably has a glass transition temperature of not higher than 10.degree. C. Examples of the monomers giving such homopolymer include an alkyl acrylate having an alkyl side chain having not less than 2 carbon atoms (e.g., ethyl acrylate, n-butyl acrylate, and 2-ethylhexyl acrylate), an alkyl methacrylate having an alkyl side chain consisting of not less than 6 carbon atoms (e.g., n-hexyl methacrylate, and 2-ethylhexyl methacrylate), and a diene (e.g., butadiene and isoprene). The values of the glass transition temperature are described in "Polymer Handbook, third edition" (by J. Brandrup and E. H. Immergut, pp. V1/209-#V1/277, John Wiley & Sons, 1989). The repeating unit B.sup.1 is different from that of A.sup.1, and is derived from a monomer whose homopolymer has a glass transition temperature of not higher than 35.degree. C. Examples of the monomers include acrylates (e.g., t-butyl acrylate, phenyl acrylate, and 2-naphthyl acrylate), methacrylates (e.g., methyl methacrylate, ethyl methacrylate, 2-hydroxyethyl methacrylate, benzyl methacrylate, 2-hydroxypropyl methacrylate, phenyl methacrylate, cresyl methacrylate, 4-chlorobenzyl methacrylate, and ethyleneglycol dimethacrylate), vinyl esters (e.g., vinyl benzoate and pivaloyloxyethylene), acrylamides (e.g., acrylamide, methylacrylamide, ethylacrylamide, propylacrylamide, butylacrylamide, tert-butylacrylamide, cyclohexylacrylamide, benzylacrylamide, hydroxymethylacrylamide, methoxyethylacrylamide, dimethylaminoethylacrylamide, phenylacrylamide, dimethylacrylamide, diethylacrylamide, .beta.-cyanoethylacrylamide, and diacetoneacrylamide), methacrylamides (e.g., methacrylamide, methylmethacrylamide, ethylmethacrylamide, propylmethacrylamide, butylmethacrylamide, tert-butylmethacrylamide, cyclohexylmethacrylamide, benzylmethacrylamide, hydroxymethylmethacrylamide, methoxyethylmethacrylamide, dimethylaminoethylmethacrylamide, phenylmethacrylamide, dimethylmethacrylamide, diethylmethacrylamide, and .beta.-cyanoethylmethacrylamide), styrenes (e.g., styrene, methylstyrene, dimethylstyrene, trimethylenestyrene, ethylstyrene, isopropylstyrene, chlorostyrene, methoxystyrene, acetoxystyrene, dichlorostyrene, bromochlorostyrene, and methyl vinylbenzoate), divinylbenzene, acrylonitrile, methacrylonitrile, N-vinylpyrrolidone, N-vinyloxazolidone, vinylidene chloride, and phenylvinyl ketone. In order to improve the stability of the latex, the polymer represented by the aforementioned formula (II) may be copolymerized with a monomer having an anionic functional group (e.g., carboxyl group and sulfonic acid group) described in Japanese Patent Publications No. 60-15935, No. 45-3832 and No. 53-28086, and U.S. Pat. No. 3,700,456. Examples of the monomers include acrylic acid, methacrylic acid, itaconic acid, maleic acid, monoalkyl itaconate (e.g., monomethyl itaconate and monoethyl itaconate), monoalkyl maleate (e.g., monomethyl maleate and monoethyl maleate), citraconic acid, styrenesulfonic acid, vinylbenzylsulfonic acid, vinylsulfonic acid, acryloyloxyalkylsulfonic acid (e.g., acryloyloxymethylsulfonic acid, acryloyloxyethylsulfonic acid, and acryloyloxypropylsulfonic acid), methacryloyloxyalkylsulfonic acid (e.g., methacryloyloxymethylsulfonic acid, methacryloyloxyethylsulfonic acid, and methacryloyloxypropylsulfonic acid), acrylamidoalkylsulfonic acid (e.g., 2-acrylamide-2-methylethanesulfonic acid, 2-acrylamide-2-methylpropanesulfonic acid, and 2-acrylamide-2-methylbutanesulfonic acid), methacrylamidoalkylsulfonic acid (e.g., 2-methacrylamide-2-methylethanesulfonic acid, 2-methacrylamide-2- methylpropanesulfonic acid, and 2-methacrylamide-2-methylbutanesulfonic acid). The above acids may be in the form of salts with alkali metals (e.g., Na, K) or an ammonium ion. In the formula (II), each of x, y and z represents a weight percent ratio of each repeating unit component C.sup.1, A.sup.1 or B.sup.1, respectively. The values of x, y and z are in the ranges of 0.5 to 40 (preferably 0.5 to 30, more preferably 1 to 20), 60 to 99.5 (preferably 70 to 99.5, more preferably 75 to 99), and 0 to 50 (preferably 0 to 35, more preferably 0 to 25), respectively. The above-described monomer having an anionic functional group can be freely used, independently of the glass transition temperature of its homopolymer, in accordance with the purpose such as stabilizing the latex. The amount of the monomer preferably used is in the range of 0.5 to 20 wt. %, more preferably 1 to 10 wt. % per the total amount of the polymer. Preferred examples of the polymer latexs represented by the formula (II) are given below. In the following copolymers, the weight percent ratios of the components are shown in the brackets. II-3 x/y/z=95/2/3 PA2 II-4 x/y/z=92/5/3 PA2 II-5 x/y/z=89/8/3 PA2 II-6 x/y/z=81/16/3 PA2 II-7 x/y/z=72/25/3 PA2 Examples 3 and 4 in Japanese Patent Provisional Publication No. H5-204073, PA2 Example 2a in Japanese Patent Provisional Publication No. H6-194768, and PA2 Example 1 in Japanese Patent Provisional Publication No. H6-227431; PA2 Example 1 in Japanese Patent Provisional Publication No. H8-76305, and PA2 Examples A and K in Japanese Patent Provisional Publication No. H8-69069; PA2 Example 1 in Japanese Patent Provisional Publication No. H8-76305. The polymer latex is prepared by the well-known emulsion polymerization process, and the particle size preferably is the range of 0.01 to 0.1 .mu.m. The emulsion polymerization process is preferably performed in the following manner. In water or a mixed solvent consisting of water and an organic solvent (e.g., methanol, ethanol, acetone) compatible with water, the monomer is emulsified and then polymerized by a radical polymerization initiator at a temperature of 30.degree. C. to approx. 100.degree. C., preferably 40.degree. C. to approx. 90.degree. C. The amount of the water-compatible organic solvent generally is in the range of 0 to 100 volume %, preferably 0 to 50 volume % per that of water. In the process, a radical polymerization initiator is used in an amount of 0.05 to 5 wt. % per that of the monomer, and if desired an emulsifier is used in an amount of 0.1 to 10 wt. % per that of the monomer. As the radical polymerization initiator, azobis compounds, peroxides, hydroperoxides, and redox solvents are employable. Examples of them include potassium persulfate, ammonium persulfate, tert-butyl peraquate, benzoyl peroxide, isopropyl carbonate, 2,4-dichlorobenzyl peroxide, methylethyl ketone peroxide, cumene hydroperoxide, dicumyl peroxide, 2,2'-azobisisobutylate, and 2,2'-azobis(2-amidinopropane)hydrochloride. As the emulsifier, water-soluble polymers as well as anionic, cationic, amphoteric and nonionic surface active agents are employed. Examples of them include sodium laurate, sodium dodecylsulfate, sodium 1-octoxycarbonylmethyl-1-octoxycarbonylmethanesulfonate, sodium laurylnaphthalenesulfonate, sodium laurylbenzenesulfonate, sodium laurylphosphate, cetyltrimethylammonium chloride, dodecyltrimethyleneammonium chloride, N-2-ethylhexylpyridinium chloride, polyoxyethylenenoneylphenyl ether, polyoxyethylenesorbitanlauric ester, polyvinyl alcohol, and the emulsifier and water-soluble polymers disclosed in Japanese Patent Publication No. 53-6190. Needless to say, the emulsion polymerization conditions such as polymerization initiator, concentration, reaction temperature and reaction time can be optionally selected in consideration of its purpose. In the polymerization process, the components such as monomer, surface active agent and medium may be beforehand placed in a reaction container and then the initiator may be added to polymerize the monomer, or otherwise all or a portion of each component may be dropwise added to perform the polymerization. The monomer giving the repeating unit C.sup.1 in the formula (II) and the polymer latex having an active methylene group can be prepared in the manners described in U.S. Pat. Nos. 3,459,790, 3,619,195, 3,929482, and 3,700,456; West Germany Patent No. 2,442,165; EP 13,147; and Japanese Patent Provisional Publications No. 50-73625 and No. 50-146331. Preferably, the polymer latex having an active methylene group has the core/shell structure. In the photographic material, the ratio of the polymer latex content in the emulsion layer to that in the photo-insensitive hydrophilic colloid layer (content per gelatin in the emulsion layer/that in the photo-insensitive layer) preferably is not less than 1.0, more preferably in the range of 0 to 0.9. In the case that the photographic material has two or more emulsion layers provided on one side, the above polymer latex content per gelatin is calculated based on the total amounts of gelatin and the polymer latex in the layers. If the photographic material has two or more photo-insensitive layers provided on one side, at least one of them satisfies the above condition for the polymer latex content. The amount of the polymer latex used for the invention is not particularly restricted as long as the above condition is satisfied, but preferably in the range of 10 mg/m.sup.2 to 10 g/m.sup.2, more preferably 100 mg/m.sup.2 to 1.5 g/m.sup.2. In each layer, the polymer latex content per gelatin generally is in the range of 5 to 400 wt. %, preferably 10 to 200 wt. %. In the invention, the photosensitive emulsion layer preferably contains tabular silver halide particles having an aspect ratio of 2 to 30, in an amount of 50 to 100%, more preferably 80 to 100% per the total silver halide particles in terms of the projected area. The average aspect ratio of the tabular silver halide particles preferably is in the range of 3 to 20. The "average aspect ratio" here means a value calculated by dividing a mean diameter of a circle corresponding to a projected area by a mean thickness of particles. Generally. an emulsion containing tabular particles of a large aspect ratio has a low sensitivity for blue light, but has a high sensitivity for green light. Therefore, the tabular particles having a large aspect ratio is preferably used for the invention. The silver halide composition used for the invention is not particularly restricted, and examples of the silver halide composition include silver chloride, silver bromide, silver iodide, silver bromide chloride, silver bromide iodide and silver bromide chloride iodide. Preferred are silver bromide iodide, silver bromide chloride iodide containing 0 to 0.5 mol % of silver iodide, and silver (bromide) chloride (iodide) containing 50 to 100 mol % of silver chloride, because those silver halides give an emulsion having a low sensitivity for blue light. The preferred tabular silver halide particles can be prepared in the known manner. Particularly preferred are the monodispersed hexagonal tabular particles (described in Japanese Patent Provisional Publication No. 63-151618) and the tabular silver halide particles having an average thickness of not more than 1 .mu.m (described in Japanese Patent Provisional Publications No. 62-115435, No. H6-43605, and No. H6-43606). As the surface protective layer provided on the emulsion layer, known material can be used for the invention. In the protective layer, hydrophilic colloid such as gelatin is contained as a binder. As the transparent support, known material such as polyethylene terephthalate (PET) can be used. Preferred examples of the silver halide photographic materials (radiographic films) employed in combination with with the radiographic intensifying screen for the present invention, and preferred examples of the components for the photographic material are as follows: More detailed description of the preferable photographic material and the components thereof is given in Japanese Patent Provisional Publication No. H6-67305. In the invention, it is also preferred to shift the distribution of spectroscopic sensitivity of the photographic material in order to effectively use the luminescence emitted by the fluorescent dye or pigment. The method for shifting is well-known and, for example, additives such as thiocyanates, monomethine dyes, azaindene compounds, imidazole dyes and silver iodide are used. The photographic material and the radiographic intensifying screen may be singly combined to use. However, in many cases, the photographic material having silver halide emulsion layers provided on both sides and a pair of the intensifying screens (which are referred to as "front screen" and "back screen") are used in combination. Examples of the developer preferably used for the development of the photographic material of the invention include polyhydroxybenzenes (such as hydroquinone, which is generally used for the treatment process of conventional medical silver halide photographic materials); and ascorbic acid, erysorbic acid (diastereomer of ascorbic acid), and their alkaline metal salts (e.g., lithium salts, sodium salts, potassium salts). The time for the treatment in dry-to-dry of the photographic material preferably is in the range of 20 to 100 seconds, more preferably 25 to 50 seconds. In the treatment process, each of the developer and the fixer is preferably supplied in an amount of 25 to 200 mL, more preferably 40 to 160 mL, further preferably 60 to 130 mL. In the process for producing the silver halide photographic material, the coated and dried material is wound up into a roll under an absolute humidity of not more than 1.4 wt. %, preferably 0.6 to 1.3 wt. %. Thus wound roll of the material is further processed under an absolute humidity of not more than 1.4 wt. %, preferably 0.6 to 1.3 wt. %. The "absolute humidity (wt. %)" here means a percent ratio of weight (kg) of water vapor in wet air to that (kg) in dry air, and hence it indicates the degree of wet air. The photographic material thus produced is preferably stored in a heat-sealed moisture-proof bag in which the absolute humidity is kept in the range of not more than 1.4 wt. %, preferably 0.6 to 1.1 wt. %. Prior to the storage in the bag, it is particularly preferred for the photographic material to season under an absolute humidity of not more than 1.4 wt. % after the above process. EXAMPLE 1 Production of Radiographic Intensifying Screen A 1) Preparation of Support Having Light-reflecting Layer Containing Titanium Dioxide 500 g of rutile type titanium dioxide powder having a mean grain size of 0.28 .mu.m (CR95 [trade name], available from Ishihara Industries Co., Ltd.) and 100 g of acrylic binder resin (Cryscoat P1018GS [trade name], available from Dainippon Ink & Chemicals, Inc.) were placed in methyl ethyl ketone, and mixed to prepare a coating liquid having a viscosity of 10 PS. The coating liquid was then evenly applied by means of a doctor blade onto a polyethylene terephthalate film (thickness: 250 .mu.m) containing titanium dioxide powder, and then dried to form a light-reflecting layer. The thickness of the dried light-reflecting layer was 40 .mu.m. The volume filling content of titanium dioxide in the light-reflecting layer was 48%, and a diffuse reflectivity at a wavelength of 545 nm (which corresponds to the main peak of the luminescence emitted by terbium activated gadolinium oxysulfide Gd.sub.2 O.sub.2 S:Tb phosphor) was 95.5%. 2) Preparation of a Phosphor Sheet 250 g of terbium activated gadolinium oxysulfide (Gd.sub.2 O.sub.2 S:Tb, mean grain size: 3.5 .mu.m, amount of Tb per 1 mol of Gd: 0.003[0.3 mol %]), 8 g of polyurethane binder resin (Pandex T5265M [trade name], available from Dainippon Ink & Chemicals, Inc.), 2 g of epoxy binder resin (Epicoat 1001 [trade name], available from Yuka Shell Epoxy Kabushiki Kaisha), 10 mg of the fluorescent dye No. 25 (coumarin-6, compound No. 44,263-1, available from Aldrich) and 0.5 g of isocyanate compound (Colonate HX [trade name], available from Nippon Polyurethane Kogyo Kabushiki Kaisha) were placed in methyl ethyl ketone, and mixed by means of a propeller mixer to prepare a coating liquid having a viscosity of 25 PS (at 25.degree. C.). The coating liquid was then applied onto a temporary support (polyethylene terephthalate sheet having a surface beforehand coated with silicon releasing agent), and dried to form a phosphor layer. The formed phosphor layer was then peeled off from the temporary support to prepare a phosphor sheet. 3) Fixing Phosphor Sheet onto Light-reflecting Layer The above-prepared phosphor sheet was placed on the light-reflecting layer prepared in the above 1), and then pressed to fix by means of a calender roll under a pressure of 400 kgw/cm.sup.2 at 80.degree. C. The thickness of the resultant phosphor layer was 100 .mu.m, and the volume filling content of the phosphor was 68%. 4) Preparation of Surface Protective Layer A polyethylene terephthalate (PET) film (thickness: 6 .mu.m) containing anatase type titanium dioxide (A220 [trade name], available from Ishihara Industries Co., Ltd.) in an amount of 3 wt. % was prepared. The scattering length and the haze of the film observed at 545 nm were 30 .mu.m and 50, respectively. The film was laminated on the phosphor layer using a polyester adhesion. Thus, a radiographic intensifying screen (Screen A) comprising a support, a light-reflecting layer, a phosphor layer and a surface protective layer was produced. Production of Radiographic Intensifying Screens B, C and D for Invention The above-mentioned procedures were repeated except for using each of the fluorescent dyes shown in Table 2 in the shown amount in place of 10 mg of the fluorescent dye No. 25, to prepare each of the radiographic intensifying screens (Screens B, C and D) for the invention. Production of Radiographic Intensifying Screens X. Y and Z for Comparison The procedures for the preparation of Screen A were repeated except for using a non-fluorescent yellow dye (oil yellow 3G [trade name], available from Orient Industries Co., Ltd.) in each amount shown in Table 2 in place of 10 mg of the fluorescent dye No. 25, to prepare each of radiographic intensifying screens (Screens X and Y) for comparison. Further, the procedures for the preparation of Screen A were repeated except for using no fluorescent dye to prepare a radiographic intensifying screen (Screen Z) for comparison. Measurement of Emission Spectrum Each screen produced above was excited with X-rays (40 KVp, tungsten target tube), and an emission spectrum was measured by means of a modified fluorophotometer (F4010, manufactured by HITACHI, Ltd.). The results are shown in FIGS. 1 to 3. FIG. 1 shows an emission spectrum of Screen B containing the fluorescent dye according to the invention. FIG. 2 shows an emission spectrum of Screen Z containing no fluorescent dye. FIG. 3 shows an emission spectrum of Screen X containing a yellow absorbing dye for comparison. Comparison between FIG. 1 and FIG. 2 confirms the following fact. Since the fluorescent dye absorbs bright emission lines of Gd.sub.2 O.sub.2 S:Tb phosphor in the wavelength region shorter than 500 nm, the screen B of the invention exhibits a broad emission spectrum having the maximum peak at 520 nm. This means that the light emitted by Screen B of the invention has a relatively small amount of blue component (light in the wavelength region of 380 to 500 nm) and a relatively large amount of green component (light in the wavelength region of 500 to 570 nm). FIG. 3 indicates the following fact. Although the yellow dye contained in Screen X absorbs the blue component of the emitted light to reduce the cross-over, the sensitivity contributed by the blue component is reduced. FIG. 4 shows an emission spectrum of Screen B excited at 417 nm. As shown in FIG. 4, the fluorescent dye No. 25 emits a luminescence having a maximum peak at approx. 520 nm and the half-width of 58 nm. FIG. 5 shows a spectrum indicating spectral sensitivity of the photographic material (sample 1) described below, and indicates that the sample is more sensitive to the green component than the blue component. According to FIG. 5, the luminescence emitted by the fluorescent dye preferably has a maximum peak in the range of 490 to 600 nm. Since a relatively small amount of the blue component is absorbed by the photographic material, the cross-over can be effectively reduced using a combination of the above photographic material and the intensifying screens emitting a relatively small amount of the blue component. Production of Silver Halide Photographic Material (Sample 1) Silver halide photographic material (Sample 1) was produced in the following manner. (Preparation of Emulsion A: {111} tabular particles of AgBr) 6.0 g of KBr and 7.0 g of gelatin (weight average molecular weight: 15,000) were placed in 1 litter of water, and then heated to 55.degree. C. To the prepared dispersion liquid at that temperature, 38 ml of a mixture of 37 mL of aqueous AgNO.sub.3 solution (AgNO.sub.3 : 4.00 g) and 5.9 g of Kr was added with stirring for 37 seconds by a double-jet method. After that, 18.6 g of gelatin was further added. The resultant dispersion liquid was heated to 70.degree. C., and then 89 mL of aqueous AgNO.sub.3 solution (AgNO.sub.3 : 9.80 g) was added for 22 minutes. 7 mL of 25% aqueous ammonia solution was further added, and then the liquid was stored for 10 minutes for physical ripening while the temperature was kept at the same level. After the physical ripening, 6.5 mL of 100% aqueous acetic acid solution was added. Further, 435 mL of aqueous AgNO.sub.3 solution (AgNO.sub.3 : 153 g) and 677 mL of aqueous KBr solution (KBr: 573 g) were added for 37 minutes by a double-jet method while the value of pAg was kept at 8.5. After that, 15 mL of 2 N aqueous potassium thiocyanate solution was added. Thus obtained liquid was then stored for 5 minutes for physical ripening while the temperature was kept at the same level. The resultant liquid was cooled to 35.degree. C. The AgBr particles formed in thus prepared liquid had the following characteristic values on the shape. After the above procedures, soluble salts were removed by a sedimentation method. The liquid was again heated at 40.degree. C., and then 30 g of deionized gelatin treated with alkali, 2.35 g of phenoxyethanol and 0.8 g of sodium polystyrenesulfonate (thickener) were added. After that, the values of pH and pAg were adjusted to 5.90 and 8.00 with sodium hydroxide and silver nitrate, respectively. (Chemical Sensitization) The above-prepared particles were subjected to chemical sensitization with stirring at 56.degree. C. in the following manner. To the above liquid, Thiosulfonate compound-1 (shown below) was added in the amount of 1.times.10.sup.-4 mol per 1 mol of the silver halide, and then AgI fine particles having a mean diameter of 0.10 .mu.m was further added in an amount of 0.1 mol % based on the total amount of silver. After 5 minutes, 1 wt. % KI solution was added in an amount of 1.times.10.sup.-3 mol per 1 mol of the silver halide, and then the liquid was stored for 3 minutes. Then, 1.times.10.sup.-6 mol/mol Ag of thiourea dioxide was added, and the resulting liquid was further stored for 22 minutes for reduction sensitization. After that, 3.times.10.sup.-4 mol/mol Ag of 4-hydroxy-methyl-1,3,3a,7-tetraazaindene and Sensitizing dyes-1, 2 and 3 (shown below) were added in each amount shown below. Further, 1.times.10.sup.-2 mol/mol Ag of calcium chloride, 1.times.10.sup.-5 mol/mol Ag of chloroaurate and 3.0.times.10.sup.-3 mol/mol Ag of potassium thiocyanate were added. Following that, 6.times.10.sup.-6 mol/mol Ag of sodium thiosulfate and 4.times.10.sup.-6 mol/mol Ag of Selenium compound-1 (shown below) were added. After 3 minutes, 0.5 g/mol Ag of nucleic acid was added. The obtained liquid was stored for 40 minutes, and then Water-soluble mercapto compound-1 (shown below) was added. The resulting liquid was cooled to 35.degree. C. to prepare the aimed emulsion. ##STR135## (Preparation of Coating Liquid for Emulsion Layer) Into the emulsion chemically sensitized in the above manner, the following compounds were added in the amounts described below per 1 mol of the silver halide, to prepare a coating liquid for emulsion layer. Gelatin (including that 80 g contained in the emulsion) Dextran (average molecular weight: 39,000) 10.0 g Sodium polyacrylate 5.1 g (average molecular weight: 400,000) Sodium polystyrenesulfonate 1.2 g (average molecular weight: 600,000) Potassium iodide 78 mg Hardening agent: 1,2-bis(vinyl- 4.3 g sulfonylacetoamide) ethane Compound A-1 42.1 mg Compound A-2 10.3 g Compound A-3 0.11 g Compound A-4 8.5 mg Compound A-5 0.43 g Compound A-6 0.04 g Compound A-7 15 g Dye emulsion a (solid content) 0.50 g Dye emulsion m (solid content) 30 mg NaOH (with which pH was adjusted at 6.1). ##STR136## Dye emulsions a and m in the above were prepared in the following manner. (Preparation of Dye Emulsion a) 60 g of Dye-1 (shown below), 62.8 g of 2,4-diamyl-phenol and 62.8 g of dicyclohexylphthalate were dissolved in 333 g of ethyl acetate at 60.degree. C. Then, 65 mL of 5 wt. % sodium dodecylbenzenesulfonate aqueous solution, 94 g of gelatin and 581 mL of water were added and emulsified at 60.degree. C. for 30 minutes using a dissolver. Further, 2 g of p-methyl hydroxybenzoate and 6 L of water were added, and the thus obtained mixture was cooled to 40.degree. C. The mixture was condensed using a ultrafilter (Labo-Module ACP1050 [trade name], available from Asahi Chemical Industry Co., Ltd.) to reduce the amount to 2 kg. Thereafter, 1 g of p-methyl-hydroxybenzoate was added to prepare Dye emulsion a. ##STR137## (Preparation of Dye Emulsion m) 10 g of Dye-2 (shown below) was dissolved in a mixed solvent of 10 mL of tricresyl phosphate and 20 mL of ethyl acetate. Thus prepared solution was added to 100 mL of 15 wt. % aqueous gelatin solution containing 750 mg of Anionic surface active agent-1 (shown below), and emulsified to prepare Dye emulsion m. ##STR138## (Preparation of Coating Liquid for Dye Layer) The coating liquid was prepared so that the resultant dye layer might contain the following compounds in the amounts described below (amount of each component was based on the layer provided on a single face). Gelatin 0.25 g/m.sup.2 Compound A-8 1.4 mg/m.sup.2 Sodium polystyrenesulfonate 5.9 mg/m.sup.2 (average molecular weight: 600,000) Dye dispersion A (used as a dye) shown in Table 1 Compound A-8 in the above is shown below: ##STR139## Dye dispersion A in the above was prepared in the following manner. (Preparation of Dye dispersion A) Each dye shown in Table 1 was not dried, but treated in the form of wet-cake. 2.5 mmol of each dye, 1.2 cc of 25 wt. % aqueous solution of Dispersing aid V (shown below) and water were mixed to prepare 32 g of a slurry. The prepared slurry and 120 g of zirconia beads (mean diameter: 1 mm) were placed in a vessel, and then dispersed for 6 hours using a mixer (1/16G sand grinder mill, available from Eyemex Co., Ltd.). Water was then added so that the content of the dye might be 2 wt. %, to prepare a dye dispersion. Following the above, a photographic gelatin was added into the prepared dispersion so that the solid dye and the photographic gelatin might be incorporated in the same amount (5 wt. %). The antiseptic (Additive D shown below) and distilled water were further added so that the amount of Additive D might be 2,000 ppm per that of gelatin. After that, the resulting emulsion was cooled to store in the form of jelly. Thus, Dye dispersion A containing dispersed solid fine particles was prepared. The mean particle size of the solid fine particles was 0.4 .mu.m (in the case of F-16). ##STR140## (Preparation of Coating Liquid for Surface Protective Layer) The coating liquid was prepared so that the resultant surface protective layer might contain the following compounds in the amounts described below. Gelatin 0.780 g/m.sup.2 Sodium polyacrylate 0.025 g/m.sup.2 (average molecular weight: 400,000) Sodium polystyrenesulfonate 0.0012 g/m.sup.2 (average molecular weight: 600,000) Matting agent 1 0.072 g/m.sup.2 (average particle size: 3.7 .mu.m) Matting agent 2 0.010 g/m.sup.2 (average particle size: 0.7 .mu.m) Compound A-9 0.018 g/m.sup.2 Compound A-10 0.037 g/m.sup.2 Compound A-11 0.0068 g/m.sup.2 Compound A-12 0.0032 g/m.sup.2 Compound A-13 0.0012 g/m.sup.2 Compound A-14 0.0022 g/m.sup.2 Compound A-15 0.030 g/m.sup.2 Proxcell (available from ICI) 0.0010 g/m.sup.2 NaOH (with which pH was adjusted at 6.8) Matting agents 1 and 2, and Compounds A-9 to A-15 in the above are as follows: ##STR141## ##STR142## ##STR143## (Preparation of Support) (1) Formation of First Undercoating Layer The surface of a biaxially stretched polyethylene terephthalate film (thickness: 175 .mu.m) containing Dye-1 (shown above) in an amount of 0.04 wt. % was subjected to corona discharge treatment. Onto one surface of the film, the following coating liquid was applied in the amount of 4.9 cc/m.sup.2 by a wire-converter. The coated liquid was then dried at 185.degree. C. for 1 minute to form a first undercoating layer. Also onto the other surface, a first undercoating layer was formed in the same manner as described above. (Coating liquid for first undercoating layer) Solution of butadiene-styrene copolymer latex 158 mL (solid content: 40%, weight ratio of butadiene/styrene: 31/69) 4 Wt. % 2,4-dichloro-6-hydroxy-s-triazine 41 mL sodium salt solution Distilled water 801 mL In the above latex solution, Compound A-16 (shown below, emulsifier) was added in an amount of 0.4 wt. % per the solid content of the latex. ##STR144## (2) Formation of Second Undercoating Layer Onto each of the first undercoating layers formed on both surfaces of the support, the coating liquid having the following components in the amounts described below was applied (amount of each component was based on the liquid for coating a single layer) by a wire-bar coater, and then dried at 155.degree. C. to form a second undercoating layers. The amount of the liquid for coating a single layer was 7.9 cc/m.sup.2. Gelatin 0.06 g/m.sup.2 Dye dispersion B (used as a dye shown in Table 1 for a single layer) Compound A-17 1.8 g/m.sup.2 Compound A-18 0.27 g/m.sup.2 Matting agent (polymethyl 2.5 g/m.sup.2 methacrylate having the mean particle size of 2.5 .mu.m) Dye dispersion B in the above was prepared in the same manner as Dye dispersion A. Compounds A-17 and A-18 are shown below: ##STR145## (Preparation of Photographic Material) Each surface of the support prepared above was simultaneously and superposingly coated with the above prepared coating liquids by an extruding method, and then the coated liquids were dried to prepare a photographic material (Sample 1) comprising a support, a dye layer, an emulsion layer and a surface protective layer overlaid in order. The amount of silver provided on a single surface was 1.5 g/m.sup.2. (Measurement of Swelling Ratio) After storing at 40.degree. C. (60%RH) for 7 days, the above photographic material (Sample 1) was soaked in distilled water at 21.degree. C. for 3 minutes and then frozen using liquid nitrogen. The frozen material was cut perpendicularly to the surface by means of a microtome, and then freezedried at -90.degree. C. The section of the thus treated material was observed by means of a scanning electron microscope, to measure the thickness (Tw) of the swelled sample. Independently, the section of the dry sample was also observed by means of a scanning electron microscope, to measure the thickness (Td). Based on thus obtained Tw and Td, the swelling ratio was calculated in accordance with the following formula: EQU Swelling ratio (%)={(Tw-Td)/Td}.times.100. The values of Sample 1 were as follows: Td=2.7 .mu.m, Tw=7.0 .mu.m, and the swelling ratio=159%. Measurement of Absorption Spectrum of Dye in Dye Layer and/or Undercoating Layer In order to measure the absorption spectrum of the dye contained in the dye layer and/or the undercoating layers of Sample 1, the above procedure was repeated except for providing no emulsion layers to produce a sample comprising the same layers as those of Sample 1 except for the emulsion layers. The absorption spectrum of thus produced sample was measured by means of a normal automatic recording spectrophotometer equipped with an integrating sphere. In the measurement, a polyethylene terephthalate film was used as a reference. The results (wavelength of the maximum peak, the ratio of the absorption coefficient at 550 nm to that at 450 nm) are set forth in Table 1. TABLE 1 sample dye layer u.c. layer.sup.1) peak No. dye amount.sup.3) dye amount.sup.3) (nm) ratio.sup.2) 1 F-16 24 mg/m.sup.2 -- 555 5.3 2 F-16 34 mg/m.sup.2 -- 555 5.3 3 F-16 46 mg/m.sup.2 -- 555 5.3 4 F-16 14 mg/m.sup.2 -- 555 5.3 5 F-16 8 mg/m.sup.2 -- 555 5.3 6 -- F-3 14 mg/m.sup.2 550 4.5 7 F-3 20 mg/m.sup.2 F-3 20 mg/m.sup.2 550 4.5 8 F-19 50 mg/m.sup.2 -- 530 1.5 9 -- F-19 20 mg/m.sup.2 530 1.5 Remarks: .sup.1) "u.c. layer" stands for the undercoating layer. .sup.2) "ratio" stands for the ratio of the absorption coefficient at 550 nm to that at 450 nm. .sup.3) The amount in the above is that of the dye included in the layer provided on a single face of the support. [Evaluation 1 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] The sample of silver halide photographic material (Sample 1) produced in the above described manner was evaluated in combination with each of Screens A to D, and X to Z. (Exposure to X-rays) The photographic material was placed in contact with the surface of the screen, and then an X-ray generator (X-ray tube), a photographic material and an intensifying screen were arranged in order. In this arrangement, the photographic material was stepwise exposed to X-rays at different doses (which are adjusted by varying the distance between the screen and the X-ray tube). The dose was varied by the step width of logE=0.15. The used X-ray tube was DRX-3724HD [trade name], available from Toshiba Corporation, in which X-rays were generated by tungsten target and a pulse generator (80 KVp, three-phase), and then passed through 3 mm thick aluminum equivalent material including aperture to make the focal spot size of 0.6 mm.times.0.6 mm. The X-rays generated from the tube were made to pass through a filter of water having the pass of 7 cm (which absorbs X-rays in nearly the same amount as a human body). (Development of Photographic Material) The exposed photographic material was developed in a developing agent (CED2 [trade name], available from Fuji Photo Film Co., Ltd.) and fixed in a fixing agent (CEF2 [trade name], available from Fuji Photo Film Co., Ltd.) by means of a roller conveyor automatic developing machine (CEPROS-M2 [trade name], available from Fuji Photo Film Co., Ltd.) for 45 seconds (dry-to-dry: SP mode). Each of the developing agent and the fixing agent was supplied in an amount of 10 cc per a 10.times.12 inch size. 1) The developed material was divided into two sheets. From one sheet, the emulsion layer not having been in contact with the intensifying screen was removed by enzyme treatment. The optical density of the emulsion layer remaining on the support was measured to determine the sensitivity of the intensifying screen. The results are expressed in relative values in which the sensitivity of screen Z is set at 100. 2) From another sheet, the emulsion layer having been in contact with the intensifying screen was removed. The optical density of the emulsion layer remaining on the support was measured to determine the sensitivity, and then the difference between the determined sensitivity and that determined in the above 1) was calculated. Based on the obtained difference, the degree of cross-over is determined in accordance with the aforementioned equation. The results are set forth in Table 2. TABLE 2 amount of dye cross-over screen (mg/phosphor 1 kg) sensitivity (%) (Examples) A f.dye.sup.1) No.25 40 mg 105 7.5 B f.dye.sup.1) No.25 120 mg 108 7 C f.dye.sup.1) No.25 15 mg 100 8.5 D f.dye.sup.1) No.23 60 mg 105 8 (Comparison Examples) X yellow dye 40 mg 95 10.5 Y yellow dye 120 mg 85 8 Z -- 100 12 Remark: .sup.1) "f.dye" stands for fluorescent dye. The results in Table 2 indicate that each of the intensifying screens according to the invention (Screens A, B, C and D) has a high sensitivity and gives a low cross-over (consequently, gives an image of high sharpness). Particularly, Screen B exhibited the highest sensitivity and the lowest cross-over. In contrast, the intensifying screens for comparison (screens containing a conventional absorbing dye) exhibited a low sensitivity although they reduced cross-over. The results shown in Table 2 are expected before on the basis of the emission spectra shown in FIGS. 1-3. EXAMPLE 2 Production of Silver Halide Photographic Materials (Samples 2-5) The procedures of Example 1 were repeated except that the dye was incorporated into the dye layer in the amount shown in Table 1, to produce each of the silver halide photographic materials (Samples 2 to 5). [Evaluation 2 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] With respect to each combination shown in Table 3, the sensitivity and the cross-over were measured in the same manner as described in Evaluation 1. The results are set forth in Table 3. TABLE 3 material screen sensitivity cross-over (%) (Examples) Sample 1 B 108 7 Sample 2 B 108 5 Sample 3 B 108 2 Sample 4 B 108 11 Sample 5 B 108 17 (Comparison Examples) Sample 1 Z 100 12 Sample 2 Z 100 10 Sample 3 Z 100 7 Sample 4 Z 100 14 Sample 5 Z 100 18 The above results indicate that each combination of the invention (which comprises the photographic material and the intensifying screen containing a fluorescent dye) effectively reduces the cross-over without lowering the sensitivity. Table 3 confirms that the intensifying screen is advantageously combined with the photographic material giving a cross-over in a degree of not less than 10%. EXAMPLE 3 Production of Silver Halide Photographic Materials (Samples 6-9) The procedures of Example 1 were repeated except that the dye was incorporated into the dye layer and/or the undercoating layer in the amounts shown in Table 1, to produce each of the silver halide photographic materials (Samples 6 to 9). [Evaluation 3 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] With respect to each combination shown in Table 4, the sensitivity and the cross-over were measured in the same manner as described in Evaluation 1. The results are set forth in Table 4. TABLE 4 material screen sensitivity cross-over (%) (Examples) Sample 6 D 105 10 Sample 7 D 105 3 Sample 8 D 105 5 Sample 9 D 105 12 (Comparison Examples) Sample 6 X 100 14 Sample 7 X 100 9 Sample 8 X 100 7 Sample 9 X 100 14 The above results indicate that each combination of the invention effectively reduces the cross-over without lowering the sensitivity. In particular, according to Tables 1 and 4, the photographic material having a large ratio of the absorption coefficient at 550 nm to that at 450 nm remarkably reduces the cross-over. EXAMPLE 4 Production of Radiographic Intensifying Screens E to I (1) The procedures of Example 1 were repeated except for the following. In the procedure, the amount of the coating liquid for phosphor layer was changed so that the layer might have a thickness of 80 .mu.m after calender treatment. Thus, the screen containing the fluorescent dye was prepared (Screen E). (2) The procedures of Example 1 were repeated except for the following. In the procedure, 50 g of the phosphor particles having an average particle size of 2.0 .mu.m and 200 g of those having an average particle size of 6.2 .mu.m were incorporated (chemical composition of the phosphor was not changed), and the amount of the coating liquid for phosphor layer was changed so that the layer might have a thickness of 120 .mu.m after calender treatment. Thus, the screen according to the invention was prepared (Screen F). The volume filling content of the phosphor in the phosphor layer was 72%. (3) The procedures of the above (2) were repeated except for the following. In the procedure, the amount of the coating liquid for phosphor layer was changed so that the layer might have a thickness of 95 .mu.m after calender treatment. Thus, the screen according to the invention was prepared (Screen G). (4) The procedures of Example 1 were repeated except for the following. The phosphor layer was formed by superposingly providing two component layers in which one component layer (lower component layer) contained the phosphor particles (chemical composition of the phosphor was not changed) having the average particle size of 3.0 .mu.m (variation coefficient: 45%) and the other component layer (upper component layer) contained those having an average particle size of 6.2 .mu.m (variation coefficient: 30%). The lower and the upper component layers had thicknesses of 80 .mu.m and 100 .mu.m, respectively, after calender treatment. Thus, the intensifying screen according to the invention was prepared (Screen H). The volume filling content of the phosphor in the phosphor layer was 70%. (5) The procedures of Example 1 were repeated except for the following. The phosphor layer was formed by superposingly providing two component layers in which one component layer (lower component layer) contained the phosphor particles (chemical composition of the phosphor was not changed) having an average particle size of 3.0 .mu.m (variation coefficient: 40%) and the other component layer (upper component layer) contained those having an average particle size of 6.2 .mu.m (variation coefficient: 30%). The lower and the upper component layers had thicknesses of 80 .mu.m and 240 .mu.m, respectively, after calender treatment. Thus, the intensifying screen according to the invention was prepared (Screen I). Production of Silver Halide Photographic Material (Sample 10) The procedures for preparing Emulsion A were repeated except for changing the conditions (such as the temperature for forming the particles and the conditions for chemical sensitization) to prepare a monodispersed {111} tabular silver iodobromide emulsion (Emulsion B). The particles of the silver iodobromide had the following characteristic values on the shape. The procedures of Example 2 were repeated except for the following. In the procedure, the dye F-16 was incorporated into the dye layer in an amount of 50 mg/m.sup.2 (based on the layer provided on a single face) and the emulsion layer was formed by superposingly applying Emulsions A and B. Thus prepared photographic material (Sample 10) exhibits almost the same sensitivity and contrast as those of a commercially available radiographic film (UR2 [trade name], available from Fuji Photo Film Co., Ltd.). Production of Radiographic Intensifying Screens P to T for Comparison Each procedure of (1) to (5) in Example 4 was repeated except for not adding the fluorescent dye No. 25, to produce each of the screens for comparison (Screens P, Q. R, S and T). [Evaluation 4 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] A pair of intensifying screens shown in Table 5 and Sample 10 were combined to give a combination so that each of the screens might be in contact with each surface of Sample 10. Thus prepared combinations are shown in Table 5. In Table 5, the intensifying screen placed on the side near the X-ray generator is referred to as "front screen" and that on the opposite side is referred to as "back screen". Each combination was exposed to X-ray, and then developed to measure the sensitivity in the manner described in Evaluation 1. The results are expressed in relative values in which the sensitivity of the pair of Screens P and Q is set at 100. In order to determine the sharpness, a radiographic image of a rectangular chart for measuring MFT (made of molybdenum, thickness: 80 .mu.m, spatial frequency 0-10 1p/mm) was obtained. The chart was placed in front of an X-ray tube at distance of 2 m. The dose of X-rays in the exposure process was determined so that the average difference between the highest and the lowest density of the developed image might be 1.0. The density of the radiographic image of the developed sample was measured to obtain a density profile by means of a microdensitometer under the condition that the aperture was a slit of 30 .mu.m.times.500 .mu.m (scanning direction.times.vertical direction) and the sampling distance was 30 .mu.m. This procedure was repeated twenty times and the obtained values were averaged to obtain a density profile on which CTF was calculated. Thereafter, the peak corresponding to the pulse of each spatial frequency in the density profile was observed to calculate a density contrast of each frequency. The calculated density contrast was then normalized with the density difference at a frequency of 0 to obtain the value of CTF (2 1p/mm) indicating the degree of sharpness. The obtained values concerning sensitivity and sharpness (CTF) were set forth in Table 5. TABLE 5 front back sharpness material screen screen sensitivity (CTF) (Examples) Sample 10 E F 104 0.610 Sample 10 G H 140 0.53 Sample 10 G I 210 0.37 (Comparison Examples) Sample 10 P Q 100 0.585 Sample 10 R S 130 0.510 Sample 10 R T 190 0.36 UR-2 HGM2 HGM2 100 0.500 The results shown in Table 5 indicate that the combination of the intensifying screen according to the invention (i.e., each of Screens E-I) and Sample 10 (a photographic material according to the invention) has a high sensitivity and gives an image of high sharpness. EXAMPLE 5 Production of Silver Halide Photographic Material (Samples 11 and 12) The procedures for producing Sample 10 in Example 4 were repeated except for changing the sensitivity of emulsion, to prepare two photographic materials (Samples 11 and 12) according to the invention. Samples 11 and 12 had the same structure, sensitivity and contrast as those of UR-1 and UR-3 ([trade name], available from Fuji Photo Film Co., Ltd.), respectively. [Evaluation 5 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] Each of the above produced photographic materials (Samples 11 and 12) was evaluated in combination with a pair of screens in the same manner as described in Evaluation 4, to obtain satisfying results like those shown in Table 5. [Evaluation 6 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] The photographic materials produced in Examples 4 and 5 were developed by means of an automatic developing machine CEPROS-30, CEPROS-S or CEPROS-P in place of CEPROS-M2 ([trade names], available from Fuji Photo Film Co., Ltd.) for 30, 60 or 120 seconds (dry-to-dry), to obtain images of high quality exhibiting no residual color. In this developing process, the following treatment solutions were used. (Preparation of Developing Solution) The developing solution consisting of the following components was prepared. The components were as follows: Ethylenediaminepentaacetic acid 8.0 g Sodium sulfite 20.0 g Sodium carbonate monohydrate 52.0 g Potassium carbonate 55.0 g Sodium erysorbic acid 60.0 g 4-Hydroxymethyl-4-methyl-1-phenyl-3-pyrazolidone 13.2 g 3,3'-Diphenyl-3,3'-dithiopropionic acid 1.44 g Diethylene glycol 50.0 g ##STR146## (Preparation of Supplementary Developing Solution) The above developing solution was also used as a supplementary developing solution. (Preparation of Developing Mother Solution) Into 2 litters of the above developing solution, the following starter was added in an amount of 55 mL per 1 litter of the developing solution, to prepare developing mother solution (pH: 9.5). (Preparation of Starter) The starter consisting of the following components was prepared. The components were as follows: Potassium bromide 11.1 g Acetic acid 10.8 g (Preparation of Condensed Fixing Solution) The condensed fixing solution consisting of the following components was prepared. The components were as follows: Water 0.5 litter Ethylenediaminetetraacetic acid dihydrate 0.05 g Sodium thiosulfate 200 g Sodium bisulfite 98.0 g Sodium hydroxide 2.9 g (Preparation of Supplementary Fixing Solution) The above condensed fixing solution was diluted twice with first drained wash-water, to prepare a supplementary fixing solution. (Preparation of Fixing Mother Solution) Two litters of the above condensed fixing solution was diluted with water to adjust the volume to 4 litters. The value of pH was 5.4. (Preparation of Supplementary Wash-Water) The supplementary wash-water was prepared in the following manner. Glutaric aldehyde 0.3 g, and Diethylenetriaminepentaacetic acid 0.5 g EXAMPLE 6 Production of Radiographic Intensifying Screens J to M The procedures of Example 1 were repeated except for using each of the phosphors shown in Table 6 in the corresponding amount, to produce a radiographic intensifying screens (Screen J to M) according to the invention. [Evaluation 7 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] With respect to the sensitivity and the cross-over, the combination of each intensifying screen shown in Table 6 and the photographic material (Sample 2) was evaluated in the same manner as described in Evaluation 1. The results are set forth in Table 6. Table 6 shows a maximum wavelength and a half-width of the luminescence emitted by the fluorescence dye contained in each intensifying screen excited at 417 nm. TABLE 6 amount of dye w.l..sup.2) h.w..sup.3) c.o..sup.5) screen.sup.1) (mg/phosphor 1 kg) (nm) sens..sup.4) (%) (Examples) J f.dye.sup.6) No.51 120 mg 540 75 101 6 K f.dye.sup.6) No.47 120 mg 490 82 95 8 L f.dye.sup.6) No.23 120 mg 513 63 106 6 M f.dye.sup.6) No.48 120 mg 524 83 103 6 B f.dye.sup.6) No.25 120 mg 520 58 108 5 (Comparison Examples) Z -- -- -- 100 10 Y yellow dye 120 mg -- -- 85 6 Remarks: .sup.1) Every screen was combined with Sample 2. .sup.2) "w.l." stands for the maximum wavelength. .sup.3) "h.w." stands for the half-width. .sup.4) "sens." stands for the sensitivity. .sup.5) "c.o." stands for the cross-over. .sup.6) "f.dye" stands for fluorescent dye. The results in Table 6 indicate that the combination of the intensifying screen according to the invention (i.e., each of Screens B and J-M) and Sample 2 (a photographic material according to the invention) has a high sensitivity and gives an image of high sharpness. According to Table 6, a particularly preferred image of a low cross-over can be given with a high sensitivity by the intensifying screen containing the fluorescent dye which emits a luminescence having the maximum peak in the wavelength region of 500 to 555 nm. If the maximum peak shifts to a shorter wavelength, the luminescence emitted by the fluorescent dye does not effectively enhance the sensitivity. This fact can be expected on the basis of the spectral sensitivity shown in FIG. 5. In contrast, if the fluorescent dye emits a luminescence having the peak at a longer wavelength, its excitation spectrum (absorption spectrum) accordingly shifts to a longer wavelength, and hence the dye absorbs an effective portion of the emission (maximum peak: 545 nm) of Gd.sub.2 O.sub.2 S:Tb phosphor. Consequently, the intensifying screen containing such fluorescent dye (e.g., Screen J) is liable to lower in the sensitivity. However, if the absorption spectrum does not shift (in other words, if proper Stokes' shift is ensured), the peak of luminescence emitted by the fluorescent dye preferably is close to 545 nm where the photographic material has a relatively high spectral sensitivity. The results in Table 6 also indicate that the fluorescent dye emitting luminescence of a narrow half-width gives a high sensitivity. FIG. 6 shows the excitation spectrum of Screen B (i.e., the absorption spectrum of the fluorescent dye in Screen B) monitored at 520 nm. According to FIG. 6, the fluorescent dye contained in Screen B (i.e, coumarin-6) well absorbs a light in the wavelength region of 380 to 500 nm while it does not at a wavelength near 545 nm. Since Screen B exhibits excellent properties, the fluorescent dye like coumarin-6 (which absorbs light in 380 to 500 nm and then emits luminescence of narrow halfwidth at near 545 nm) is particularly preferred. EXAMPLE 7 Production of Radiographic Intensifying Screens B2 and 3 The procedures for producing Screen B in Example 1 were repeated except for using Gd.sub.2 O.sub.2 S:Tb phosphor containing the amount of Tb as shown in Table 7, to produce each radiographic intensifying screen according to the invention (Screens B2 and B3). Production of Radiographic Intensifying Screens Z2 and Z3 for Comparison The procedures for producing Screen Z in Example 1 were repeated except for using Gd.sub.2 O.sub.2 S:Tb phosphor including the amount of Tb as shown in Table 7, to produce each radiographic intensifying screen for comparison (Screens Z2 and Z3). [Evaluation 8 of Combination of Radiographic Intensifying Screen and Silver Halide Photographic Material] With respect to the sensitivity and the cross-over, the combination of each screen shown in Table 7 and the photographic material (Sample 2) was evaluated in the same manner as described in Evaluation 1. The results are set forth in Table 7. TABLE 7 amount of Tb amount of dye c.o..sup.3) screen.sup.1) (mol/1 mol Gd) (mg/phosphor 1 kg) sens..sup.2) (%) (Examples) B 0.003 f.dye.sup.4) No.25 120 mg 108 5 B2 0.0015 f.dye.sup.4) No.25 120 mg 105 6 B3 0.010 f.dye.sup.4) No.25 120 mg 103 5 (Comparison Examples) Z 0.003 -- 100 10 Z2 0.0015 -- 92 13 Z3 0.010 -- 98 9 Remarks: .sup.1) Every screen was combined with Sample 2. .sup.2) "sens." stands for the sensitivity. .sup.3) "c.o." stands for the cross-over. .sup.4) "f.dye" stands for fluorescent dye. The results in Table 7 indicate that the fluorescent dye according to the invention is effective even if the amount of the contained in Gd.sub.2 O.sub.2 S:Tb phosphor is varied. Further, according to Table 7, the fluorescent dye is particularly effective when the amount of Tb is in the range of 0.001 to 0.02. EXAMPLE 8 The procedures of Example 1 were repeated except that a phosphor layer-forming coating dispersion was prepared by placing 250 g of Gd.sub.2 O.sub.2 S:Tb phosphor, 6 g of a polyurethane binder resin (Pandex T5265M, trade name), 1 g of an epoxy binder resin (Epikote 1001), 10 mg of Fluorescent No. 25, and 0.25 g of an isocyanate compound (Colonate HX) in methyl ethyl ketone. Thus, Radiographic Intensifying Screen A2 was prepared. Intensifying Screen A2 was evaluated in its radiographic characteristics in combination with Photographic Material (Sample 1). The results indicate that Intensifying Screen A2 has almost the same favorable characteristics as those of Intensifying Screen A. EXAMPLE 9 The procedures of Example 4 for the preparation of Radiographic Intensifying Screens E, F, G, H and I were repeated except for decreasing the amount of binder to the same level of Example 8, to prepare Radiographic Intensifying Screens E2, F2, G2, H2, and I2. These Intensifying Screens were evaluated in its radiographic characteristics in combination with Photographic Material (Sample 10). The results indicate that these Intensifying Screens have almost the same favorable characteristics as those of Intensifying Screens of Example 4.
	summary
	abstract	The known fully ceramic microencapsulated fuel (FCM) entrains fission products within a primary encapsulation that is the consolidated within a secondary ultra-high-temperature-ceramic of Silicon Carbide (SiC). In this way the potential for fission product release to the environment is significantly limited. In order to extend the performance of this fuel to higher temperature and more aggressive coolant environments, such as the hot-hydrogen of proposed nuclear rockets, a zirconium carbide matrix version of the FCM fuel has been invented. In addition to the novel nature to this very high temperature fuel, the ability to form these fragile TRISO microencapsulations within fully dense ZrC represent a significant achievement.
039830500	summary	BACKGROUND OF THE INVENTION This invention relates to the storage of high-level wastes for an extended period of time. More specifically, this invention relates to an improvement in the method for the storage of high-level radioactive wastes as solids in metal canisters by making the canisters self-sealing. A problem facing the nuclear industry which has received much attention is how to dispose of radioactive wastes so that they will never contaminate the biosphere with radioactivity. While disposal of these wastes in a form and in an environment in which no contamination of the biosphere is possible under any conceivable circumstances for the entire period that radioactivity is at a dangerous level is the ultimate objective of waste management engineers, no such disposal procedure has as yet gained wide acceptance. As an alternative or supplement to ultimate disposal, an engineered storage employing buildings, vaults, tanks, etc. which will require continuous surveillance and maintenance for up to 100 years may be employed. A number of factors must be considered in the development of these storage methods which are to last for 100 years. For example, integrity of the storage container, cost of solidifying the wastes, retrievability of the wastes if necessary for reprocessing and cost of storage must all be considered. Large volumes of liquid waste have been and are being stored in large tanks. Due to the tremendous cost of storing liquids, considerable effort has gone into programs for developing methods for solidifying these wastes for longterm storage. For example, the use of cement for the fixation of radioactive wastes has been studied extensively, and from a production standpoint, is probably the most convenient way in which to solidify waste. All that is required is that the material to be solidified be mixed with water and the mixture cast into the desired monolithic shape and allowed to set. However, some radioactive salts are readily leachable from concrete and the retrievability of the radioactive material from the concrete may pose some future problems. Studies are also being conducted into methods for solidifying the liquid wastes by calcining them in a fluidized-bed or batch-type pot calciner or transforming the wastes into glass. In either case, the calcine or glass would then be sealed into metal canisters for storage. Because of the heat developed by the radioactive waste, the canisters must be cooled, either by air or by storage in a water-filled tank. In either environment, the canister is subject to failure of its integrity which may result in the release of radioactive material to the environment. SUMMARY OF THE INVENTION A method has been developed which eliminates many of the problems attendent upon the storage of radioactive calcine wastes in a metal canister for long periods of time in a moisture-containing environment such as open-air or water-tank storage. In the method of the present invention, dry cement powder is added to the metal canister containing the calcined wastes so that the cement powder is in contact with the inner surface of the wall of the canister before the canister is sealed whereby, should the canister wall fail and develop an opening to the environment, moisture from the environment will enter the canister, mix with the Portland cement, forming concrete which will harden, seal the opening and prevent the escape of any radioactivity from the canister into the environment. It is therefore the object of the invention to provide an improved method for the safe storage of calcined high-level radioactive waste. It is a further object of the invention to provide an improved method for the safe storage of calcined high-level radioactive wastes in metal canisters for long-term storage in a moisture-containing environment. It is a further object of the invention to provide a method for providing a self-sealing metal canister for the long-term storage of calcined high-level radioactive wastes. BRIEF DESCRIPTION OF THE PREFERRED EMBODIMENT These and other objects of the invention may be met by mixing cement powder with the solid calcined high-level radioactive waste in a ratio of from about 1 part cement powder to from about 3 to about 10 parts by weight of calcine before the calcine is placed in the metal canister and sealed for storage in a moisture-containing environment. It has been assumed that the wastes to be stored will be stable, dry, solid, high-level radioactive wastes. The reference design waste form is a calcine product from Light Water Reactor Uranium (PWR-2) fuels, typically packaged in a stainless steel canister, 12.75 inches outside diameter by 10 feet long, containing 6.3 cubic feet of waste from reprocessing 3.15 metric tons of fuel and generating approximately 3.25 kilowatts of decay heat at 10 years age. The ratio of cement to calcine will depend upon the physical form of the calcine, i.e., whether it is a fine granular friable material such as sand or of a coarser form such as gravel. In general, a fine granular powder is limited to a ratio of 3 to 1 while a coarser granular form may vary up to a ratio of 10 parts to 1 part cement by weight. The type of cement may be any of the ordinary Portland cements which have the ability to harden in a high-moisture environment such as where the canister has been placed in a water tank for storage. It may be desirable to use an expansive cement such as Atlas MX (Universal Atlas Co. Div. of U.S. Steel) which will expand about 2 percent as it sets. Because of the high temperatures which the radioactive material in storage may reach, a cement having a high alumina content may also be considered. The cement may be mixed with the calcine in various ways. For example, the cement powder can be added to the calciner to be mixed with the calcine as it is formed. This would enhance the incorporation of the oxide waste with the cement powder. The cement can also be mixed with the calcine as it is fed into the canister. This would give an even distribution of cement with the calcine within the canister. As another alternative, the cement powder can be inserted into the canister as an outer annulus, with the calcined waste in the center of the canister. This may be accomplished, for example, by inserting a cylindrical sleeve into the canister which is slightly smaller in diameter than the inner wall of the canister so that an annular space about 1 inch deep is formed between the sleeve and the wall, and filling this space with cement powder while filling the center of the sleeve with calcine. The sleeve can then be removed and the canister sealed for storage. This method has the advantage that the amount of cement necessary to provide the self-sealing feature for the canister would be minimized. It is important that the inner surface of the wall of the canister be in contact with the cement powder so that a breach anywhere in the canister wall which would permit the entrance of moisture either from the cooling basin or from the air will result in the water mixing with and wetting the cement so that it can set, plug the breach and thus prevent the inflow of any more water. The formation of cement under conditions of leak will, of course, not provide the usual proportioned well-mixed cement-water paste. However, a leak due to corrosion or to a crack in the canister will first open only a very small surface for the entry of water. Since the canister is sealed and the water is only at the pressure of the gravity head in the pool, the water will enter slowly. The temperature is favorable for quick setting of cement, since the temperature of the outer wall of the canister is expected to be about 128.degree.F. Thus, while the point adjacent to the breach in the canister wall may have too much water to form a good cement, the inside portions of cement will not have excess water and a zone of setting proportions will encircle the hole. The leaching of the calcine will be reduced and the integrity of the calcine and of the container will be maintained for a longer time, so that maintenance or removal of the canister can be carried out if desired. ADVANTAGES OF THIS METHOD a. The properties and further accessibility of the dry calcine are not changed in the normal situation where the canister does not leak. b. The amount of concrete formation, which may change the accessibility of the calcine, is limited to the amount of water leaked. c. The formation of monolithic concrete involves the doubling of the volume of solid previously present as dry calcine, thus closing any void space in the calcine and forming a seal against any further water entry or exit of calcine or water. d. This method is compatible with the formation of a concrete monolith of the calcine prior to canistering and thus should be deemed desirable for this type of interim storage of high-level wastes. e. No expensive reagents and a minimum of equipment are required to carry out this method of the interim storage of high-level radioactive wastes.
042232295	summary	BACKGROUND OF THE INVENTION The use of X-rays for diagnostic purposes, and radiotherapy, expecially as applicable for treatment of malignancies, has been found necessary and desireable in many phases of dentistry and medicine. Unnecessary and uncontrolled subjection of the human body, or portions thereof, to radiation, however, including X-rays for diagnostic purposes, or radiation exposure connected with radiotherapy treatment of malignancies, are increasingly recognized as being harmful. The medical and dental professions accordingly have taken steps attempting to reduce, in so far as possible, the subjection of patients to such harmful radiation, occuring either inadvertently or during intended treatment but perhaps to specifically different parts of the body, or resulting from stray, scattered and surplus rays. Recent studies have shown, for example, that oral complications occur in patients undergoing radiotherapy treatments for malignancies, even though the malignancies were not in the head and neck. An article in The Journal of the American Dental Association, September 1978, Volume 97, No. 3, pps. 468-472, by Stephen T. Sonis, Andrew L. Sonis and Alan Lieberman, entitled "Oral Complications in Patients Receiving Treatment for Malignancies Other Than of The Head and Neck" discusses the results of a recent study in this connection and reference is made therein to several additional publications & articles reporting the results of other studies. The following ones of the articles are pertinent in regard to the background of the present invention: Ref. 1. DelRegato, J.A. Dental lesions observed after roentgen therapy in cancer of the buccal cavity, pharynx, and larynx. American Journal of Roentgenology Radium Therapy and Nuclear Medicine, 42:404, Sept. 1939. Ref. 2 King, E.R., Elzay, R.P. and Dettman, P.M., Effects of ionizing radiation in the human oral cavity and oropharynx, results of a survey. Radiology, 91:990, November 1968. The use of X-rays for diagnostic purposes in dentistry has also recently been of some concern. Research has generally led the medical and dental professions to avoid unnecessary exposure of patients to X-rays. It has been found that even very limited amounts of exposure to radiation, especially in children, sometimes causes damage to such glands as the pituitary and thyroid. In efforts to avoid such problems, techniques and apparatus have been developed attempting to either absorb or otherwise shield body areas of a patient from undesired exposure or from stray, or scattered x-rays such as those which normally tend to scatter from the principal stream of X-rays. Preferably the only X-rays allowed to contact human tissue are those necessary in the procedure. Some apparatus of this type is disclosed, for example, in United States Patents Re. 25,773; 3,304,422 and 3,304,423. These patents disclose dental X-ray shields and X-ray aiming means, in conjunction with X-ray film holding devices used in the taking of X-rays of teeth. The stated purpose, and suggested result, is to greatly minimize possible injury to the patient from scattered and surplus X-rays contacting tissues other than those which are intended to be subjected to X-rays. It is also known to use protective garments or covers, in the nature of aprons and the like, to shield patients and/or certain body areas from stray X-rays during the course of X-ray examination or treatment. An example of an apron type of protective shield is shown in U.S. Pat. No. 3,233,248. Generally, however, such shields and/or protective covers are very heavy and uncomfortable, and varied sizes are required for efficient use with different individual patients. Another example of a protective shield is shown in U.S. Pat. No. 3,569,713. This patent discloses a shield used in dentistry which is adapted for positionment on, and partially around, the neck of a patient, and is a shield for the thyroid gland of the patient. There is additionally an increasing awareness in the medical and dental professions of possible serious damages which can be inflicted on teeth, gingiva, dentition, periodontal bone, parotid, sublingual, and salivary glands, and other related near body areas, by harmful side effects of radiation therapy, particularly when the patient receives large radiation doses in the head and neck regions for example. Research, resulting in part in the above noted articles, which while alluding generally to detrimental and dangerous conditions resulting from negligent use of radiation, fail to completely appreciate the problems, and the techniques and apparatus advanced in the articles, as also in the prior patents, have failed to solve some of the existing problems. A primary purpose of the present invention is to provide techniques and apparatus which help to overcome some of the existing problems, and to a very substantial extent fulfill a need in the medical and dental professions. SUMMARY OF THE INVENTION The present invention is primarily directed to an appliance devised to overcome some of the problems which are resulting effects of radiation, either directly or indirectly applied to teeth and surrounding patient areas and tissues, as also glands, in areas of a patient being treated. More specifically, the invention teaches a radiation shield intended to protect the teeth, gingiva (gums), periodontal bone and salivary glands against harmful side effects of radiation therapy. The shield, can consist of or include specifically different materials which are known to protect against radiation, such as lead in sheet form, or lead carried in or by another material, one known material for example consisting of multilayered plastic sheets, an intermediate one of the layers containing fine particles of lead. The shield can vary in thickness and other physical characteristics depending upon the intensity of radiation required for treatments, and also as equated to the patient being treated. In a preferred form of an appliance for practicing the invention, a composite or combined unit is used including an intraoral portion and an extraoral portion, selectively joinable. The intraoral portion is so designed as to substantially cover the teeth, gingiva, periodontal bone and related tissue to prevent impingement thereon of the radiation rays, and the extraoral portion, which in effect serves as a selectively usable attachment to the intraoral portion, provides a shield which covers the extraoral anatomy of the face and provides protection for the parotid and sublingual and salivary glands. The conditions among others which the oral radiation protector appliance is intended to protect against include: 1. Salivary gland atrophy and xerostomia (dry mouth). The salivary glands, when exposed to radiation atrophy (shrink) and their function markedly decreases or ceases altogether. This condition lasts from weeks to a permanent loss of function. The patient finds it difficult to chew, swallow and digest food. The food will also adhere to the dentition and periodontal areas increasing the liklihood of caries and periodontal disease. 2. The appliance is also intended to provide protection to the teeth. The dentition is very susceptible to the effects of radiation. Therapeutic radiation in the range of 2500 r to 5000 r causes great injury to teeth and bone. Most organs such as thyroid and brain radiation require over 5000 r. The oral radiation protector appliance will allow a patient to receive large doses in the head and neck region and yet preserve the integrity of the teeth. Should a patient fail to have protection, the dentition becomes susceptible to a condition known as "Radiation Caries." This condition is characterized by demineralization and breaking down of enamel. The teeth become brittle, usually getting cavities in the cervical region, and often breaking off at the gumline. 3. The appliance will also provide protection to children whose teeth are just erupting or are still unerupted. Children often require radiation treatment for head and neck tumors. Hemangioma's in children are also treated by radiation. Radiation therapy, whether direct or indirect, frequently results in stunted undeveloped roots, retarded eruption, and anodontia (missing teeth). The oral radiation protector appliance will aid in preventing this condition. 4. The alveolar bone which holds the teeth has been seen to resorb and become porous as a result of direct and indirect radiation. The teeth loosen and the patient becomes susceptible to periodontitis (pyohorrea). 5. Alveolar bone that has been exposed to radiation has a reduced blood supply and infected teeth are much more severe and dangerous due to this limited blood supply. 6. It has in many instances been the accepted principle to extract the teeth that are in an area to be radiated, as noted in the above mentioned articles. The oral radiation protector appliance in some cases make this procedure unnecessary, since the radiation would not severely damage the teeth and alveolar bone. 7. The oral radiation protector appliance also minimizes a condition known as osteoradionecrosis. The chronic pathosis in this condition is characterized by infection, pain and necrosis. Sequestrae of bone and overlying mucosa are common and in some cases deformity results. This infection can last from months to years, and has also been known to be fatal. The above listed articles discuss in greater detail some of these problems and it is to be noted that the application of the radiation rays discussed therein is to areas removed from direct impingement on teeth and areas proximate thereto. The direct impingement is understandably more devastating.
046541900	description	DETAILED DESCRIPTION The emergency feedwater system of the present invention provides for the separation of the feedwater supply into two separate subsystems and may be used in connection with two loop, three loop or four loop pressurized water reactor plants, the loop designation referring to the number of generators associated with a pressurized water nuclear reactor. Referring now to FIG. 1, an emergency feedwater system 1, is illustrated for use with a four loop system, one containing four steam generators S.sub.1, S.sub.2, S.sub.3, S.sub.4. As is conventional in such a four loop system, each of the steam generators S.sub.1, S.sub.2, S.sub.3 and S.sub.4 has a respective main feedwater system which supplies water to a respective generator through an inlet line 3, the line terminating at the steam generator at a nozzle 5, and having a check valve 7 therein, the check valve 7 being positioned in the inlet line 3 at a location outside the containment wall 9 which contains the reactor and associated equipment. In accordance with the present invention, two separate subsystems, designated as A and B are provided, each subsystem servicing a pair of generators. As indicated, area A services steam generators S.sub.1 and S.sub.2. The subsystems are located at separate physical locations and preferably, one such subsystem is provided on the opposite side of the containment from the other subsystem. Each subsystem contains the components for charging the steam generators with emergency feedwater, and in the interest of brevity, specific description of the subsystem area A will be made. The subsystem, area A, comprises an emergency feedwater tank 11 containing a supply of emergency feedwater 13. Leading from the emergency feedwater tank 11 are a pair of emergency feedwater lines 15 and 17. Emergency feedwater line 15 communicates with the inlet line 3 to steam generator S.sub.2 at 19, at a location between check valve 7 and nozzle 5, while the emergency feedwater line 17 communicates with the inlet line 3 to steam generator S.sub.1 at 21, at a location between check valve 7 and nozzle 5. In emergency feedwater line 15, there is provided a motor operated pump 23 which is electrically operated and has an electrical power source for activating the pump to discharge water from the emergency feedwater tank 11. Also provided in emergency feedwater line 15, between pump 23 and the communication of the line to inlet line 3 of steam generator S.sub.2 at 19, there is a cavitating venturi orifice 25. In emergency feedwater line 17, there is provided a steam turbine driven pump 27. A steam supply line 29 provides steam from the main steam system of the steam generator S.sub.1 and contains a valve means 31, such as a pneumatically-operated steam admission valve, which is in normally closed position. Activation of the steam driven pump will also discharge water from the emergency feedwater tank 11. In emergency feedwater line 17, between pump 27 and the communication of the line to inlet line 3 of steam generator S.sub.1 at 21, there is a cavitating venturi orifice 33. The cavitating venturies 25 and 33 are sized to cavitate and choke emergency feedwater flow in the emergency feedwater lines 15 and 17 to a specified flow. The cavitating venturies serve several purposes. In the event of a steamline or feedline rupture, the pumps 23 and 27 will discharge to a reduced pressure. In such cases, the cavitating venturies will choke the flow in each line to the specified flow and thereby prevent the pumps from being damaged by runout. Also, in the event of a steamline rupture, with all pumps operating, the cavitating venturies 25 and 33 prevent an excessive flow of emergency feedwater to the steam generators which could cause an unacceptably high cooldown rate of reactor coolant system components. Also, by limiting the emergency feedwater flow to the steam generators in the short term (i.e., before operator action can be assumed), the cavitating venturies prevent the steam generators from being filled solid with water and prevent the attendant problem of steamline flooding. In addition, in the event of a steamline rupture inside the containment, the cavitating venturies limit the emergency feedwater system contribution to the mass and energy released to the containment. Also, with use of a cavitating venturi in each of the emergency feedwater lines to each steam generator, a nearly balanced flow distribution can be maintained in the event of a steamline or feedline break, when one steam generator is depressurized to atmospheric pressure and the other intact steam generator is at design pressure. During normal emergency feedwater flow rates, the venturies will not cavitate and the permanent head loss caused by the cavitating venturies will be considerably less than that of an equivalent orifice. Flow modulating valves 35 are provided in each emergency feedwater lines 15 and 17 between the respective pumps 23 and 27 and the cavitating venturi orifices 25 and 33. These flow modulating valves 35 are normally open, fail open, air operated, hand controlled valves. A local namual override is provided on these valves to allow positioning of the valve in the event of loss of air, or a failure in the valve control circuitry. First and second connecting lines 37 and 39 are also provided which communicate between the emergency feedwater lines 15 and 17, one before and the other following the location of the flow modulating valves 35. The first connecting line 37, located between the pumps 23 and 27 and the flow modulating valves 35, contains valve means, such as motor operated valves 41 which are in a normally open position. These valves 41 allow the individual pumps to be remotely isolated from each other in the event of a passive failure, such as a pipe leak, while a local manual override is provided on each valve 41 to allow manual positioning in the event of control failure. The second connecting line 39, located between the flow modulating valves 35 and the cavitating venturi orifices 25 and 33, contains valve means, such as air operated valves 43, in a normally open position. These valves 43 allow the individual pumps to be remotely isolated from each other in the event of a passive failure or to establish individual steam generator flow control. Closure of one of the valves also terminates flow to a faulted steam generator following a feedline or steamline break to the generator. Each of the emergency feedwater lines 15 and 17 also are provided with valve means 45, such as an air operated, normally open, fail open, isolation valve downstream of the flow modulating valves 35. These valves 45 are used to isolate emergency feedwater flow to a faulted steam generator following a main feedline or main steamline rupture. These valves 45 can also be used for maintenance operations, or in the event of a passive failure, such as a pipe leak, and during such situations, serve as a barrier between the emergency feedwater system and the high temperature, high pressure water in the main feedwater system or in the steam generators. Valve means, such as manual valves 47, normally locked open, are located in the emergency feedwater lines 15 and 17 between the tank 11 and the pumps 23 and 27. These valves isolate an individual pump for maintenance operations, or in the event of a passive failure, such as a pipe leak. Additional valve means 49, such as locked open manual gate valves, are located at each pump discharge, in emergency feedwater lines 15 and 17, which valves are normally locked open and are closed only for pump maintenance. Check valves are also provided, as indicated at 51, at each pump discharge, in emergency feedwater lines 15 and 17, so as to prevent high pressure discharge from the operating pump from flowing in reverse direction through the non-operating pump and back into the low pressure suction piping and tank. Check valves 53 are also located in each emergency feedwater line 15 and 17 near the point where each of these lines communicate with the main feedwater lines 3. These check valves 53 prevent the flow of high temperature, high pressure water from the main feedwater system or the steam generators into the emergency feedwater lines when the emergency feedwater pumps are not operating. In order to assure reliability, the two subsystems are separated into two redundant load groups or electrical power trains (not shown). Each of these two electrical power trains is connected to both a preferred and a standby power supply. The preferred power supply consists of one or more circuits from the transmission network (offsite source) and a standby power supply consists of two emergency generators, such as diesel generators. Therefore, the vital bus in each of the two power trains is connected to either an offsite power source or to one emergency generator if offsite power is lost. During normal plant operation, offsite power is usually supplied through the plant startup transformer. However, should a loss of offsite power occur with a resultant plant trip, the standby power supply of the emergency generators would be available to supply system power requirements. The two electrical power trains provide sufficient physical and electrical separation and redundancy to prevent the occurrence of a common mode failure between the two subsystems. The emergency feedwater system is therefore able to accept a complete loss of one electrical train coincident with a loss of offsite power and still meet all system requirements. An alternate emergency feedwater source (not shown) is also provided, such as a single tank or reservoir, or several tanks and/or reservoirs. The alternate emergency feedwater source need not be a safety grade, but contains a sufficient quantity of water to allow the plant to be maintained in hot standby condition. Water from the alternate emergency feedwater source is chargeable through line 55 to the emergency feedwater tanks 11. An orificed recirculating line 57 is provided from each pump discharge back to the emergency feedwater tank 11. The orifice is sized to provide the required amount of recirculation flow for pump protection in the event the discharge flow paths are isolated. While the previous description refers to the components of subsystem A for servicing steam generators S.sub.1 and S.sub.2, it is to be understood by reference to the drawing, that subsystem B contains the components for servicing steam generators S.sub.3 and S.sub.4, which are located at a physically distinct location. The pumps 23 and 27 are sized such that, in events not involving a steamline or feedline rupture, only one pump is capable of supplying the minimum required emergency feedwater flow to the minimum number (2) of steam generators within one minute of system actuation, with the steam generators at a pressure equal to the setpoint of the lowest set safety relief valve in the main steam system, plus the accumulation of the safety valve. In the event of a main feedline or steamline rupture, it is postulated that none of the emergency feedwater flow in the affected subsystem will reach the steam generators. In such cases, the sizing of the emergency feedwater pumps 23 and 27 is such that either pump in the unaffected subsystem will be capable of supplying the minimum required emergency feedwater flow to the minimum number (2) of effective steam generators with the steam generator pressure as described above. The steam supply line 29 for each turbine pump 27 is connected to the main steam line from one steam generator only (steam generator S.sub.1 for the turbine pump in subsystem A and steam generator S.sub.3 for the turbine pump in subsystem B). The steam supply line to each turbine is fitted with a steam admission valve 31 which is a pneumatically-operated valve arranged to fail open on loss of air or electrical power. The steam admission valve 31 uses both subsystem A and subsystem B powered actuation trains to operate redundant solenoid valves to vent the air and open the steam admission valve. This ensures that a single failure of an actuation train will not incapacitate both the turbine driven pump 27 and the motor driven pump 23 in either subsystem. The emergency feedwater system 1 is not operated during normal plant operations, but remains in a state of readiness to provide emergency feedwater to the steam generators in the event of transient or accident conditions. In the event of such an occurrence, the emergency feedwater pumps are automatically started as follows: ______________________________________ Signal Pumps Started ______________________________________ Low-Low level in 2/4 Motor Driven Pumps 23 level channels in any one steam generator Low--low level in 2/4 Turbine Driven Pumps 27 level channels in any two steam generators Safety Injection Motor Driven Pumps 23 ______________________________________ Since all valves in the emergency feedwater lines 15 and 17 are open, the automatic startup of the pump, as indicated above, will result in the immediate delivery of emergency feedwater into the steam generators. The system is designed to supply at least the minimum required flow, within one minute of the actuation signal, to at least 2 effective steam generators, and to continue this delivery for an indefinite period without operator action. When operator action can be taken (after an assumed 30 minute delay) the emergency feedwater flow rate is adjusted by positioning the hand controlled flow modulating valves 35 so as to restore and maintain the steam generator water levels within the narrow control range. With the reactor tripped, and the emergency feed-water system supplying water to the steam generators at a rate equivalent to the rate at which steam is being removed to dissipate core decay heat and the heat input of one reactor coolant pump (assumed to be operating), the plant is in a stable hot standby condition. The plant can be maintained in this condition for a period limited only by the amount of water in the tank 11 and alternate emergency feedwater source. If the initiating event can be resolved, plant power operations can be resumed. Normal feedwater flow to the steam generators, by the main feedwater system, is resumed and the emergency feedwater pumps are manually stopped. If the initiating event cannot be resolved, a plant cooldown must be performed. In this case, the emergency feedwater system continues to supply feedwater to the system generators throughout the cooldown until the primary system hot leg temperature is reduced to the desired level. The residual heat removal system of the plant is then activated and the emergency feedwater system is secured by manually stopping the emergency feedwater pump. The residual heat removal system continues the cooldown to cold shutdown conditions. An embodiment of the present invention for use with a three loop system, a system containing three steam generators, indicated as S.sub.5, S.sub.6 and S.sub.7, is illustrated in FIG. 2. This embodiment 101 has the components of the four loop embodiment, with the numerals on the components corresponding to those in FIG. 1, with four emergency feedwater lines, 15, 15, 17 and 17 provided, except that one of the feedwater lines 15 and 15 of each subsystem combine to form a common discharge line to the third generator. As illustrated, emergency feedwater line 17 of subsystem A services steam generator S.sub.5, while emergency feedwater line 17 of subsystem B services steam generator S.sub.6. The third steam generator S.sub.7 has an inlet line 103 connected to a main feedwater system, the line 103 terminating at the steam generator S.sub.7 at a nozzle 105, and a check valve 107 is provided in line 103. The other emergency feedwater line 15 of subsystem A and the other emergency feedwater line 5 of subsystem B combine to form a common discharge line 109 which communicates at 111 with inlet line 103 at a location between check valve 107 and nozzle 105. As illustrated, this point of communication 111 may be inside the containment, i.e., on the inside of containment wall 9. In addition to the cavitating venturis 25 and 33 in emergency feedwater lines 15 and 17, a further cavitating venturi 113 is provided in the common discharge line 109. Also, an additional check valve 115 is provided in common discharge line 109 between the cavitating venturi 113 and the communication 111 of line 109 with inlet line 103. An embodiment of the present invention for use with a two loop system, a system containing two steam generators, indicated as S.sub.8 and S.sub.9, is illustrated in FIG. 3. This embodiment of the emergency feedwater system 201 also has two subsystems, A and B. Each of the two steam generators S.sub.8 and S.sub.9 has a main feedwater system which supplies water through an inlet line 203 to a steam generator through a nozzle 205, and a check valve 207 is present in each line 203, outside containment wall 209. There are two subsystems, A and B, provided which are located at physically distinct locations. Referring to subsystem A, which has identical components of subsystem B, an emergency feedwater tank 211 contains a supply of emergency feedwater 213, which is discharged to a pair of emergency feedwater lines 215 and 217. Emergency feedwater lines 215 and 217 combine to form a common discharge line 219, which common discharge line 219 communicates with the inlet line 203 to steam generator S.sub.8 at a location between check valve 207 and nozzle 205. In emergency feedwater line 215, there is provided a motor operated pump 223 which is electrically operated, while in emergency feedwater line 217 there is provided a steam turbine driven pump 225. A steam supply line 227 provides steam from the main steam system of the steam generator S.sub.8 and contains a valve means 229, such as a pneumatically-operated, normally-closed, steam admission valve. Activation of either the motor operated pump 223 or the steam driven pump 225 will discharge water from the emergency feedwater tank 211 through the respective emergency feedwater lines 215 and 217. A cavitating venturi orifice 231 is provided in the common discharge line 219. Flow modulating valves 233 are located in each of the emergency feedwater lines 215 and 217 between the respective pumps 223 and 225 and the common discharge line 219. The flow modulating valves 233 are normally open, fail open, air operated valves having a local manual override. These valves 233 will be normally open when the system is activated. They are provided to allow operator control of the emergency feedwater flow rates to the steam generator so that, in the long term, after operator action can be assumed, steam generator water levels can be maintained in a narrow control range. A safety grade air supply is provided for each valve 233, and a local manual override is provided to allow positioning in the event of loss of air, or a failure in the valve control circuitry. Valve means, such as normally open motor operated valves 235 are provided in each emergency feedwater line 215 and 217 downstream from the flow modulating valves 233. The valves 235 allow the emergency feedwater lines to be remotely isolated for maintenance in the event of a passive failure, such as a pipe leak. These valves 235 are also used to isolate emergency feedwater flow to a faulted steam generator following a minor feedline or main steamline rupture. A local manual override is provided on each valve 235 to allow manual positioning in the event of control failure. Locked open manual gate valves 237 are located between the emergency feedwater tank and each pump, while further locked open manual gate valves 239 are located between each pump and the flow modulating valves 233 in each of the emergency feedwater lines 215 and 217. Valves 237 are used to isolate an individual pump for maintenance operation in the event of a passive failure, such as a pipe leak. Valve 239 are closed only for pump maintenance. Check valves 241 are provided in each emergency feedwater line 215 and 217 between the respective pumps 223 and 225 and the common discharge line 219. In the event that any one pump in a subsystem is started, these check valves 241 prevent the high pressure discharge of the operating pump from flowing in a reverse direction through the non-operating pump and back into the low pressure emergency feedwater tank 211. Also, a check valve 243 is located in the common discharge line 219 near the point of communication 221 with the main feedwater inlet line 203. This check valve 243 prevents the flow of high temperature, high pressure water from the main feedwater system or the steam generator from flowing into the emergency feedwater system when the emergency feedwater pumps 223 and 225 are not operating. As with the previous embodiments, two redundant load groups or electrical power trains (not shown) are used with the two subsystems to assure reliability. An alternate emergency feedwater source (not shown) is provided which charges the feedwater tank 211 through line 245. Also, an orificed recirculation line 247 is provided from each pump discharge back to the emergency feedwater tank 211. The operation of the subsystem in the two loop embodiment is similar to that of the other embodiment, with the pumps sized such that only one pump in either subsystem is capable of supplying the minimum required feedwater flow to an effective steam generator in the event of a main feedline or steamline rupture affecting one of the steam generators. The steam supply line for each turbine driven pump is connected to the main steamline from one steam generator only, and fitted as in the previously described embodiments.
046769455	claims	1. In a nuclear reactor facility having fuel bundles, a system for the insertion of a fuel bundle into a position wherein a plurality of vertically arranged fuel bundles surround and are adjacent to said position, said system comprising, in combination, a plurality of separate and individual centering devices secured to and disposed on top of each fuel bundle adjacent said position, each such centering device having a generally box-like cap configuration on the upper end of each fuel bundle and including: a top wall, first and second side walls, each secured along an upper edge to said top wall, a rear plate attached along opposite vertical edges to said first and second side walls, a front inclined wall joined along an upper edge to the top wall and attached along opposite vertical edges to said first and second side walls, a plurality of pad means secured to the lower edge of said first and second side walls, said front inclined wall and said rear plate for mounting each said centering device on top of an associated fuel bundle, pin means carried by at least two of said pad means engageable with an associated aperature in said fueld bundle for locating and laterally fixing each said centering device on top of its respective fuel bundle, each said front inclined wall of each of said centering devices being orientated on top of its respective fuel bundle to slope upwardly and away from the position whereby upon downward insertion of a fuel bundle into said position any contact between the lower end of the fuel bundle being inserted with a front inclined wall of a centering device will laterally deflect the fuel bundle into said position, each said centering device further including central socket means secured to said top wall, and an elongated handling pole means pivotally attached to said socket means. a top wall, first and second side walls, each secured along an upper edge to said top wall, a rear plate attached along opposite vertical edges to said first and second side walls and having an upper edge portion spaced from and terminating substantially below said top wall to define a fluid flow passageway therebetween, a front inclined wall joined along an upper edge to the top wall and attached along opposite vertical edges to said first and second side walls, said inclined wall sloping upwardly and rearwardly toward the rear plate, a plurality of four pad means secured beneath each of the four vertical edge intersections between said first and second side walls, said front inclined wall and said rear plate for mounting said centering device on top of an associated fuel bundle, and pin means carried by at least two of said four pad means for engagement with an associated aperature in a fuel bundle for locating and laterally fixing said centering device on top thereof, each said centering device further including central socket means secured to said top wall, and an elongated handling pole means pivotally attached to said socket means. 2. The combination of claim 1 wherein said socket means and pole means are disposed such that said pole means is movable to an inclined position to avoid interference with a fuel bundle being inserted into said position. 3. The combination of claim 2 wherein said socket means includes a downwardly facing conical portion and wherein said pole means includes a pin extension projecting through said socket means and having a chamfer plate attached thereto whereby upon lifting of said centering device by said pole means, said chamfer plate will engage and align with said conical portion. 4. The combination of claim 3 wherein said pin means comprises a pair of conical dowels disposed on diagonally opposite corners of each said centering device. 5. In a nuclear reactor facility having fuel bundles, a system for the insertion of a fuel bundle into a position wherein a plurality of vertically arranged fuel bundles surround and are adjacent to said position, said system comprising a centering device to assist in the insertion of a fuel bundle into an empty position of a nuclear reactor having a plurality of vertically arranged fuel bundles surrounding and adjacent said position, said centering device having a generally box-like cap configuration for placement upon the upper end of a fuel bundle and comprising: 6. The combination of claim 5 wherein said socket means and pole means are disposed such that said pole means is movable to an inclined position to avoid interference with a fuel bundle being inserted into said position. 7. The combination of claim 6 wherein said socket means includes a downwardly facing conical portion and wherein said pole means includes a pin extension projecting through said socket means and having a chamfer plate attached thereto whereby upon lifting of said centering device by said pole means, said chamfer plate will engage and align with said conical portion. 8. The combination of claim 7 wherein said pin means comprise a pair of conical dowels disposed on diagonally opposite corners of said centering device. 9. The combination of claim 5 wherein a lower edge portion of said first and second side walls, said rear plate and said front inclined wall is arched upwardly toward said top wall to define side fluid flow openings to permit a low resistance escape for any reactor core cooling fluid flowing upwardly through the fuel rods of the fuel bundle below said centering device.
	abstract	One embodiment of the present invention includes a process for production and recovery of no-carrier-added radioactive tin (NCA radiotin). An antimony target can be irradiated with a beam of accelerated particles forming NCA radiotin, followed by separation of the NCA radiotin from the irradiated target. The target is metallic Sb in a hermetically sealed shell. The shell can be graphite, molybdenum, or stainless steel. The irradiated target can be removed from the shell by chemical or mechanical means, and dissolved in an acidic solution. Sb can be removed from the dissolved irradiated target by extraction. NCA radiotin can be separated from the remaining Sb and other impurities using chromatography on silica gel sorbent. NCA tin-117m can be obtained from this process. NCA tin-117m can be used for labeling organic compounds and biological objects to be applied in medicine for imaging and therapy of various diseases.
	description	The present invention refers to an electron exit window foil. More particularly, the present invention relates to an electron exit window foil for use in a corrosive environment and operating at a high performance. Electron beam devices may be used to irradiate objects with electrons, e.g. for surface treatment. Such devices are commonly used within the food packaging industry, where electron beams are providing efficient sterilization of packages, e.g. plastic bottles or packaging material to be later converted into a package. A main advantage with electron beam sterilization is that wet chemistry, using e.g. H2O2, may be avoided thus reducing the high number of components and equipment required for such wet environments. An electron beam device typically comprises a filament connected to a power supply, wherein the filament is emitting electrons. The filament is preferably arranged in high vacuum for increasing the mean free path of the emitted electrons and an accelerator is directing the emitted electrons towards an exit window. The electron exit window is provided for allowing the electrons to escape from the electron beam generator so they may travel outside the electron beam generator and thus collide with the object to be sterilized and release its energy at the surface of the object. The electron exit window typically consists of thin electron permeable foil that is sealed against the electron beam generator for maintaining the vacuum inside the electron beam generator. A cooled support plate in the form of a grid is further provided for preventing the foil to collapse due to the high vacuum. Ti is commonly used as the foil material due to its reasonable good match between high melting point and electron permeability, as well as the ability to provide thin films. A problem with a Ti film is that it may oxidize, leading to reduced lifetime and operational stability. In order to achieve a long lifetime of the exit window, a maximum temperature of approximately 250 C should not be exceeded during the operation of the electron beam device. Typically, a high performance electron beam device is designed to provide 22 kGy at up to 100 m/min at 80 keV when used for sterilizing a running web of material. A plain Ti foil may thus not be used with such high performance electron beam devices, since the amount of emitted electrons transmitted through the window may cause temperatures well above this critical value. In filling machines, i.e. machines designed to form, fill, and seal packages, sterilization is a crucial process not only for the packages, but for the machine itself. During such machine sterilization, which preferably is performed during start-up, the outside of the exit window will be exposed to the chemicals used for machine sterilization. A highly corrosive substance such as H2O2, which is commonly used for such applications, will affect the exit window by means of etching the Ti. Different solutions for improving the properties of the exit window have been proposed to overcome the above-mentioned drawbacks. EP0480732B describes a window exit foil consisting of a Ti foil, and a protective layer of Al that is forming an intermetallic compound by thermal diffusion treatment of the Ti/Al construction. This solution may be suitable for relatively thick exit windows, i.e. windows allowing a protective layer being thicker than 1 micron. However, an intermetallic compound is not acceptable on a thin Ti foil since it would reduce its physical strength. EP0622979A discloses a window exit foil consisting of a Ti foil and a protective layer of silicon oxide on the side of the exit foil facing the object to be irradiated. Although the Ti foil may be protected by such layer, silicon oxide is very brittle and may easily crack in the areas where the foil is allowed to flex, i.e. the areas between the grids of the supportive plate when vacuum is provided. This drawback is making the foil of EP0622979A unsuitable for applications where the exit foil is exhibiting local curvatures, such as electron beam devices using a grid-like cooling plate arranged in contact with the exit foil. An object of the present invention is to reduce or eliminate the above-mentioned drawbacks. A further object is to provide an electron exit foil that is able to decrease the heat load as well as the corrosion on the foil. An idea of the present invention is thus to provide an electron beam generator having a prolonged operating lifetime, requiring a reduced service, and being more cost-effective than prior art systems due to inexpensive coating processes and the appliance of well-established X-ray manufacturing processes. According to a first aspect of the invention, an electron exit window foil for use with a high performance electron beam generator operating in a corrosive environment is provided. The electron exit window foil comprises a sandwich structure having a film of Ti, a first layer of a material having a higher thermal conductivity than Ti, and a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment. The first layer may be arranged between the film and the second layer, or the film may be arranged between the first layer and the second layer. The second layer may comprise at least two layers of different materials, which is advantageous in that different mechanical and/or physical properties of the foil, such as erosion resistance and strength, may be tailor made for the particular application. The first layer may be selected from a group consisting of materials having a ratio between thermal conductivity and density being higher than of Ti. The first layer may be selected from the group consisting of Al, Cu, Ag, Au, or Mo, and the second layer may be selected from the group consisting of Al2O3, Zr, Ta, or Nb. The corrosive environment may comprise H2O2. Hence, the foil may be implemented in electron beam devices operating in machines being subject to corrosive sterilizing agents, such as for example filling machines within the food packaging industry. The electron exit window foil may further comprise at least one adhesive coating between the Ti film and first layer or the second layer. Said adhesive coating may be a layer of Al2O3 or ZrO2 having a thickness between 1 and 150 nm. This is advantageous in that any reaction or material diffusion is prevented at the film/layer interface or the adhesion between the Ti film and a layer or between two layers is improved. According to a second aspect, an electron beam generator configured to operate in a corrosive environment is provided. The electron beam generator comprises a body housing and protecting an assembly generating and shaping the electron beam, and a support carrying components relating to the output of the electron beam, said support comprising an electron exit window foil according to the first aspect of the invention. The advantages of the first aspect are also applicable for the second aspect of the invention. According to a third aspect of the invention, a method for providing an electron exit window foil for use with a high performance electron beam generator operating in a corrosive environment is provided. The method comprises the steps of providing a film of Ti, providing a first layer of a material having a higher thermal conductivity than Ti onto a first side of said film, and providing a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment. The step of providing a flexible second layer may comprise arranging said flexible second layer onto a second side of said film. The step of providing a flexible second layer may comprise arranging said flexible second layer onto said first layer. At least one of steps of providing a first layer or providing a flexible second layer may be preceded by a step of providing an adhesive coating onto said film. According to a fourth aspect of the invention, a method for providing a high performance electron beam device is provided. The method comprises the steps of attaching a film of Ti onto a frame, processing said film by providing a first layer of a material having a higher thermal conductivity than Ti onto a first side of said film, and providing a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment, and attaching said foil-frame subassembly to a tube housing of an electron beam device for sealing said electron beam device. The advantages of the first aspect of the invention are also applicable for the third and the fourth aspects of the invention. With reference to FIG. 1 an electron beam device is shown. The electron beam device 100 comprises two parts; a tube body 102 housing and protecting an assembly 103 generating and shaping the electron beam, and a supportive flange 104 carrying components relating to the output of the electron beam, such as a window foil 106 and a foil support plate 108 preventing the window foil 106 from collapsing as vacuum is established inside the device 100. Further, during operation of the electron beam device the foil 106 is subject to excessive heat. Thereby, the foil support plate 108 also serves the important purpose of conducting heat generated in the foil 106 during use away from the foil 106 of the device. By keeping the foil temperature moderate a sufficiently long lifetime of the foil 106 may be obtained. With reference to FIG. 2, an electron exit window is shown comprising the foil 106 and the foil support 108. The support 108 is arranged inside the electron beam device such that vacuum is maintained on the inside of the exit window. This is indicated by P1 and P2 in FIG. 2, where P1 denotes atmospheric pressure outside the exit window and P2 represents vacuum on the inside. During manufacturing, the foil support plate 108, being of copper, is preferably attached to the flange 104 forming a part of the tube body 102. The flange 104 is generally made of stainless steel. The window foil 106 is then bonded onto a separate frame thus forming a foil-frame sub assembly. The foil 106 is subsequently coated, in order to improve its properties regarding for instance heat transfer. The foil-frame subassembly is subsequently attached to the tube body 102 to form a sealed housing. In an alternative embodiment, the exit window foil 106 is attached directly to the flange, being attached to the support plate, before the flange is welded to the tube body. In this embodiment, the exit window foil is consequently coated prior to being attached to the tube body 102. With reference to FIG. 3a-f, different embodiments of an electron exit window foil 106a-f are shown. Starting with FIG. 3a, the foil 106a comprises a thin film of Ti 202. The Ti film 202 has a thickness of approximately 5 to 15 microns. On a first side of the Ti film 202, a thermally conductive layer 204 is arranged. The thermally conductive layer 204 is provided in order to transfer heat along the exit foil such that a reduced temperature is achieved across the entire foil 106. The thermally conductive layer 204 is provided by means of any suitable process, such as sputtering, thermal evaporation, etc, and should allow for a sufficient improvement in thermal conductivity for lowering the temperature of the electron exit window foil 106a while still allowing the foil to bend into the apertures of the support plate 108 when vacuum is applied. Preferably, the material of the thermally conductive layer 204 is chosen from the group consisting of Al, Cu, Ag, Au, and Mo. Although other materials, such as Be, may have a higher ratio between thermal conductivity and density they are considered as poisonous and hence not preferred, especially in applications in which the electron beam device is arranged to process consumer goods. On the other side of the Ti film a protective layer 206 is arranged. The protective layer 206 is provided by means of any suitable coating process, such as sputtering, thermal evaporation, etc. Preferably, the material of the protective layer is chosen from the group consisting of Al2O3, Zr, Ta, and Nb due to their resistance against hydrogen peroxide containing environments. It should thus be understood that the protective layer 206 is facing the atmospheric environment, i.e. the objects to be sterilized. The thickness of the thermally conductive layer 204 is preferably between 1 and 5 microns and the thickness of the protective layer 206 is substantially less than 1 micron. Preferably, the thickness of the protective layer 206 is approximately 200 nm. By keeping the window foil 106 as thin as possible, the electron output is maximized. The thickness of the protective layer 206 should thus be designed such that it is capable of protecting the Ti film from a) corrosion by hydrogen peroxide or other aggressive chemical agents which may be provided in the particular application, and b) corrosion caused by the plasma created by the electrons in the air. Further, the thickness of the protective layer 206 should ensure tightness and physical strength, such that the second layer 206 is flexible in order to allow the entire foil to bend and conform to the apertures of the support plate 108 when vacuum is applied. A yet further parameter may be the density, for allowing electron transmittance through the protective layer 206. By arranging the thermally conductive layer 204 and the protective layer 206 on opposite sides of the Ti foil, stress in the layers may be reduced. For example, if using Al as the thermally conductive layer and Zr as the protective layer, the Ti foil arranged in between those layers will reduce some of the stress induced upon heating. This is due to the fact that the coefficient of thermal expansion of Ti lies between the corresponding value of Al and Zr. FIG. 3b shows another embodiment of a foil 106b. Here, the thermally conductive layer 204 and the protective layer 206 are provided on the same side of the Ti film 202 such that the protective layer 206 is coated directly on the thermally conductive layer 204. This structure may be advantageous for electron beam devices, for which the electron exit window foil must be mounted to the tube housing before coating, i.e. not allowing coating of the side of the Ti foil facing the interior of the electron beam device. FIGS. 3c and 3d show two different embodiments similar to what has been previously described with reference to FIGS. 3a and b. However, in FIGS. 3c and 3d the protective layer 206 comprises at least two layers of different materials; a first layer 208 and a second layer 209. The first layer 208 and the second layer 209 of the protective layer 206 are both selected from the group consisting of Al2O3, Zr, Ta, and Nb, or alloys thereof. It should however be understood that each one of the layers 208, 209 could per se be a sandwich of two or more protective layers. For example, the corrosion protection layer 206 itself could be a multilayer structure comprising an oxide, a metal, an oxide, a metal, etc. According to a specific embodiment such multilayer structure may be formed by a first layer of ZrO2, a second layer of Zr, a third layer of ZrO2, and a fourth layer of Zr. This is advantageous in that a potential disruption in one of the sub layers does not induce a significant reduction of the overall corrosion protection of the protective layer 206. In order to achieve good adherence between the different layers/films of the electron exit window foil, adhesive barrier coatings may be provided at the interface. Such coatings may be a thin layer of Al2O3 or ZrO2, having a thickness between 1 to 150 nm, preferably between 50 and 100 nm. The use of such coatings is advantageous in that they prevent any reaction or diffusion of material at the interface between Ti and the thermally conductive layer and/or the protective layer. Reaction or diffusion may result in the formation of an intermetallic compound which negatively changes the characteristics of the materials involved. In the case of a thin Ti foil it may get reduced physical strength. Further, the presence of intermetallic compounds may reduce the thermal conductivity and the corrosion protective ability of the thermally conductive layer 204 and the protective layer 206 respectively. FIG. 3e describes a further embodiment similar to that of FIG. 3a but provided with barrier coatings of the kind described above. The electron exit window foil 106e comprises a sandwich structure having a film of Ti, a first layer 204 of Al having a higher thermal conductivity than Ti, and a flexible second layer 206 of Zr being able to protect said film 202 from a corrosive environment, wherein the second layer 206 is facing the corrosive environment. The thermally conductive layer 204 of aluminium (Al) is arranged on a first side of the titanium (Ti) foil. A first barrier coating 210a of zirconium oxide (ZrO2) is provided in between said Ti film 202 and said Al layer 204. On the other side of the Ti film 202 the protective layer 206 of zirconium (Zr) is arranged. A second barrier coating 210b of zirconium oxide (ZrO2) is provided in between the Ti film and the Zr layer 206. This embodiment is advantageous in that the Ti foil 202 is surrounded on one side by Al as the thermally conductive layer and on the other side by Zr as the protective layer. Since the coefficient of thermal expansion of Ti lies between the corresponding values of Al and Zr, some of the stress induced during heating of the foil will be reduced. As an alternative one or both of the barrier coatings 210a, 210b may instead be made of aluminium oxide (Al2O3). It is an advantage if the barrier coatings are based on a material provided in either the thermally conductive layer or in the protective layer. For example, if the protective layer is zirconium and the thermally conductive layer is aluminium, it is preferred that either aluminium oxide or zirconium oxide are used for the barrier coatings. This is due to the fact that the layers are applied by a sputtering machine. In a such machine sputter targets are used, one for each material that should be deposited. One and the same target can be used for both e.g. zirconium and zirconium oxide. The same applies for aluminium and aluminium oxide. Hence, it is preferred if the barrier coating is an oxide of a material used in either the corrosion protection or the thermal conductivity layer. FIG. 3f shows an embodiment similar to that of FIG. 3b but provided with barrier coatings. The electron exit window foil 106f comprises a sandwich structure having a film of Ti, a first layer 204 of Al having a higher thermal conductivity than Ti, and a flexible second layer 206 of Zr being able to protect said film 202 from a corrosive environment, wherein the second layer 206 is facing the corrosive environment. The thermally conductive layer 204 and the protective layer 206 are provided on the same side of the Ti film 202. On top of the Ti film 202 a first barrier coating 210a is coated. The barrier coating 210a is made of aluminium oxide (Al2O3). The thermally conductive layer 204 of aluminium (Al) is coated on said first barrier coating 210a. On the thermally conductive layer 204 there is in turn coated a second barrier coating 210b. Said barrier coating 210b is also made of aluminium oxide (Al2O3). Finally, the protective layer 206, being made of zirconium (Zr), is coated on said second barrier coating 210b. As an alternative one or both of the barrier coatings 210a, 210b may instead be made of zirconium oxide (ZrO2). In both the embodiments of FIGS. 3e and 3f the protective layer 206 may be a multilayer structure as described in relation to FIGS. 3c and 3d. The invention has mainly been described above with reference to a few embodiments. However, as is readily appreciated by a person skilled in the art, other embodiments than the ones disclosed above are equally possible within the scope of the invention, as defined by the appended patent claims.
043057837	summary	This invention relates generally to plasma devices, particularly such devices of the tokamak type. More particularly, the present invention relates to the generation of toroidal magnetic fields in such devices utilizing liquid metal to form coils for generating toroidal magnetic fields. Tokamak devices are devices in which plasma is created in a toroidal space and is confined therein by an appropriate combination of toroidal and poloidal magnetic fields. Such devices are useful in the study and analysis of plasmas, and particularly in the generation, confinement, study and analysis of hydrogenic plasmas. Such devices are among the most useful of known plasma devices for the reaction of deuterium and tritium with the production of high energy neutrons as reaction products. The present invention finds particular utility in respect to such devices and their applications, including experimental devices and the use thereof in experimentation and investigation in respect to toroidal plasma devices of the tokamak type. In tokamak devices, gases are disposed in a toroidal confinement vessel. The gases are ionized to produce a plasma that is heated and confined by appropriate magnetic and electrical fields. The principal field is a toroidal magnetic field conventionally created by electrical coils linking the torus. A serious difficulty with such coils, particularly where high fields are created in a small space, has been occasioned by the very great mechanical forces and stresses created in the coils and their supports. In accordance with the present invention, the toroidal field coil is formed of a single turn of liquid metal, whereby the fluid nature of the metal relieves all internal stresses; a pressure vessel contains the liquid metal and isolates the metal mechanically from the outside environment.
046769463	abstract	A thermal insulating blanket used for insulating pipes and equipment in nuclear power plant containment buildings is resistant to tearing if impinged upon by a high force liquid stream (i.e., during a loss of coolant accident--LOCA) and if torn is cut into small pieces which will not clog the protective screen of the emergency cooling system recirculating sump. The blanket consists of a filler layer of thermal insulating fibers, such as glass fiber wool or ceramic fiber wool, a waterproof sheet, and a wire mesh casing surrounding the filler layer and sheet. The wire casing dissipates the force of high force liquid streams which may strike the blanket during a LOCA and in the event that the filler layer is torn, the filler layer is cut into small pieces as it passes through the casing. Preferably, a wire mesh spetum lies within the filler layer to assist in cutting up the torn filler layer.
	abstract	A pharmaceutical pig is used to transport a syringe containing a liquid radiopharmaceutical from a radiopharmacy to a medical facility for administration to a patient. The pharmaceutical pig includes an elongate polymer cap that is removably attached to an elongate polymer base. The elongate polymer cap includes a cap shell that completely encloses a cap shielding element and the elongate polymer base includes a base shell that completely encloses a base shielding element. Preferably the polymer utilized for the cap shell and the base shell is polycarbonate resin, e.g., LEXAN®. An inner liner is not utilized and the cap shielding element and the base shielding element, which are preferably, but not necessarily, made of lead, are completely sealed and unexposed.
	claims	1. A multi-channel ion implantation system comprising:a beam source that generates a beamlet array; anda beamline assembly that comprises:a mass analyzer module that operates on the beamlet array to remove ions having a non-selected mass energy product; anda beam formation component that combines the beamlet array into a single ion beam after the ions having the non-selected mass energy product have been removed from the beamlets. 2. The system of claim 1, wherein the mass analyzer module comprises a first array of magnets defining channels between pairs of magnets through which the beamlets travel subjected to magnetic fields, wherein the first array of magnets deflect the beamlet array by a first angle. 3. The system of claim 2, wherein the mass analyzer module further comprises a slit that blocks the ions having the non-selected mass energy product and passes ions having a selected mass energy product. 4. The system of claim 1, wherein the beam formation component comprises vertical deflection plates positioned between rows of beamlets of the array that cause the beamlets to diverge in a vertical direction. 5. The system of claim 1, wherein the beam formation component comprises a scanning mechanism that scans the beamlets in a vertical direction. 6. A multi-channel ion implantation system comprising:a beam source and triode extraction assembly that generate a beamlet array; anda beamline assembly comprising an array of channels through which the beamlet array passes, the beamline assembly comprising:a first array of magnets comprised of pairs of permanent magnets defining a first magnetic field between the pairs of magnets;first beamguide spacers positioned in between columns of the first array of magnets, wherein the first beamguide spacers comprise channels through which the beamlet array passes;a first slit positioned downstream of the first array of magnets and having a first resolution that permits passage of ions having a selected mass energy product and substantially blocks other ions having a non-selected mass energy product;horizontal partition plates positioned downstream of the first array of magnets and positioned between rows of magnets of the first array of magnets;a second array of magnets positioned downstream of the vertical deflection plates and comprised of pairs of permanent magnets defining a second magnetic field between the pairs of magnets;second beamguide spacers positioned in between columns of the second array of magnets, wherein the second beamguide spacers comprise channels through which the beamlet array passes;a second slit positioned downstream of the second array of magnets having a second resolution that permits passage of ions having the selected mass energy product and substantially blocks other ions having a non-selected mass energy product;vertical deflection plates positioned downstream of the second slit and in between rows of the beamlet array that cause the beamlets within the array to diverge in a vertical direction; anda drift region positioned downstream of the vertical deflection plates that cause the beamlets within the array to diverge in a horizontal direction and form into a single ion beam. 7. The system of claim 6, further comprising an end station comprising a target that is implanted by the single ion beam. 8. The system of claim 6, wherein the first magnetic field deflects the beamlet array by 45 degrees in a horizontal direction. 9. The system of claim 8, wherein the second magnetic field deflects the beamlet array by 45 degrees opposite the deflection of the first magnetic field. 10. A method for generating a low energy, high current ion beam comprising:generating an array of beamlets;performing a mass analysis on the array of beamlets that causes ions having a selected mass energy product to deflect at a first angle;blocking other ions having a non-selected mass energy product;diverging the array of beamlets in a horizontal direction and a vertical direction to form the low energy, high current ion beam; andmeasuring beam current uniformity of the low energy, high current ion beam and adjusting beam current of individual beamlets of the array of beamlets according to the measured beam current uniformity. 11. A multi-channel ion implantation system comprising:a beam source that generates a beamlet array, the beam source comprising: a plasma source, a power source, and a triode extraction assembly; anda beamline assembly that comprises:a mass analyzer module that operates on the beamlet array to remove ions having a non-selected mass energy product; anda beam formation component that combines the beamlet array into a single ion beam. 12. A multi-channel ion implantation system comprising:a beam source that generates a beamlet array; anda beamline assembly that comprises:a mass analyzer module that operates on the beamlet array to remove ions having a non-selected mass energy product, the mass analyzer module comprising:a first array of magnets defining channels between pairs of magnets through which the beamlets travel subjected to magnetic fields, wherein the first array of magnets deflect the beamlet array by a first angle; anda second array of magnets defining channels between pairs of magnets through which the beamlets travel, wherein the second array of magnets deflect the beamlet array by a second angle; and a beam formation component that combines the beamlet array into a single ion beam. 13. The system of claim 12, wherein the second angle is opposite the first angle. 14. The system of claim 12, wherein the mass analyzer module further comprises horizontal partition plates positioned between the first array of magnets and the second array of magnets that block separating rows of that mitigate cross channel contamination. 15. The system of claim 14, wherein the horizontal partition plates further comprise electrodes associated with individual beamlets of the beamlet array. 16. The system of claim 15, wherein the electrodes are biased to generate an electric field across the individual beamlets. 17. A multi-channel ion implantation system comprising:a beam source that generates a beamlet array; anda beamline assembly that comprises:a mass analyzer module that operates on the beamlet array to remove ions having a non-selected mass energy product, the mass analyzer module comprising:a first array of magnets defining channels between pairs of magnets through which the beamlets travel subjected to magnetic fields, wherein the first array of magnets deflect the beamlet array by a first angle, wherein the first array of magnets includes beamguide spacers that physically separates columns of the magnets; and a beam formation component that combines the beamlet array into a single ion beam. 18. The system of claim 17, wherein the beamguide spacers have the channels formed therein. 19. The system of claim 18, wherein the channels formed within the beamguide spacers have a size and shape selected to substantially permit ions having the selected mass energy product to pass and substantially block ions having other mass energy products. 20. A multi-channel ion implantation system comprising:a beam source that generates a beamlet array; anda beamline assembly that comprises:a mass analyzer module that operates on the beamlet array to remove ions having a non-selected mass energy product;a beam formation component that combines the beamlet array into a single ion beam; anda beamline current adjustment module downstream of the mass analyzer module that is controllable to individually adjust beamlet currents of the beamlet array. 21. The system of claim 20, wherein the beamline current adjustment module comprises a number of horizontal plates positioned between rows of the beamlet array, wherein the number of horizontal plates comprise electrodes that controllably applies electric fields to the beamlet array. 22. The system of claim 21, further comprising a controller that controls the electric fields generated by the current adjustment module to obtain a desired beam current distribution. 23. A method for generating a low energy, high current ion beam comprising:generating an array of beamlets;performing a mass analysis on the array of beamlets that causes ions having a selected mass energy product to deflect at a first angle;performing a second mass analysis on the array of beamlets that causes ions having a selected mass energy product to deflect at a second angle;blocking other ions having a non-selected mass energy product; anddiverging the array of beamlets in a horizontal direction and a vertical direction to form the low energy, high current ion beam. 24. A method for generating a low energy, high current ion beam comprising:generating an array of beamlets;performing a mass analysis on the array of beamlets that causes ions having a selected mass energy product to deflect at a first angle;blocking other ions having a non-selected mass energy product; anddiverging the array of beamlets in a horizontal direction and a vertical direction to form the low energy, high current ion beam; wherein diverging the array of beamlets in the horizontal direction comprises employing a drift region having a length selected to provide an amount of horizontal divergence. 25. The system of claim 20, further comprising an acceleration assembly positioned downstream of the vertical deflection plates that accelerates the beamlets to a selected energy.
	abstract	A method and apparatus for a fret resistant fuel rod for a Boiling Water Reactor (BWR) nuclear fuel bundle. An applied material entrained with fret resistant particles is melted or otherwise fused to a melted, thin layer of the fuel rod cladding. The applied material is made of a material that is chemically compatible with the fuel rod cladding, allowing the fret resistant particles to be captured in the thin layer of re-solidified cladding material to produce an effective and resilient fret resistant layer on an outer layer of the cladding.
	description	Priority is claimed to Japanese Patent Application No. 2011-142691, filed Jun. 28, 2011, and International Patent Application No. PCT/JP2012/065386, the entire content of each of which is incorporated herein by reference. 1. Technical Field The present invention relates to a charged particle beam irradiation device that irradiates a body to be irradiated with a charged particle beam. 2. Description of the Related Art In the related art, a charged particle beam irradiation device disclosed in the related art is known as the charged particle beam irradiation device used for radiation therapy. The related art describes a particle beam treatment device that scans a charged particle beam by a scanning method and irradiates a body to be irradiated, such as a part of body with cancer, with the beam and that includes an inert gas chamber that is provided on a path of the charged particle beam, an integral gas supply pipe that supplies inert gas into the inert gas chamber, and a control device that controls the amount of gas supply of a gas supply pipe based on the difference between the internal and external pressures of the inert gas chamber. In the aforementioned particle beam treatment device, isolating films are provided at an inlet and an outlet of the inert gas chamber to secure airtightness, and inert gas is supplied into the inert gas chamber to prevent damage of the isolating films by the difference between the internal and external pressures. Also, by arranging this inert gas chamber on the path of the charged particle beam, scattering of the charged particle beam under the influence of air is avoided and thus, irradiation position accuracy is improved. According to an embodiment of the present invention, there is provided a charged particle beam irradiation device including an accelerator configured to accelerate charged particles and emit a charged particle beam; an irradiation unit configured to irradiate a body to be irradiated with the charged particle beam; a duct configured to transport the charged particle beam emitted from the accelerator to the irradiation unit; a tubular body that is arranged on a propagation path of the charged particle beam within the irradiation unit, has inert gas filled therein, and has particle beam transmission films transmitting the charged particle beam therethrough at an inlet and an outlet thereof; a gas supply unit configured to supply the inert gas into the tubular body; and a leak valve configured to leak the inert gas inside the tubular body to the outside when the internal pressure of the tubular body is equal to or higher than a set pressure. The gas supply unit has a plurality of supply lines having different amounts of supply of inert gas. In the particle beam irradiation device of the related art mentioned above, to realize the control of the amount of supply of the inert gas, there are problems because it is necessary to provide a control device that determines a gas flow rate based on the difference between the internal and external pressures of the inert gas chamber, an actuator for driving the flow control valve of the gas supply pipe, or the like and in that the configuration or control of the device is complicated. Thus, it is desirable to provide a charged particle beam irradiation device that can manage the amount of gas supply to a tubular body with a simple configuration. According to the charged particle beam irradiation device related to the embodiment of the invention, since there is almost no necessity for changing the amount of supply of the inert gas to the tubular body in normal use, the amount of gas supply to the tubular body can be easily managed by adopting a configuration in which a line with a suitable amount of supply is selected from the plurality of supply lines based on a situation. Moreover, according to the charged particle beam irradiation device, compared to the configuration of the related art in which the amount of supply is always controlled, the amount of supply of the inert gas to the tubular body can be managed with an extremely simple configuration. Additionally, since the inert gas is leaked through the leak valve when the internal pressure of the tubular body rises excessively, it is easy to manage the internal pressure of the tubular body within a desired range. In the charged particle beam irradiation device related to the embedment of the invention, the plurality of supply lines may include a pressure-maintaining supply line for maintaining the pressure inside the tubular body at a predetermined value, an adjusting supply line for adjusting the pressure within the tubular body, and a substituting supply line for substituting air inside the tubular body with the inert gas, the amount of supply of the inert gas of the adjusting supply line is larger than the amount of supply of the inert gas of the pressure-maintaining supply line, and the amount of supply of the inert gas of the substituting supply line may be larger than the amount of supply of the inert gas of the adjusting supply line. According to the charged particle beam irradiation device related to the embodiment of the invention, management of the amount of gas supply based on a situation can be realized by performing gas supply by the pressure-maintaining supply line at a normal time and using the adjusting supply line with a slightly larger amount of supply at the time of adjustment, such as maintenance. Additionally, when the air inside the tubular body is substituted with the inert gas, at the time of setting of the device, efficient substitution of the inert gas is possible by using the substituting supply line with a large amount of supply. The charged particle beam irradiation device related to the embodiment of the invention may further include a suction pump configured to suction air inside the tubular body, and the suction pump suctions the air inside the tubular body, using at least one supply line among the plurality of supply lines. According to the charged particle beam irradiation device related to the embodiment of the invention, the air inside the tubular body can be suctioned using the supply lines of the gas supply unit when the air inside the tubular body is substituted with the inert gas. Thus, it is not necessary to separately provide a line for the pump, and reduction in the number of pipes and simplification of the configuration of the device can be achieved. A preferred embodiment of the invention will be described below in detail with reference to the drawings. In addition, the terms “upstream” and “downstream” mean the upstream (cyclotron side) and downstream (patient side) of a charged particle beam to be emitted, respectively. As shown in FIG. 1, the charged particle beam irradiation device 1, which is a device used for cancer treatment through radiotherapy, or the like, includes a treatment table 2 on which a patient is put. In the charged particle beam irradiation device 1, irradiation of a charged particle beam P emitted from a cyclotron (accelerator) 3 is performed to a tumor (a body to be irradiated) of a patient on the treatment table 2. The charged particle beam P is obtained by accelerating particles with charges at high speed, for example, is a proton, a heavy particle (heavy ion) beam, or the like. The charged particle beam irradiation device 1 includes a rotating gantry 4 that is rotatable 360 degrees around the treatment table 2 with a rotation axis R as a center, an irradiation nozzle (irradiation unit) 5 that is attached to the interior of the rotating gantry 4 and is movable to arbitrary rotational positions by the rotating gantry 4, and a beam transportation line 6 that connects the cyclotron 3 and the irradiation nozzle 5 together. The beam transportation line 6 is a path along which the charged particle beam P emitted from the cyclotron 3 is transported to the irradiation nozzle 5. The beam transportation line 6 includes a vacuum duct 7 through which the charged particle beam P passes. The interior of the vacuum duct 7 is maintained in vacuum, and the charged particle beam P under transportation is prevented from being scattering due to air or the like. Additionally, the beam transportation line 6 includes a deflecting magnet 8 that deflects the charged particle beam P along the vacuum duct 7, and a converging magnet 9 that converges the beam diameter of the charged particle beam P under transportation. The irradiation nozzle 5 irradiates a diseased part of the patient on the treatment table 2 with the charged particle beam P. The irradiation nozzle 5 is detachably configured with respect to the rotating gantry 4. The irradiation nozzle 5 has an extension duct (tubular body) 10 connected to the vacuum duct 7 of the beam transportation line 6, and a quadrapole magnet 11 and a scanning magnet 12 arranged around the extension duct 10. The quadrapole magnet 11 is provided to converge the beam diameter of the charged particle beam P, which has entered the extension duct 10, with a magnetic field. The scanning magnet 12 is provided to scan the charged particle beam P that has entered the extension duct 10 and handles the charged particle beam as a scanning beam. As shown in FIGS. 1 and 2, the extension duct 10 is a hollow member arranged on a propagation path of the charged particle beam P within the irradiation nozzle 5. Helium gas (inert gas) is filled into the extension duct 10. The extension duct 10 is constituted by an inlet portion 10a through which the charged particle beam P enters from the vacuum duct 7, a central portion 10b that is scanned with the charged particle beam P by the scanning magnet 12, and an outlet portion 10c through which the charged particle beam P is irradiated toward the patient. The inlet portion 10a, the central portion 10b, and the outlet portion 10c are cylindrical portions, respectively, and the diameters (thicknesses) thereof become larger in order of the inlet portion 10a, the central portion 10b, and the outlet portion 10c. Since it is necessary to secure a scanning space for the scanning beam more widely as the extension duct 10 is closer to the patient, internal spaces are more widely formed in order of the inlet portion 10a, the central portion 10b, and the outlet portion 10c. Kapton films (particle beam transmission films) (kapton is a registered trademark) 13 and 14 for trapping the helium gas within the extension duct 10 are arranged at the inlet of the inlet portion 10a, and the outlet of the outlet portion 10c. The kapton films 13 and 14 have properties of isolating the helium gas within the extension duct 10 from the external atmosphere and transmitting the charged particle beam P therethrough. Although the kapton films 13 and 14 can transmit the charged particle beam P therethrough without attenuating the charged particle beam due to their thinness, since the strength thereof decreases based on thinness, there is a possibility that the films may be damaged if the difference between the internal and external pressures of the extension duct 10 becomes large. The possibility of the damage occurring is particularly large in the kapton film 14 on the outlet side with a large cross-sectional diameter. Additionally, since a small amount of helium gas leaks through the kapton films 13 and 14 from the extension duct 10, it is necessary to supply the helium gas into the extension duct 10. The charged particle beam irradiation device 1 has a gas supply unit 20 that supplies the helium gas into the extension duct 10. The gas supply unit 20 is constituted by a helium gas container 22 that is installed at a deck 21 outside the rotating gantry 4, and a supply pipe 23 that connects the helium gas container 22 and the extension duct 10 together. The deck 21 is provided independently from the rotating gantry 4, and the position of the helium gas container 22 is constant irrespective of the rotation of the rotating gantry 4. In addition, the helium gas container 22 may be fixed to the rotating gantry. The supply pipe 23 is constituted by a gas-container-side pipe 24, a duct-side pipe 25, and first to third pipes 26 to 28. The gas-container-side pipe 24 is connected to the helium gas container 22, and helium gas is introduced from the helium gas container 22. The duct-side pipe 25 is connected to the outlet portion 10c of the extension duct 10, and the introduced helium gas is supplied to the extension duct 10. The first pipe 26, the second pipe 27, and the third pipe 28 are pipes that are provided in parallel between the gas-container-side pipe 24 and the duct-side pipe 25. One ends of the first to third pipes 26 to 28 branch in three ways with a three-way valve 32 provided at the end of the gas-container-side pipe 24 as a starting point, and the other ends of the first to third pipes 26 to 28 are collectively connected to the duct-side pipe 25. Pipes into which the helium gas flows is switched by operating the three-way valve 32. Additionally, the first to third pipes 26 to 28 include valves 29, 30, and 31 which determine the amounts of gas supply, respectively. The gas supply unit 20 has a pressure-maintaining supply line A, an adjusting supply line B, and a substituting supply line C. Here, the supply lines mean not pipes themselves but flow channels through which the helium gas flows. The pressure-maintaining supply line A is a supply line for maintaining the pressure within the extension duct 10 at a predetermined value at a normal time. As this predetermined value, for example, a pressure slightly higher than the atmospheric pressure is selected. The pressure-maintaining supply line A is constituted by the gas-container-side pipe 24, the duct-side pipe 25, and the first pipe 26. In the pressure-maintaining supply line A, the helium gas of the helium gas container 22 moves in order of the gas-container-side pipe 24, the first pipe 26, and the duct-side pipe 25, and is supplied into the extension duct 10. In addition, the gas-container-side pipe 24 and the duct-side pipe 25 are pipes that are common in the respective supply lines. The amount of supply of the helium gas by the pressure-maintaining supply line A is set to such an amount (for example, 0.5 L/min) that the pressure within the extension duct 10 can be maintained at a pressure slightly higher than the atmospheric pressure. This amount of supply is fixed to a constant value by the valve 29 of the first pipe 26. The adjusting supply line B is a supply line used at the time of when the pressure adjustment of the extension duct 10, such as maintenance. The adjusting supply line B is constituted by the gas-container-side pipe 24, the duct-side pipe 25, and the second pipe 27. In the adjusting supply line B, the helium gas to the helium gas container 22 moves in order of the gas-container-side pipe 24, the second pipe 27, and the duct-side pipe 25 and is supplied into the extension duct 10. The amount of supply of the helium gas in the adjusting supply line B is set to a larger value (for example, 2.0 L/min) than the amount of gas supply of the pressure-maintaining supply line A. This amount of supply is fixed to a constant value by the valve 30 of the second pipe 27. The substituting supply line C is a supply line used when air within the extension duct 10 is substituted with the helium gas. The substituting supply line C is constituted by the gas-container-side pipe 24, the duct-side pipe 25, and the third pipe 28. In the substituting supply line C, the helium gas to the helium gas container 22 moves in order of the gas-container-side pipe 24, the third pipe 28, and the duct-side pipe 25 and is supplied into the extension duct 10. The amount of supply of the helium gas in the substituting supply line C is set to a larger value (for example, 10 L/min) than the adjusting supply line B. This amount of supply is fixed to a constant value by the valve 31 of the third pipe 28. The amount of supply of the helium gas in the substituting supply line C is preferably equal to or higher than 10 times of the amount of supply of the pressure-maintaining supply line A. In the pressure-maintaining supply line A, the adjusting supply line B, and the substituting supply line C, only one line selected by switching of the three-way valve 32 is used. In addition, the arrangement or configuration of the pressure-maintaining supply line A, the adjusting supply line B, and the substituting supply line C are not limited to that shown in FIG. 2. The charged particle beam irradiation device 1 includes a suction pump 33 for suctioning the air within the extension duct 10. The suction pump 33 is used when the air within the extension duct 10 is substituted with the helium gas. A suction pipe 34 is connected to a suction port of the suction pump 33, and the suction pipe 34 is connected to a switching valve 35 provided in the middle of the gas-container-side pipe 24. By switching the switching valve 35, a flow channel connected to the helium gas container 22 is closed and a flow channel connected to the suction pump 33 is opened. By driving the suction pump 33 in this state, suction of the air within the extension duct 10 is performed. The suction pump 33 performs suction of the air within the extension duct 10, using any of the pressure-maintaining supply line A, the adjusting supply line B, and the substituting supply line C. Additionally, the charged particle beam irradiation device 1 includes a leak valve 36 that leaks the helium gas to the exterior of the extension duct 10 when the internal pressure is equal to or higher than a set pressure. The leak valve 36 is provided at the outlet portion 10c of the extension duct 10. The set pressure of the leak valve 36 is preferably set within a range of 1.5 times to 2 times the atmospheric pressure. By providing such a leak valve 36, it is possible to prevent a situation where the helium gas is superfluously supplied into the extension duct 10 due to the difference between the internal and external pressures of the extension duct 10 becomes large and the kapton films 13 and 14 are torn off. Additionally, the charged particle beam irradiation device 1 includes a pressure gauge 37 that displays the internal pressure of the extension duct 10. The pressure gauge 37 is attached to a side surface of the outlet portion 10c of the extension duct 10. A meter display unit of the pressure gauge 37 protrudes from the irradiation nozzle 5 and is arranged at a position that is easily seen from a treatment room within the rotating gantry 4. The meter display unit of the pressure gauge 37 shows a value higher than the atmospheric pressure through the supply of gas from the pressure-maintaining supply line A at a normal time. If abnormality, such as gas leak, occurs in the extension duct 10, the meter display unit of the pressure gauge 37 drops to the atmospheric pressure, and thereby, the abnormality of the extension duct 10 is detected. In addition, various internal pressure sensors may be adopted instead of the pressure gauge. According to the charged particle beam irradiation device 1 related to the present embodiment described above, since it is not essentially necessary to change the amount of supply of the helium gas to the extension duct 10 in normal use, the amount of gas supply to the extension duct 10 can be easily managed by adopting a configuration in which a line with a suitable amount of supply is selected from the three supply lines A to C based on a situation. Additionally, since the helium gas is leaked through the leak valve when the internal pressure of the extension duct 10 rises excessively, it is easy to manage the internal pressure of the extension duct 10 within a desired range. Moreover, according to the charged particle beam irradiation device 1, compared to the configuration of the related art in which the amount of supply is always controlled, it is not necessary to provide an actuator or the like that drives a complicated control device or valve, and the amount of supply of the helium gas to the extension duct 10 can be managed with an extremely simple configuration. Additionally, since a complicated control device is not used, a control system trouble can be avoided, highly reliable management of the amount of gas supply can be realized, and the manufacturing costs of the device can be reduced from simplification. Additionally, in the charged particle beam irradiation device 1, management of the amount of gas supply based on a situation can be realized by performing gas supply by the pressure-maintaining supply line A at a normal time and using the adjusting supply line B with a slightly larger amount of supply at the time of adjustment, such as maintenance. Additionally, when the air within the extension duct 10 is substituted with the helium gas, such as at the time of setting of the device, efficient substitution of the helium gas is allowed by using the substituting supply line C with a large amount of supply. Moreover, according to the charged particle beam irradiation device 1, air suction can be performed using the supply lines A to C for the helium gas when the air within the extension duct 10 is suctioned by the suction pump 33. Thus, it is not necessary to separately provide a line for the pump, and reduction in the number of pipes and simplification of the configuration of the device can be achieved. The invention is not limited to the aforementioned embodiment. For example, not the rotating gantry 4 that rotates 360 degrees but a gantry that can only oscillate less than 360 degrees (for example, 180 degrees) may be used for the charged particle beam irradiation device 1. Additionally, instead of the rotation irradiation using the rotating gantry, stationary irradiation in which the rotating gantry is not used and the irradiation nozzle is fixed may be used. The invention can also be effectively applied to the stationary irradiation. Additionally, the accelerator may be a synchrotron, a synchrocyclotron, a linear accelerator, or the like. Additionally, although the management of the amount of gas supply in the charged particle beam irradiation device 1 can mainly be manually performed, an aspect in which the management is controlled automatically, such as switching a supply line based on the difference between the internal and external pressures of the extension duct 10, can be adopted. Additionally, films that partition off the interior and exterior of the extension duct 10 are not limited to the kapton films 13 and 14. Moreover, the films or sheets that are thin to a degree where the strength of the charged particle beam P to pass is not affected may be used, and can secure the airtightness within the extension duct 10. Moreover, the gas filled into the extension duct 10 is not limited to the helium gas, and rare gas and other suitable inert gas may be used. The invention is available for the charged particle beam irradiation device that can manage the amount of gas supply to a tubular body with a simple configuration. It should be understood that the invention is not limited to the above-described embodiment, but may be modified into various forms on the basis of the spirit of the invention. Additionally, the modifications are included in the scope of the invention.
	description	1. Field The present disclosure relates to an iodine absorbent material containing a salt, and a radioactive iodine removal system using the same, and more particularly, to an iodine absorbent material applicable when volatile fission products are abnormally emitted from nuclear facilities, and a radioactive iodine removal system using the same. 2. Discussion of Related Technology Radioactive iodine (131I) is one of fission products formed as a fissile material undergoes nuclear fission, which is a radioactive isotope that accounts for approximately 2.8% of a total of fission products formed as uranium undergoes nuclear fission by means of thermal neutrons (fission yield: 2.8%). The molecular iodine of this element has a high level of radioactivity and exists as gas, and thus, the absorptiveness in human body by respiration is high, therefore is a nuclide whose initial spread is of particular concern when a nuclear accident takes place. In general, iodine is present in the form of cesium iodide (CsI) in a spent nuclear fuel, and released to the outside in the CsI form. Since there is a high possibility that the first exposed environment would be the inner part of a coolant, a method of suppressing an oxidation reaction of iodide (I−) into iodine (I2). The released iodide (I−) ion is stably dissolved in the coolant, but the molecular iodine (I2) is volatile. In this background, a method of controlling a pH value of a coolant and redox conditions has been used for a long period of time. A level of volatility of the iodine present in an aqueous solution may be lowered by controlling the pH value of the coolant and the redox conditions as described above. However, the iodine is released into an internal atmosphere of a reactor containment building due to an increase in temperature of the coolant, or an increase in a level of volatility of the iodine under an environment exposed to radiation. Also, since the iodine easily reacts with organic compounds and increase a level of volatility while forming organic iodide, methods of minimizing contact with an organic compound have been used. In general, a method of removing iodine released in the atmosphere includes a method using an absorbent material. A representative absorbent material includes a carbon-based absorbent material such as activated carbon (i.e., charcoal), and a silver. J. G. Jolley and H. G. Tompkins from the US Idaho Falls Institute evaluated a level of absorption of various organic iodines at room temperature and a high-temperature desorption characteristic using silver zeolite. Also, H. Faghihian et al. (J. Radioanalytical and Nuclear Chemistry, (2002) 254: 545-550) disclosed that two kinds of natural zeolites (clinoptilolite and natrolite) may be used as a radioactive iodine absorbent by measuring a level of absorption of radioactive iodine into both of the natural zeolites. Japanese Patent Application Publication No. 2003-302493 (Oct. 24, 2003) discloses a method of fixing radioactive iodine gas using a silicate-based mineral having micropores, and Japanese Patent Application Publication No. Hei 5-126995 (May 25, 1993) discloses a method of separating an iodine species by precipitation by treating waste water containing radioactive iodine with silver nitrate. Also, Korean Patent Application Publication No. 2000-0008867 (Feb. 15, 2000) discloses a method of converting a chemical species of radioactive iodine in an aqueous solution or waste water into molecular iodine (I2) to remove the chemical species from a gaseous phase to activated carbon. As one of representative absorbent materials, a silver-based radioactive iodine absorbent material has two limits. First, since silver belongs to the group of noble metals, an absorbent material using silver is very expensive compared with its effects. Second, silver has high reactivity with iodine, but also exhibits high reactivity with chloride ions. Since many nuclear power plants are generally installed on the seashore, and salt aerosol is present at a high concentration in the atmosphere within several kilometers from the seashore. Accordingly, absorption performance of the silver-based absorbent material in a seashore region may be degraded. Also, performance of the carbon-based absorbent material may be degraded when other materials such as moisture are present in the atmosphere. The above-described methods known in the related art are effective when iodine is present in a high concentration in a closed space. However, once the iodine is released into the atmosphere and is mixed with a large amount of moisture and salt, decontamination efficiency is significantly degraded using the current capture method. The foregoing discussion is to provide general background information, and does not constitute an admission of the prior art. One aspect of the present invention is directed to a material capable of effectively removing radioactive iodine, which is harmful to human bodies, from the atmosphere, and a system for removing radioactive iodine using the same. Absorption reactivity of a carbon-based absorbent material widely used these days is degraded when there is a lot of moisture in the air, and iodine absorption performance of a silver-based absorbent material may be degraded by chloride ions present in marine environments, and is very expensive compared to its effects, but when volatile radioactive nuclide absorbent material according to the present invention is used, it is capable to provide a method to remove radioactive iodine effectively at a low cost. However, the problems to be solved according to the present invention are not limited to the above-described problems, and other problems which are not disclosed herein will be made apparent from the detailed description provided below by those skilled in the art. One aspect of the invention provides an apparatus for filtering airborne radioactive iodine. The apparatus may include: a housing defining an interior space and comprising an inlet for receiving air and an outlet for discharging the air; and a composition placed in the interior space for trapping airborne radioactive iodine between the inlet and the outlet, the composition comprising one or more salts selected from the group consisting of an alkali metal chloride and an alkaline earth metal chloride. In the foregoing apparatus, the one or more salts may be present in the form of particles. The particles may have a diameter of about 0.1 μm to about 5,000 μm. The particles may be porous. The apparatus may further comprise a filter located at each of the inlet and the outlet and configured to allow the air to pass therethrough and inhibit the particles from passing therethrough. Still in the foregoing apparatus, the composition may comprise a solution of the one or more salts. The housing may comprise a liquid drain configured to drain an aqueous solution containing the trapped iodine. The apparatus may further comprise a heater configured to heat the one or more salts. The composition may further comprise one or more bases. Another aspect of the invention provides an apparatus for filtering airborne radioactive iodine. The apparatus may include: a housing defining an interior space and comprising an inlet for receiving air and an outlet for discharging the air; and a composition placed in the interior space for trapping airborne radioactive iodine between the inlet and the outlet, the composition comprising a mixture of one or more salts selected from the group consisting of an alkali metal chloride and an alkaline earth metal chloride and one or more bases. In the foregoing apparatus, the one or more salts may be present in the form of particles having a diameter of about 0.1 μm to about 5,000 μm. The one or more salts and the one or more bases may be substantially homogeneously mixed. The one or more bases are selected from the group consisting of NaOH, KOH, NH4OH, Ca(OH)2 and Na3PO4. The housing may comprise a liquid drain configured to drain an aqueous solution containing the trapped iodine. Still in the foregoing apparatus, the apparatus may be located at a nuclear facility selected from the group consisting of a nuclear power plant, a nuclear fuel processing plant, a nuclear waste processing plant, and a radioactive material waste storage. The apparatus may further comprise a blower configured to blow the air to flow from the inlet to the outlet. The apparatus may further comprise a plurality of containers, each of which contains the composition. Still another aspect of the invention provides a method of filtering airborne radioactive iodine. The method may include: providing the foregoing apparatus at a nuclear facility selected from the group consisting of a nuclear power plant, a nuclear fuel processing plant, a nuclear waste processing plant, and a radioactive material waste storage; blowing air to flow from the inlet to the outlet and contact the composition placed in the interior space, whereby airborne radioactive iodine is trapped in the interior space. In the foregoing method, at least part of the trapped iodine may be present in an aqueous solution, wherein the method may further comprise draining the aqueous solution containing iodine from the housing. The method may further comprise heating the one or more salts so as to remove moisture from the one or more salts. Yet another aspect of the present invention provides a volatile radioactive nuclide absorbent material including a salt which is a chloride of an alkali metal or an alkaline earth metal. According to one exemplary embodiment of the present invention, the volatile radioactive nuclide absorbent material may be used to absorb a radioactive material abnormally leaked from a nuclear facility. According to another exemplary embodiment of the present invention, the volatile radioactive nuclide absorbent material may absorb a volatile radioactive nuclide present in a gaseous or aerosol phase. According to still another exemplary embodiment of the present invention, the radioactive nuclide may be iodine. According to another exemplary embodiment of the present invention, the alkali metal may be selected from the group consisting of lithium, sodium, potassium, rubidium, cesium, and francium. According to still another exemplary embodiment of the present invention, the alkaline earth metal may be selected from the group consisting of calcium, strontium, barium, radium, beryllium, and magnesium. According to still another exemplary embodiment of the present invention, the salt may be selected from the group consisting of sodium chloride, calcium chloride, or potassium chloride, and may be in the form of particles having a diameter of about 0.1 μm to about 5,000 μm. According to yet another exemplary embodiment of the present invention, the volatile radioactive nuclide absorbent material may include a pH control compound. A further aspect of the present invention provides a system for removing radioactive iodine. Here, the system includes an absorption reactor isolated from an external atmosphere and carrying the volatile radioactive nuclide absorbent material defined in claim 1, an inlet configured to allow an external atmosphere to flow into the absorption reactor, an outlet configured to discharge the external atmosphere passed through the absorption reactor to the outside, a filter installed in the front of the outlet to prevent external leakage of a salt included in the absorbent material, a drain configured to release moisture formed in the absorption reactor to the outside, and a heater configured to induce recycling of the salt and release of the radioactive iodine. Hereinafter, embodiments of the present invention will be described in detail. However, the present invention is not limited to the embodiments disclosed below, but can be implemented in various forms. The following embodiments are described in order to enable those of ordinary skill in the art to embody and practice the present invention. The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of exemplary embodiments. The singular forms “a,” “an” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises,” “comprising,” “includes” and/or “including,” when used herein, specify the presence of stated features, integers, steps, operations, elements, components and/or groups thereof, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components and/or groups thereof. With reference to the appended drawings, exemplary embodiments of the present invention will be described in detail below. Embodiments of the present invention provide a volatile radioactive nuclide absorbent material containing a salt which is a chloride of an alkali metal or an alkaline earth metal which is effective for removing gaseous radioactive iodine, and a radioactive iodine removal system using the same. The absorption performance of the carbon-based absorbent material, which is a current representative radioactive iodine absorbent material, is degraded when there is moisture in the air, and in the case of silver-based absorbent material, the problem not only lies in the absorption performance that gets degraded by iodine, but also in the high cost. Based on the facts described above, there is an urgent demand for developing a new iodine absorbent material that may be effectively used when a large amount air contaminated by radioactive iodine is released from a nuclear facility into the atmosphere, and a radioactive iodine removal system using the same. Therefore, the present inventors have completed the invention by studying on an inexpensive volatile radioactive nuclide absorbent material that can be used to effectively remove radioactive iodine that is harmful to human bodies, in the atmosphere. The present inventors have conducted research on a method capable of effectively absorbing a large amount of radioactive iodine, and found a radioactive iodine absorbent material containing a salt. As a result, the present inventors have developed a removal system capable of effectively capturing the radioactive iodine using the absorbent material. When a radioactive material is abnormally leaked as cooling of a nuclear fuel in a nuclear facility is suspended, radioactive iodine may be released into the atmosphere due to its strong volatility. In this case, the volatile radioactive nuclide absorbent material according to one exemplary embodiment of the present invention may not only selectively absorb iodine in the atmosphere since the volatile radioactive nuclide absorbent material contains a salt which is a chloride of an alkali metal or an alkaline earth metal, but may also selectively absorb iodine even when the absorbent material is already combined with steam, water aerosol, etc. Therefore, the volatile radioactive nuclide absorbent material according to embodiments of the present invention, which contains a salt which is a chloride of an alkali metal or an alkaline earth metal chloride, may be used in a radioactive iodine removal system capable of effectively absorbing a volatile radioactive nuclide present in a gaseous or aerosol phase. According to one exemplary embodiment of the present invention, the volatile radioactive nuclide absorbent material containing the salt which is the chloride of the alkali metal or the alkaline earth metal chloride possesses a function of absorbing radioactive iodine, and the salt may include at least one selected from the group consisting of chlorides of an alkali metal and/or an alkaline earth metal. In embodiments, the alkali metal may be selected from the group consisting of lithium (Li), sodium (Na), potassium (K), rubidium (Rb), cesium (Cs), and francium (Fr), and the alkaline earth metal may be selected from the group consisting of calcium (Ca), strontium (Sr), barium (Ba), radium (Ra), beryllium (Be), and magnesium (Mg), but the present invention is not particularly limited thereto. In one embodiment, the salt is sodium chloride, calcium chloride, or potassium chloride. The iodine absorbent material containing a salt according to embodiments of the present invention may improve absorption performance by increasing the reactive surface using the salt in the form of microparticles. However, the salt is a representative chemical species causing corrosion of metals, and to reduce mobility and dispersibility of the salt, a diameter of 0.1 micrometers (μm) or more that can be controlled through a conventional filter is required. In embodiments, a salt in the form of particles having a diameter of about 0.1 μm to about 5,000 μm, but not particularly limited as long as the diameter can be controlled through the filter. Meanwhile, iodine is present in the form of molecular iodine (I2) having strong volatility rather than in the form of iodide ions (I−) having high solution stability under a low pH condition, and when under the basic condition, it is present in the form of iodide (I−) or iodate (IO3−) ions having high solution stability, as shown in the following Scheme 1. Therefore, a pH control compound capable of maintaining a basic pH value is added to the iodine absorbent material containing a salt according to embodiments of the present invention to effectively inhibit the absorbed iodine from being desorbed and released again. Scheme 1I2+2OH−→I−+IO−+H2O (K=30)3IO−→2I−+IO3− (K=1020) Accordingly, to improve retention stability of the iodine absorbed onto the volatile radioactive nuclide absorbent material containing a salt which is a chloride of an alkali metal or an alkaline earth metal, the apparatus in accordance with embodiments of the present invention may further include a pH control compound such as a basic material. A group of pH control compounds includes hydroxide salts, phosphates, etc., and representative pH control compounds include NaOH, KOH, NH4OH, Ca(OH)2, Na3PO4, etc., but the present invention is not particularly limited thereto. The volatile radioactive nuclide absorbent material containing a salt according to one exemplary embodiment of the present invention may absorb iodine in the atmosphere using a salt (i.e., a chloride) that can be easily manufactured at a low cost, and may also be reused since the performance is maintained for a long period of time when a device capable of removing moisture by an increase in temperature is simply installed. Therefore, one embodiment of the present invention provides a system for removing radioactive iodine, which includes an absorption reactor isolated in an external atmosphere, and carrying the volatile radioactive nuclide absorbent material according to embodiments of the present invention, an inlet configured to allow an external atmosphere to flow into the absorption reactor, an outlet configured to discharge the external atmosphere passed through the absorption reactor to the outside, a filter installed in the front of the outlet to prevent external leakage of a salt included in the absorbent material, a drain configured to release moisture formed in the absorption reactor to the outside, and a heater configured to induce recycling of the salt and release of the radioactive iodine. A schematic view of the system is shown in FIG. 1. However, systems for removing radioactive iodine are not particularly limited as long as they include the volatile radioactive nuclide absorbent material according to embodiments of the present invention. According to embodiments of the present invention, the volatile radioactive nuclide absorbent material containing a salt or the radioactive iodine removal system using it, may be installed in a nuclear power plants that have radioactive iodine leakage possibility or in an interim storage facility of spent nuclear fuel and be properly used as a main or auxiliary means that collect and remove the radioactive iodine of the inside facility. Hereinafter, Examples are provided to aid in understanding the present invention. However, it should be understood that detailed description provided herein is merely intended to provide a better understanding of the present invention, but is not intended to limit the scope of the present invention. To determine absorption performance of an absorbent material containing a salt, an experimental device for removing iodine was manufactured, and an image of the experimental device is shown in FIG. 2. As shown in FIG. 2, a first reactor, which is an iodine generator, is filled with an aqueous solution in which molecular iodine (I2) is dissolved. A tubular absorption reactor is charged with sodium chloride (NaCl) particles. Then, a second reactor configured to dissolve and capture iodine contained in a carrier gas is filled with a sodium hydroxide (NaOH) solution. The first reactor and the absorption reactor, and the absorption reactor and the second reactor are each connected with glass tubes, respectively, to manufacture a radioactive iodine removal experimental device. To determine absorption performance of an absorbent material containing a salt, iodine absorption performance of two salts, sodium chloride and calcium chloride, was experimentally measured. For iodine generation, molecular iodine (I2) was dissolved in an aqueous solution, and added to the first reactor manufactured in Example 1. Carrier air was bubbled into the first reactor at a flow rate of 2 cc per second and evaporated iodine molecules, and passed through an absorption reactor (diameter: 2 cm, and length: 46 cm) filled with a salt. Thereafter, the carrier air passed through the absorption reactor was again passed through the second reactor filed with 200 milliliters (mL) of a 0.1 M NaOH aqueous solution and dissolved the iodine contained in the carrier air in a sodium hydroxide solution. Then, an iodine content dissolved in the sodium hydroxide solution was measured using a UV/VIS spectrometer, and calculated the amount of iodine absorbed onto the absorption reactor. The absorption reactors were charged respectively with 195 grams (g) of sodium chloride (NaCl) having an average particle diameter of 0.5 millimeters (mm), and 90 g of calcium chloride (CaCl2) having an average particle diameter of 0.1 mm to perform two absorption experiments. The absorption experiment results on both of the chlorides are listed in the following Table 1. TABLE 1Amount ofAmount ofDecontam-Absorbentinjectedpassedinationmaterialiodine gas (mg)iodine (mg)factorNaCl particles0.423Not detected∞CaCl2 particles0.134Not detected∞*Not detected: A detection limit is 0.012 mg as measured using a UV/VIS spectroscopic method As listed in Table 1, when a certain amount of iodine was passed through the absorption reactor together with air, the iodine species was completely absorbed to both of the absorbent materials, and no iodine species was detected in the carrier air passed through the absorption reactor. From these results, it was confirmed that by using alkali metal such as sodium chloride, or an alkaline earth metal chloride such as calcium chloride, radioactive iodine in the atmosphere can be easily collected. Moreover, it is easily inferred that when a reaction cross section is increased using porous absorbent material or form the absorption reactor in multi-step, an iodine absorption capacity can be further increased. A liquid may be formed in an environment in which an absorbent acts due to a hygroscopic characteristic of a salt under a high moisture condition. To evaluate an absorption characteristic of iodine under a high humidity condition, solubility of iodine was measured in a solution in which the salt was dissolved. FIG. 3 is an image illustrating solubility phenomenon comparison after 1 minute since the solid iodine (I2) was added into each distilled water and 5 M sodium chloride solution. Dissolution of iodine was not initiated in the distilled water (FIG. 3A), but the iodine was rapidly dissolved in the sodium chloride solution (FIG. 3B), which was easily observed from a change in color. Also, when the iodine was dissolved in 100 mL of a solution of 5 M sodium chloride at 25° C., it was observed that the solubility was increased to 0.13 g, which was approximately 4.5 times higher than the solubility (0.029 g) in distilled water, as listed in Table 2. This indicates that the iodine forms a stable chemical species in a sodium chloride solution. Therefore, it could be identified that not only a salt in a solid phase, but also a solution in which the salt is dissolved can stably capture iodine. TABLE 2Amount ofReportedTest solutiondissolved iodine atsolubility at(100 ml)25° C. (g)25° C. (g/100 mL)Distilled water0.0320.0295M NaCl0.130— From the results, it could also be identified that a radioactive iodine removal system as shown in FIG. 1 can be manufactured using the absorbent material containing a salt according to embodiments of the present invention. When a filter is installed at an inlet and an outlet of an absorption reactor to prevent an absorbent material including a salt from flowing out, and when a large amount of moisture is included in the atmosphere inflow, the iodine dissolved in moisture could be collected in the form of a solution from a lower portion of the absorption reactor. In this case, the iodine dissolved in an aqueous solution is ionized into a soluble chemical species by chloride ions included in the solution, and can be stably dissolved in the solution, and additionally adds a pH control agent in the absorbent material to be more stably presented in the aqueous solution. In addition, since the radioactive iodine has a short half-life of 8 days, the iodine absorbent material containing a salt could be repeatedly reused when several weeks pass after saturation, and the moisture is removed, and after the absorbent material is dried using a built-in heater. Moreover, a gamma-ray spectrometer can be further installed inside or outside of the absorption reactor to drain moisture collected at a lower portion of the absorption reactor, or sense the period for reuse. According to embodiments of the present invention, the volatile radioactive nuclide absorbent material containing a salt or the radioactive iodine removal system using it, has advantages of absorption functions to selectively absorb iodine in the atmosphere, and is inexpensive compared to a silver-based absorbent material. The salt can selectively absorb gaseous iodine, and even when an iodine species is present in steam or water aerosol phases, due to the hygroscopicity of the salt, an aerosol can be effectively captured, and due to this characteristic, when the radioactive iodine is present in a gaseous phase, it can be absorbed and the salt can be properly used to remove the radioactive iodine in the atmosphere. Therefore, one embodiment of the present invention is expected to be installed in a nuclear power plants that have radioactive iodine leakage possibility or in an interim storage facility of spent nuclear fuel and be vitally used as a main or auxiliary means that collect and remove the radioactive iodine of the inside facility. While embodiments of the invention have been shown and described with reference to certain exemplary embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the scope of the invention as defined by the appended claims.
	description	FIG. 1 shows a computed tomography apparatus which includes a gantry 1 on which a radiation source 2 is mounted. The X-ray detector 8 with the anti-scatter grid 3 arranged thereabove is mounted so as to face the radiation source 2. A patient 5 on a table 6 is introduced into the beam path 4. The gantry 1 rotates about the patient 5. An examination zone 7 is thus irradiated from all sides. The patient 5 is slid through the rotating gantry in the horizontal direction, so that a volume image is acquired by way of a plurality of cross-sectional images. The zone scanned during one rotation is substantially larger in the case of two-dimensional X-ray detectors 8 than in the case of single-line X-ray detectors. As a result, the patient 5 can be slid through the gantry 1 faster. The FIGS. 2 to 5 show a one-sided comb element 12 in several views. FIG. 2 is a plan view of a one-sided comb element 12. This one-sided comb element 12 is made of a material absorbing X-rays, for example brass, molybdenum, tungsten. The comb structure of the comb element 12 is formed by comb lamellae 1I1 which extend transversely of a base plate 10. The height of the comb element 12 is dependent on the specific application. A decisive criterion in this respect is the surface area irradiated by one scan. The ratio of useful radiation to scattered radiation becomes worse as the width of the surface irradiated by the X-rays per scan increases. The comb elements 12 typically have a height of from approximately 2 to 6 cm. The more scattered radiation is contained in the overall signal, the higher the anti-scatter grid must be. The width of the comb element 12, or also of the base plate 10, is governed by the width of the X-ray detector 8. An anti-scatter grid 3 as constructed from such comb elements 12 must completely cover the X-ray detector 8. In the case of large-area flat X-ray detectors, therefore, the comb elements 12 are wider than in the case of the narrower multi-line or two-dimensional X-ray detectors 8 used in computed tomography. The depth of the comb lamellae 11 and the distance D between the individual comb lamellae 11 define the pixel size of such an anti-scatter grid 3. In the case of two-dimensional X-ray detectors 8 for computed tomography apparatus the pixel size amounts to from approximately 1xc3x971 to 2xc3x975 mm2. A plurality of comb elements 12 are oriented relative to the incident X-rays in such a manner that the X-rays pass through the grid openings formed by the comb lamellae 11 and the base plate 10. X-rays are emitted by the X-ray source with a focal spot and emanate at a radiation angle from this spot. In order to achieve effective filtering or an as good as possible primary radiation transparency, the comb lamellae 11 are arranged on the base plate 11 so as to be oriented towards or focused on this focal spot. This is shown in FIG. 4. The distance Do between the comb lamellae 11 at the upper edge of the base plate 10 is smaller than the distance Du between the comb lamellae 11 at the lower edge of the base plate 10. Because the X-ray detectors 8 in computed tomography apparatus are adapted to a curvature, it is necessary to adapt the anti-scatter grid 3 accordingly. FIG. 3 shows that the depth of the comb lamellae 11 at the upper edge is less than that at the lower edge of the base plate 10. Piece-wise assembly of small anti-scatter grid segments is possible in the case of long X-ray detectors. FIG. 6 illustrates the linking of a plurality of one-sided comb elements 12. Due to the different depths of the comb lamellae 11 at the upper edge and the lower edge (FIG. 3), the anti-scatter grid 3 can be readily adapted to the curvature of the X-ray detector 8. The curvature of the anti-scatter grid 3 is also imposed by the arrangement of the grooves 14 in the frame 13. FIG. 7 illustrates the arrangement of a plurality of one-sided comb elements 12 in a frame 13 which produces an X-ray shadow. The inner side of the frame 13 is provided with grooves 14 which are shown in FIG. 8. The grooves 14 receive the sides of the base plates 10 of the plurality of one-sided comb elements 12. The comb elements 12 can be glued in or be secured in any other feasible manner. Mechanical fixation by pressing in the comb elements 12 is also feasible. An anti-scatter grid 3 is formed by linking a plurality of onesided comb elements 12. The comb lamellae 11 of one base plate 10 then adjoin the rear side of a neighboring base plate 10. The length of such an anti-scatter grid 3 can be increased at will by selection of the number of comb elements 12. A further embodiment of an anti-scatter grid 3 will be described in detail hereinafter. The FIGS. 9 to 12 show a two-sided comb element 15 and an anti-scatter grid 3 assembled from such elements and lamellae 19. FIG. 9 shows a two-sided comb element 15 with a double comb structure. It consists of a base plate 17 on both sides of which there are provided lamellae 16 and 18. The comb lamellae 16 and 18 are arranged on both sides of the base plate 17 so as to extend transversely of the comb base surface formed by the base plate 17. The above configurations for the focusing of the one-sided comb element 12 are to be used accordingly for this two-sided comb element 15. Moreover, in order to imitate the curvature of the X-ray detector 8, the comb lamellae 16 and 18 are deeper at the lower side of the base plate 17 than the comb lamellae 16 and 18 at the upper edge of the base plate 17. FIG. 11 shows the assembly of plane lamellae 19 (FIG. 10) and two-sided comb elements 15. Two-sided comb elements 15 and lamellae 19 are fitted in an alternating arrangement in a frame 13, thus forming an anti-scatter grid 3. The comb lamellae 16 and 18 adjoin the respective neighboring lamellae 19. The length of the anti-scatter grid 3 can again be increased by increasing the number of two-sided comb elements 15 and lamellae 19 used. Anti-scatter grids are used not only for computed tomography but also for radiology. In that case the anti-scatter grid 3 need not be curved, because the X-ray detector 8 is flat. Such anti-scatter grids typically have dimensions other than the grids described thus far. In these fields of application, however, fewer vibrations occur. The frames of these anti-scatter grids are larger and the comb elements 12 or 15 to be used are also larger. Because of the very high natural stability of the comb elements 15, such an embodiment of an anti-scatter grid is suitable for a very large range of applications. Several methods are available for the manufacture of such comb elements 15. Depending on the resolution or pixel size of the anti-scatter grid, the comb elements 12 or 15 can be formed, for example by means of milling, sintering or injection molding. In the case of the injection-molding method materials absorbing X-rays can be added to a basic material. An anti-scatter grid 3 can also be formed by linking two-sided comb elements 15 without arranging lamellae 19 therebetween. Instead of using a frame 13, the comb elements 12 or 15 can also be arranged while using spacers in such a manner that an anti-scatter grid is formed. Such an anti-scatter grid can be adapted to special applications by varying the distances between the comb lamellae of the comb elements. For example, it is feasible to realize a higher resolution for an inner or core area of an anti-scatter grid; this can be achieved by means of a grid with very fine meshes. The resolution could be lower at the edge area of the X-ray detector that is covered by the anti-scatter grid, so that at this area the grid openings in the anti-scatter grid may be larger.
	summary
	abstract	A hydrogen oxidation catalyst is provided, comprising a zeolite that contains at least one catalytically active noble metal or a compound thereof, wherein said zeolite is a hydrophobic zeolite. A use of the catalyst and a method for hydrogen recombination in nuclear power plants, reprocessing plants or fuel element repositories is also specified.
047325275	description	FIG. 1 gives a sectional view of an active cell comprising a horizontal slab 1 and walls 2. The plug 3 separates the active region 6 from the clean region 7. The object to be transferred 4 is suspended from the plug. The transfer will be carried out along axis 5, which is assumed to be vertical for greater clarity of the description. FIG. 2 shows a pre-enclosure cut by a vertical plane on the transfer axis. The walls 10 delimit a closed box containing two plugs, both located along the vertical transfer axis, namely: a lower plug 11 and and upper plug 12. Within said pre-enclosure, a carriage 14 can be moved horizontally by means of a control rod 15. For greater clarity, the wheels of the carriage, as well as the rails on which they roll, have been omitted from the schematic drawing. The longitudinal dimension 16 of the inside of the pre-enclosure is such that the passage along the transfer axis is clear when the carriage is located in position 17. The carriage positions 14 and 17 will be respectively referred to as the left or left-hand and right or right-hand positions throughout the following description. Sealing means 13 steady the tightness between the active cell and the pre-enclosure. The movable enclosure is schematically illustrated in FIG. 3 and can be seen to consist of a cylindrical or prism-shaped box having a wall or walls 30 parallel to transfer axis 36. The top of said enclosure is closed except for an opening for a cable 32 sealed with a bellows 34. The end of the cable is provided with a hook 33. The bottom of said enclosure includes a removable plug 31 and sealing means 35 for tightness between the enclosure and the pre-enclosure. Having thus described the physical features of the invention, the operating principle may now be explained, for example in terms of removing an object from the cell. The mutual arrangement of the components described in the foregoing is illustrated in FIG. 4 and the transfer operation is described with reference to the following figures. Items appearing in several figures will keep the same reference. Basically then, the top slab 40 of the cell comprises the plug 41 and the "object" to be transferred 42 suspended therefrom. Pre-enclosure 43 is placed on the slab in alignment with the vertical transfer axis. Plug 44 and plug 46 are in place and the carriage 45 is located to the left. The enclosure 47 is placed on the pre-enclosure in alignment with the transfer axis. Plug 48 is in place. And the hook bears the reference 49. Attention is drawn to the fact that neither the thickness of the enclosure and pre-enclosure walls, nor the sealing means between the different components are represented in the drawings, with intent. FIG. 5 shows the first step in the transfer, which consists in bringing together the plugs in pairs. Bottom plug 44 of the pre-enclosure is made fast on the slab plug 41. Plug 48 of the enclosure is made fast on the top plug 46 of the pre-enclosure and this assembled pair is placed in the carriage. A device not covered by the claims is used to lock and unlock the two plugs by means of an outside control rod (also omitted from the drawing). The purpose of these couplings is to protect from contamination the faces of the plugs which will be exposed in the clean zone. Thereafter, the carriage 45 is pulled to the right-hand position as shown in FIG. 6. FIG. 7 shows the hook 49 hooked onto the assembly made up of plugs 41 and 44 and the object 42. The bellows seal around the cable has been omitted from the drawing. Said assembly 41, 44, 42 is pulled up by the cable into the enclosure 47, as shown in FIG. 8, whereafter the carriage 45 is moved back to the left-hand position as in FIG. 9 and plug 46 is returned to block the opening at the top of the pre-enclosure as in FIG. 10. It is then possible to pick up enclosure 47 and take it elsewhere. Confinement has been maintained throughout and enclosure 47 is now closed by its plug 48 whose bottom face has never been contaminated since it has been kept in contact at all times with the equally clean top face of pre-enclosure plug 46. The said pre-enclosure effects closure of the cell thanks to the tightness of the contacts (the seals 13 of FIG. 2) and top plug 46, now back in its initial position. The devices used to couple the plugs in pairs and to lift the joined plugs are intentionally not represented and may be of any suitable known type. The seals used between the cell and the pre-enclosure and between the pre-enclosure and the enclosure may advantageously be three-way acting seals. In FIG. 11 can be seen another embodiment of the device utilized in carrying out the inventive procedure. By way of a variant, it may be envisaged to equip the movable enclosure with a bell or lining cap 50 sliding snugly with seals, to avoid contaminating the inside of the enclosure. Also, the lifting and lowering operations can be effected by means of a specialty chain with a pushing capability, such as the locking-link "rigid" chain made by SERAPID (France), housed in a tunnel attached to the enclosure. This arrangement is illustrated in FIG. 12. As the figure shows, a tunnel 60 mounted on the enclosure 47 comprises a rigid chain 61 running over two deflecting sheaves 62, 63. To simplify, the opposite sheaves and the guides for the rigid chain links have been omitted from the drawing. This device affords the advantage of operating with low head-room whilst concentrating the plug and object lifting means on the enclosure. It is particularly convenient for conveying the equipment 42 and unloading it in a maintenance cell for repair or overhaul. The foregoing description outlined the procedure according to the invention for transferring an object which can be removed vertically and upwardly. It has been indicated that the procedure can be carried out vertically downward to feed an object into a cell and it should be apparent to those knowledgeable in the art that the same procedure can be applied with slight alteration to: horizontal transfer with a plug located on a vertical wall, PA1 vertical transfer with a plug in the floor of the cell, PA1 and in general a transfer along any given axis. The present invention has been described in relation to a nuclear facility. However, it is just as applicable to the transfer of objects contaminated by chemical or bacteriological pollutants.
	claims	1. A core monitoring system comprising:a core thermal limit value monitoring device that outputs a signal to a control unit of a nuclear reactor as a result of monitoring a thermal state value of a nuclear reactor core received from a local power range monitor,the device comprising: three or more signal input processing units each configured to receive a signal output from the local power range monitor, the signal representing a state of the core that has been divided into three or more regions to be monitored, each of the signal processing units includinga calculating unit configured to calculate thermal state values of a subset of the regions to be monitored, the thermal state values being calculated based on the signal received by the signal input processing unit to determine whether signal output to the control unit is necessary;a synchronization processing unit configured to output a signal-output stop signal and a signal-output stop cancellation signal to synchronization processing units of other signal input processing units; andan output processing unit, anda control unit of a nuclear reactor, whereinif a first calculating unit of a first signal input processing unit has determined that signal output to the control unit is necessary, and a first synchronization processing unit of the first signal input processing unit has not received a signal-output stop signal from a synchronization processing unit of one of the other signal input processing units, the first synchronization processing unit is configured to transmit the signal-output stop signal to synchronization processing units of the other signal input processing units, and a first output processing unit of the first signal input processing unit is configured to output a signal to the control unit representing calculation results of the first calculating unit,if the first calculating unit has determined that signal output to the control unit is necessary, and the first synchronization processing unit has received a signal-output stop signal from a synchronization processing unit of one of the other signal input processing units, the first calculating unit is configured to not transmit the signal-output stop signal or the signal-output stop cancellation signal to the synchronization processing units of the other signal input processing units,if the first calculating unit has determined that signal output to the control unit is not necessary, the first synchronization processing unit is configured to determine whether the signal-output stop signal has been transmitted to the synchronization processing units of the other signal input processing units, and if the synchronization processing units of the other signal input processing units have received the signal-output stop signal, the first synchronization unit is configured to transmit the signal-output stop cancellation signal to the synchronization processing units of the other signal input processing units, andthe control unit is configured to control a power level of the nuclear reactor based on the signal output by the core thermal limit value monitoring device to the control unit. 2. The core monitoring system according to claim 1, whereinthe core thermal limit value monitoring device further comprises a region definition information storage unit configured to store a plurality of patterns, each pattern including a combination of region numbers in the regions to be monitored and assigned calculation processing numbers, and each signal input processing unit further comprises:a setting unit configured to assign one of the patterns stored in the region definition information storage unit to the calculating unit of the signal input processing unit;a region definition information table configured to store the assigned pattern; anda reading unit configured to output information from the region definition information table to the calculating unit of the signal input processing unit, the information identifying the subset of regions for which the calculating unit calculates thermal state values. 3. The core monitoring system according to claim 1, further comprising:a core performance calculating device configured to calculate a distribution of power in the core. 4. A core monitoring method comprising:a core thermal limit value monitoring method that uses a computer including three or more CPUs to output a signal to a control unit of a nuclear reactor as a result of monitoring a thermal state value of a nuclear reactor core, based on a local power range monitor, the core thermal limit value monitoring method comprising:a signal inputting step in which the CPUs each receive a signal output from the local power range monitor, the signal representing a state of the core that has been divided into three or more regions to be monitored;a calculating step in which the CPUs each calculate thermal state values of the regions to be monitored based on the received signal to determine whether signal output to the control unit is necessary;if a CPU has determined that signal output to the control unit is necessary, and the CPU has not received a signal-output stop signal from one of the other CPUs, the CPU performs a signal transmitting step of transmitting a signal output stop signal to the other CPUs and an outputting step of outputting a signal representing calculation results from the calculation step to the control unit;if the CPU has determined that signal output to the control unit is necessary, and the CPU has received a signal-output stop signal from one of the other CPUs, the CPU does not perform the signal transmitting step;if the CPU has determined that signal output to the control unit is not necessary, the CPU performs a determination step of determining whether the signal-output stop signal has been transmitted to the other CPUs, and if the other CPUs have received the signal-output stop signal, the CPU performs a signal cancellation step of transmitting a signal-output stop cancellation signal to the other CPUs; andcontrolling a power level of the nuclear reactor based on the signal output to the control unit by the computer including three or more CPUs.
	description	Priority is claimed as a divisional application to U.S. patent application Ser. No. 12/645,846, filed Dec. 23, 2009 and now U.S. Pat. No. 8,681,924, which claims priority as a continuation-in-part to U.S. patent application Ser. No. 12/432,509, filed Apr. 29, 2009 and now U.S. Pat. No. 8,158,962, which in turn claims priority to U.S. Provisional Patent Application Ser. No. 61/048,707, filed Apr. 29, 2008, and to U.S. Provisional Patent Application Ser. No. 61/173,463, filed Apr. 28, 2009. The aforementioned priority applications are incorporated herein by reference in their entireties. The present invention relates generally to neutron absorbing apparatus and methods used to facilitate close packing of spent nuclear fuel assemblies, and more specifically to a single-plate neutron absorber apparatus and method of manufacturing the same. In other aspects, the invention relates to methods of supporting spent nuclear fuel assemblies in a submerged environment using the single-plate neutron absorber apparatus and a fuel rack system incorporating the single-plate neutron absorber apparatus. Nuclear power plants currently store their spent fuel assemblies on site for a period after being removed from the reactor core. Such storage is typically accomplished by placing the spent fuel assemblies in closely packed fuel racks located at the bottom of on site storage pools. The storage pools provide both radiation shielding and much needed cooling for the spent fuel assemblies. Fuel racks often contain a large number of closely arranged adjacent storage cells wherein each cell is capable of accepting a spent fuel assembly. In order to avoid criticality, which can be caused by the close proximity of adjacent fuel assemblies, a neutron absorbing material is positioned within the cells so that a linear path does not exist between any two adjacent cells (and thus the fuel assemblies) without passing through the neutron absorbing material. Early fuel racks utilized a layer of neutron absorbing material attached to the cell walls of the fuel rack. However, these neutron absorbing materials have begun to deteriorate as they have been submerged in water for over a decade. In order to either extend the period over which the fuel assemblies may be stored in these fuel racks, it is necessary to either replace the neutron absorber in the cell walls or to add an additional neutron absorber to the cell or the fuel assembly. In an attempt to remedy the aforementioned problems with the deteriorating older fuel racks, the industry developed removable neutron absorbing inserts, such as the ones disclosed in U.S. Pat. No. 5,841,825 (the “'825 Patent”), to Roberts, issued Nov. 24, 1998; U.S. Pat. No. 6,741,669 (the “'699 Patent”), to Lindquist, issued May 25, 2004; and U.S. Pat. No. 6,442,227 (the “'227 Patent”), to Iacovino, Jr. et al., issued Aug. 27, 2002. As of recent times, the neutron absorbing insert has become the primary means by which adjacent fuel assemblies are shielded from one another when supported in a submerged fuel rack. Thus, newer fuel racks are generally devoid of the traditional layer of neutron absorbing material built into the structure of the fuel rack itself that can degrade over time. Instead, fuel assembly loading and unloading procedures utilizing neutron absorbing inserts have generally become standard in the industry. While the neutron absorbing inserts disclosed in the '825 Patent, the '227 Patent and the '699 Patent have proved to be preferable to the old fuel racks having the neutron absorbing material integrated into the cell walls, these neutron absorbing inserts are less than optimal for a number of reasons, including without limitation complexity of construction, the presence of multiple welds, complicated securing mechanisms, and multi-layered walls that take up excessive space within the fuel rack cells. Additionally, with existing designs of neutron absorbing inserts, the inserts themselves must be removed prior to or concurrently with the fuel assemblies in order to get the fuel assemblies out of the fuel rack. This not only complicates the handling procedure but also leaves certain cells in a potentially unprotected state. The '825 Patent discloses a neutron absorbing apparatus which includes two adjacent neutron absorbing plates and a mounting assembly with latching means configured to be secured to fuel assemblies while the fuel assemblies remain under water in a fuel storage rack. The two neutron absorbing plates of the '825 Patent are positioned orthogonally to form a chevron cross section which is placed about the fuel assemblies by insertion in the existing space between the fuel assemblies and the cell walls of a fuel storage rack. The primary embodiment of the neutron absorbing apparatus of the '825 Patent utilizes a three layer configuration consisting of a backing plate (made of aluminum or stainless steel), a neutron absorbing sheet (made of cadmium, borated stainless steel, borated aluminum, or boron in a ceramic matrix), and a cover plate (made of aluminum or stainless steel). This multi-layer embodiment is both cumbersome and difficult to manufacture. Moreover, the absence of the neutron absorbing sheet at the fold in the backing plate and at the lateral edges of the backing plate is less than optimal and provides a potential area for increased reactivity. It should be noted that the '825 Patent also discloses a second embodiment of a neutron absorbing apparatus that allegedly eliminates any loss of nuclear absorber coverage at the fold in the backing plate and at the same time simplifies the fabrication process. In this embodiment, a special single-layer backing plate made of borated aluminum or borated stainless steel is used to replace the multi-layer arrangement of the primary embodiment. This special backing plate is itself a nuclear absorber and thus no additional absorber layer is added to provide the nuclear absorption. However, for this embodiment, the '825 Patent teaches that the special backing plate must be formed by two plates arranged to form the chevron configuration and welded together at their juncture. In this regard, the '825 Patent specifically states that for this embodiment “[t]he two individual plates are necessary because the borated backing plates cannot be folded, but must [be] welded. [T]he two borated backing plates . . . are welded together along [the] seam . . . to provide the chevron formation necessary to produce [the] plates . . . of the complete invention.” For obvious reasons, welds and joints in the body of the neutron absorbing apparatus are less than optimal. Turning to the '227 Patent, a sleeve assembly for refurbishing a fuel rack having cells in which fresh or spent nuclear fuel assemblies may be stored is disclosed. The sleeve assembly of the '227 Patent has at least one elongate wall extending from the topside of a sleeve base having an opposed bottom side. The sleeve base has a flow hole extending therethrough that communicates with one of the rack base plate flow holes. A pin assembly disposed in the sleeve base flow hole has resilient tabs extending beyond the bottom side of the sleeve base for extending into a rack base plate flow hole and resiliently engaging the rack base plate when the sleeve assembly is installed in one of the cells. The pin assembly resists horizontal and vertical movements of the sleeve assembly, permits water flow into the cell and permits sleeve assembly removal tools and inspection devices to access the pin assembly. The '227 Patent discloses an embodiment of a sleeve assembly having chevron shaped walls formed by a single-plate. The '227 Patent discloses that these walls are an extruded composite of boron carbide and aluminum. The extruding process to form the chevron shaped walls is believed to be less than optimal as it is difficult to perform, yields unpredictable result, requires extremely tight tolerances and results in an inferior product. Turning now to the '669 Patent, a neutron absorber system for a nuclear fuel storage rack is disclosed that includes a neutron absorber that is adapted to attach to a plurality of cell walls of a cell of the nuclear fuel storage rack. The neutron absorber is adapted to elastically deform. Means for applying at least one stress to the neutron absorber and means for releasing the at least one stress to cause the neutron absorber to attach to the plurality of cell walls of the cell of the nuclear fuel storage rack is also disclosed. In one embodiment, the '669 Patent teaches a multi-plate longitudinal weldment to form the body of the neutron absorber system. Specifically, the '669 Patent teaches welding a metal matrix alloy corner piece to two metal matrix neutron absorber composite plates to form the chevron shape. Welds and joints in the body of the neutron absorbing apparatus are less than optimal. These welds in this embodiment render the neutron absorber system less than optimal. The '669 Patent also teaches a neutron absorber system having chevron-shaped walls that are formed of a metal composite which includes neutron absorbing material, for example, boron carbide or a metal boron alloy, such as aluminum, magnesium, titanium, aluminum/magnesium or aluminum/titanium, in combination with boron, for example. The '629 Patent also discloses that the material may be stainless steel/boron alloys and that besides boron carbide and elemental boron, any element with a high thermal neutron absorption cross section may be substituted. The '669 Patent further states generally that the first wall and the second wall of the chevron-shaped body “may be formed of a unitary material or they may be formed separately and attached to each other, for example, via standard TIG welding or by friction stir welding.” Despite this statement, the '669 Patent is devoid of any enabling teaching as to how the chevron shaped body (which consists of the two walls connected along a longitudinal edge) is formed of a unitary material of a metal composite including a neutron absorbing material. Such materials tend to be very brittle when the percentage of boron carbide becomes substantial and thus, to date, it has been generally accepted in the art that only flat plates can be satisfactorily created from such materials. The only exception being the extruding process mentioned in the '227 Patent, which as stated is less than optimal and undesirable. Therefore, the '669 Patent also fails to teach a suitable neutron absorber insert and an enabling method of manufacturing such an insert. These and other limitations of the prior art are overcome by the present invention which is described in the following detailed specifications. It is therefore an object of the present invention to provide a neutron absorbing apparatus and a fuel rack system incorporating the same for the submerged storage of fuel assemblies. Another object of the present invention is to provide a neutron absorbing apparatus having a chevron-shaped wall structure formed by bending a single plate of a metal matrix composite having neutron absorbing particulate reinforcement. Yet another object of the present invention is to provide a method of manufacturing a neutron absorbing apparatus having a chevron-shaped wall structure by bending a single plate of a metal matrix composite having neutron absorbing particulate reinforcement. Still another object of the present invention is to provide a fuel rack system incorporating a neutron absorbing apparatus having a chevron-shaped wall structure formed by bending a single plate of a metal matrix composite having neutron absorbing particulate reinforcement. A further object of the present invention is to provide a neutron absorbing apparatus for slidable insertion into a cell of a submerged fuel rack that eliminates the need for complicated mechanisms for securement to a fuel assembly. A yet further object of the present invention is to provide a neutron absorbing apparatus that can be slid into and out of a loaded cell of a submerged fuel rack without requiring removal of the fuel assembly. A still further object of the present invention is to provide a neutron absorbing apparatus having a chevron-shaped wall structure constructed of a metal matrix composite having neutron absorbing particulate reinforcement that extends the entire length of a fuel assembly. An even further object of the present invention is to provide a neutron absorbing apparatus having a chevron-shaped wall structure constructed of a metal matrix composite having neutron absorbing particulate reinforcement that extends the entire length of a fuel assembly and is adequately rigid and straight. Another object of the present invention is to provide a neutron absorbing apparatus that can be easily and repetitively slid into and out of a loaded cell of a submerged fuel rack. These and other objects are met by the present invention, which in one embodiment is a neutron absorbing apparatus comprising: a sleeve having first wall and a second wall, the first and second wall forming a chevron shape; and the first and second wall being a single panel of a metal matrix composite having neutron absorbing particulate reinforcement bent into the chevron shape along a crease. In another aspect, the invention can be a method of manufacturing a neutron absorbing apparatus comprising: a) providing a panel of a metal matrix composite having neutron absorbing particulate reinforcement; and b) bending the panel into a chevron shape having first and second walls. In yet another aspect, the invention can be a method of manufacturing a neutron absorbing apparatus comprising: a) providing a roll of boron carbide aluminum matrix composite; b) hot rolling the roll of boron carbide aluminum matrix composite; c) straightening and flattening the roll of boron carbide aluminum matrix composite using a hot roll leveler to create a panel of boron carbide aluminum matrix composite; d) shearing the panel of boron carbide aluminum matrix composite to a desired geometry; and e) bending the panel boron carbide aluminum matrix composite into a chevron shape having first and second longitudinal walls. In still another aspect, the invention can be a method of creating a useful article having neutron absorbing properties comprising: a) providing a panel of a metal matrix composite having neutron absorbing particulate reinforcement; and b) bending the panel to form a chevron shape having first and second walls. In a further aspect, the invention can be a system for supporting radioactive fuel assemblies in a submerged environment comprising: a fuel rack comprising a base plate and an array of cells; and a neutron absorbing insert slidably inserted into one or more of the cells, the neutron absorbing insert comprising a sleeve having first wall and a second wall, the first and second wall forming a chevron shape, and the first and second wall being a single panel of a metal matrix composite having neutron absorbing particulate reinforcement bent into the chevron shape. In another aspect, the invention is a neutron absorbing apparatus comprising: a plate structure having a first wall and a second wall that is non-coplanar to the first wall; the first and second walls being formed by a single panel of a metal matrix composite having neutron absorbing particulate reinforcement that is bent into the non-coplanar arrangement along a crease; and a plurality of spaced-apart holes formed into the single panel along the crease. In yet another aspect, the invention can be a system for supporting spent nuclear fuel in a submerged environment comprising: a fuel rack comprising an array of cells; a fuel assembly positioned within at least one of the cells of the fuel rack; at least one neutron absorbing insert comprising a single panel of a metal matrix composite having neutron absorbing particulate reinforcement that is bent into a chevron shape along a crease, and a plurality of spaced-apart holes formed into the single panel along the crease; and the neutron absorbing insert positioned within the cell of the fuel rack so that the sleeve is located between the fuel assembly and the walls of the fuel rack. In a still further embodiment, the invention can be a method of manufacturing a neutron absorbing apparatus comprising: a) providing a single panel of a metal matrix composite having neutron absorbing particulate reinforcement; b) forming a line of spaced-apart holes in the single panel; and c) bending the panel along the line into a chevron shape having first and second walls. In still another embodiment, the invention can be a method of manufacturing a neutron absorbing apparatus comprising: a) providing a roll of boron carbide aluminum matrix composite; b) hot rolling the roll of boron carbide aluminum matrix composite; c) straightening and flattening the roll of boron carbide aluminum matrix composite using a hot roll leveler to create a panel of boron carbide aluminum matrix composite; d) shearing the panel of boron carbide aluminum matrix composite to a desired geometry; e) forming a line of spaced-apart slits in the single panel; and f) bending the panel boron carbide aluminum matrix composite along the line into a chevron shape having first and second longitudinal walls. Referring first to FIGS. 1 and 2 concurrently, a neutron absorbing insert 100 according to one embodiment of the present invention is illustrated. The neutron absorbing insert 100 and the inventive concepts explained herein can be used in conjunction with both PWR or BWR storage requirements. The neutron absorbing insert 100 is specifically designed to be slidably inserted at strategic locations within the cell array of a submerged fuel rack. However, in some embodiments, it is to be understood that the inventive neutron absorbing insert can be used in any environment (and in conjunction with any other equipment) where neutron absorption is desirable. Furthermore, in embodiments where the invention is based solely on the method of bending a metal matrix composite having neutron absorbing particulate reinforcement (or the resulting angled plate structure), the invention can be used in any environment and/or used to create a wide variety of structures, including without limitation fuel baskets, fuel racks, sleeves, fuels tubes, housing structures, etc. The neutron absorbing insert 100 generally comprises a reinforcement assembly 120 fastened to the top end of the sleeve 110. The sleeve 110 is chevron-shaped and constructed of a boron carbide aluminum matrix composite material. However, other metal matrix composites having neutron absorbing particulate reinforcement can be used. Examples of such materials include without limitation stainless steel boron carbide metal matrix composite. Of course, other metals, neutron absorbing particulate and combinations thereof can be used including without limitation titanium (metal) and carborundum (neutron absorbing particulate). Suitable aluminum boron carbide metal matrix composites are sold under the name Metamic® and Boralyn®. The boron carbide aluminum matrix composite material of which the sleeve 110 is constructed comprises a sufficient amount of boron carbide so that the sleeve 110 can effectively absorb neutron radiation emitted from a spent fuel assembly and thereby shield adjacent spent fuel assemblies in a fuel rack from one another. In one embodiment, the sleeve 110 is constructed of an aluminum boron carbide metal matrix composite material that is 20% to 40% by volume boron carbide. Of course, the invention is not so limited and other percentages may be used. The exact percentage of neutron absorbing particulate reinforcement required to be in the metal matrix composite material will depend on a number of factors, including the thickness (i.e., gauge) of the sleeve 110, the spacing between adjacent cells within the fuel rack, and the radiation levels of the spent fuel assemblies. However, as space concerns within the fuel pond increase, it has become desirable that the sleeve 110 take up as little room as possible in the cell of the fuel rack. Thus, the sleeve 110 is preferably constructed of an aluminum boron carbide metal matrix composite material having a percentage of boron carbide greater than 25%. While the addition of boron carbide particles to the aluminum matrix alloy increases the ultimate tensile strength, increases yield strength, and dramatically improves the modulus of elasticity (stiffness) of the material, it also results in a decrease in the ductility and fracture toughness of the material compared to monolithic aluminum alloys. Prior to the current inventive manufacturing process, these properties have limited the ways in which metal matrix composites having neutron absorbing particulate reinforcement could be used, thereby leading to difficulty in fabrication of the material into usable products. However, as will be described in greater detail below, the current invention has made it possible to bend sheets of boron carbide aluminum matrix composite material (and other metal matrix composites having neutron absorbing particulate reinforcement). Thus, the walls 111, 112 of the sleeve 110 are formed into the chevron shape by bending a single sheet of boron carbide aluminum matrix composite material in an approximate 90 degree angle along its length. Of course, other angles can be achieved. This inventive process will be described in greater detail below with respect to FIGS. 9-11. Referring still to FIGS. 1 and 2, the sleeve 110 has a first longitudinal wall 111 and a second longitudinal wall 112. The first longitudinal wall 111 is integral with and joined to the longitudinal second wall 112 along crease 113. The first longitudinal wall 11 and the second longitudinal wall 112 form a chevron shaped structure (viewed from the top or bottom). The chevron shape formed by the first longitudinal wall 111 and the second longitudinal wall 112 has an approximately 90 degree angle. Of course, other angles are contemplated, both acute and obtuse. The first longitudinal wall 111 is integral with the second longitudinal wall 112 because the sleeve 110 is formed by bending a single sheet of boron carbide aluminum matrix composite along the crease 113 to form the chevron shape with the desired angle. The single sheet of boron carbide aluminum matrix composite (and thus the sleeve 110) preferably has a gauge thickness t (FIG. 7) between 0.065 to 0.120 inches, and most preferably about 0.050 inches. The crease 113 is preferably formed with an apex radii between 0.375 to 0.625 inches. Of course, the invention is not limited to any specific apex radii or gauge thickness unless specifically recited in the claims. However, these dimensions will affect process optimization parameters during the boron carbide aluminum matrix composite sheet bending procedure and should be considered, specifically the bending rate and required temperatures of the work piece and tools. The sleeve 110 has a length L that extends from its bottom edge 114 to its top edge 115. The bottom edge 114 has a skewed shape to facilitate ease of insertion of the neutron absorbing insert 100 into a cell of a fuel rack. Specifically, the bottom edge 114 of each of the first and second longitudinal walls 111, 112 taper upward and away from the crease 113. The length L of the neutron absorbing insert 100 is preferably chosen so that the sleeve 100 extends at least the entire height of the fuel assembly with which it is to be used in conjunction. More preferably, the length L is preferably chosen so that the bottom edge 114 of the sleeve 110 can contact and rest atop a base plate of a fuel rack when inserted into a cell of the fuel rack without the reinforcement assembly 120 contacting the fuel assembly loaded in that cell. In one embodiment, the length L of the sleeve 110 will be in a range between 130 and 172 inches, and more preferably between 145 and 155 inches. Of course, the invention is not so limited and any length L may be used. In some embodiments, the length L of the sleeve 110 will only extend a fraction of the fuel assembly's height. In many instances this will be sufficient to shield adjacent fuel assemblies from one another because the irradiated uranium rods do not extend the entirety of the fuel assembly's height as the fuel assembly's lid and its base structure add to its height. Each of the first and second longitudinal walls 111, 112 have a width W that extends from the crease 113 to their outer lateral edges 116. The width W is preferably in the range between 4.25 to 8.90 inches, and most preferably about 5.625 inches. Of course, the invention is not limited to any particular width W. Further, in some embodiments the width of the first and second longitudinal walls 111, 112 may be different from one another if desired. Of course, the most preferred width W of the first and second longitudinal walls 111, 112 will be decided on a case-by-case basis and will be primarily dictated by the width of the fuel assembly housing and/or the size of the cell of the fuel rack with which the neutron absorbing insert 100 will be used in conjunction. Furthermore, while the sleeve 110 is illustrated as a two-walled chevron shape embodiment, it is to be understood that the in some embodiments the sleeve 110 may have more than two longitudinal walls. For example, in an alternative embodiment, the sleeve 110 can be formed to have three longitudinal walls formed into a general U-shape. In such an embodiment, it is preferred that the longitudinal juncture between at least two of the longitudinal walls be formed by bending. However, all longitudinal junctures may be formed by bending if desired. The number of longitudinal walls will be dictated by the arrangement and shape of the cells in the fuel rack or apparatus in which the neutron absorbing insert 100 is to be used. Referring now to FIGS. 3-6C concurrently, the structural and component details of the top end of the neutron absorbing insert 100 and the reinforcement assembly 120 will be described. The top end of the sleeve 110 comprises first and second flanges 117, 118 bent inwardly toward a central axis. The comprises first and second flanges 117, 118 are bent into the top end of each of the first and second longitudinal walls 111, 112 respectively. The flanges 117, 118 extend from the inner major surfaces 101, 102 of the first and second longitudinal walls 111, 112 at an approximately 90 degree angle. The flanges 117, 118 are arranged in an approximately orthogonal relationship to one another and are separated by a gap 119 (FIG. 4). The flanges 117, 118 provide structural rigidity to the first and second longitudinal walls 111, 112 and also provide a connection area for the L-shaped reinforcement block 121. While the flanges 117, 118 are formed by bending the sheet of boron carbide aluminum matrix composite material, in other embodiments, the flanges can be connected as separate components (such as blocks) or omitted all together. Each of the flanges 117, 118 comprise a plurality of holes 103 extending through the flanges 117, 118. The holes 103 are sized and shaped so that the dowels 125 of the dowel bar 124 can slidably pass therethrough. The reinforcement assembly 120 generally comprises a reinforcement block 121 and a dowel bar 124. The reinforcement block 121 is an L-shaped solid block of aluminum. Of course, other shapes and materials can be utilized. Moreover, the reinforcement block 121 can be a plurality of blocks working together. The reinforcement block 121 serves two primary functions: (1) to provide structural rigidity and integrity to the neutron absorbing insert 100 (and the sleeve 110); and (2) to provide an adequately strong structure by which a handling mechanism can engage, lift, lower, rotate and translate the neutron absorbing insert 100. The reinforcement block 121 comprises a plurality of engagement holes 122 that provide a geometry to which a lifting tool can engage for movement of the neutron absorbing insert 100. Of course, other mechanism can be used for the interlock mechanism, such as eye books, tabs, etc. Dowel holes 123 are also provides through the reinforcement block 121. The dowel holes 123 are sized and shaped to slidably accommodate the dowel pins 125 of the dowel bar 124 in a tight fit manner. The dowel bar 124 comprises a body 126 having a top surface and a bottom surface. A plurality of dowel pins 125 protrude form the top surface of the body 126. The dowel bar 124 is preferably aluminum. When assembled, the dowel bars 124 are positioned below the flanges 117, 118 while the reinforcement bar 121 is positioned above the flanges 117, 118. The components 121, 124, 110 are properly aligned so that the dowel pins 125 are slidably inserted through the flange holes 103 and into the holes 123 on the reinforcement bar 121, thereby sandwiching the flanges 117, 118 therebetween. The dowels 125 are secured within the holes 123 of the reinforcement block 121 by any desired means, such as a tight-fit-assembly, welding, adhesion, threaded interlock, a bolt, etc. FIG. 8 is an alternative embodiment of a neutron absorbing insert 100A. The neutron absorbing insert 100A is identical to the neutron absorbing insert 100 described above with the exception that a different reinforcement mechanism 120A is utilized. As can be seen, the major difference is that the interlock holes 122A are slots extending laterally through the block body 121. The different design is utilized to accommodate a different handling tool. As mentioned above, the sleeve 110 of the neutron absorbing insert 100 is formed by bending a single sheet of boron carbide aluminum matrix composite material. Since the boron carbide aluminum metal matrix composite material (and other metal matrix composite having neutron absorbing particulate reinforcement) exhibit the high stiffness and low ductility mechanical properties—they are very difficult and/or impossible to fabricate using conventional metal work equipment and metallurgical practices. This difficulty in fabrication becomes even more difficult as the particulate reinforcement level approaches 25% volume loading or greater of ceramic particulate. At high ceramic particulate volume loadings the elongation drops by a factor of 3 to 4 compared to the monolithic conventional aluminum alloys. To further increase the difficulty of fabricating the metal matrix composite material addition of the ceramic particulate dramatically increase the flow stress by up to 25% as the reinforcement loading level increases in the aluminum matrix. In order to make possible the useful bending of silicon carbide aluminum matrix composite material, a novel and nonobvious manufacturing process has been developed, referred to herein as “hot fabrication process technology.” This process will be described in detail below. It has been through the development of this hot fabrication process technology that the formation of useful products through bending of boron carbide aluminum matrix composite material has become possible. Of course, the fundamentals of this process can be easily applied to other metal matrix composite materials having neutron absorbing particulate reinforcement, with minor process parameter optimization. In order to successfully bend an aluminum boron carbide metal matrix composite material into a “chevron” profile one must modify all equipment and process parameters compared to conventional aluminum alloys in a number of ways. In order to produce suitable panels (i.e., sheets) of aluminum boron carbide metal matrix composite material, the quality of the work rolls used in the rolling process are first improved to overcome the abrasive nature and the propensity of the rolls to dimple during the sheet fabrication process. This is done through a hot rolling step. The hot rolling is performed while maintaining the material rolling temperature between 890 to 1010° F. Because the panels are so thin, the rollers (and other tools) are also heated to temperatures corresponding to the temperature of the panel at that step so as to eliminate rapid heat loss from the panel when contact is made with the rollers (or other interfaces). Once hot rolled, the rough panels are thermally straightened and flattened. In order to straighten and flatten the panel to meet the necessary specifications—a modified roll leveler is used. The roll leveler is modified to allow for “hot” roll leveling between a 750-1000° F. operating temperature. The roll leveler is designed to accommodate high temperature leveling without seizing up. The rough hot panel is then sheared to the desired final length and width. At this time, the necessary skew is sheared into the bottom edge of the panel, resulting in the single panel 150 shown at FIG. 9A. Subsequently, a V-shaped notch 105 is cutout of the top edge of the panel 150 and the dowel holes 103 are punched therein (FIG. 9B). The flanges 117, 118 are then bent into the panel 150 by bending the panel 150 along line C-C (FIG. 9B). The panel 150 is then bent into the chevron shape along line D-D (FIG. 9C) using the hot brake press 200 illustrated in FIG. 10. In order to bend the panel 150 into the chevron profile, the brake punch 201 and die 202 of the brake press 200 are heated to a temperature above 500 degrees Fahrenheit, and preferably between 500 and 1000 degrees Fahrenheit, using immersion heaters 203. The tip of the brake punch 201 has a ⅛ inch radius while the corresponding valley of the die 202 terminates at an apex having a radius of 3/16 inch. The panel 150 is also heated to a temperature above 750° F., preferably between 890-1010° F., before bending the panel 150 into chevron profile illustrated in FIG. 9D. The last step in the process is a thermal flattening operation performed on the thermal press 300 illustrated in FIG. 11. The thermal flattening operation coins the chevron profile of the panel 150 to meet a 90°+/−2° apex angle and flatten the longitudinal walls to meet the customer flatness and twist specification. This thermal flattening/coin operation is performed in a specially designed fixture/tool 300 which develops a minimum pressure of 20 pounds per square inch and uniform pressure distribution over the entire length of the chevron profiled panel 150. FIG. 12 illustrates a device 400 for checking the flatness and straightness of the final chevron-shaped sleeve panel 150. The device 400 has a plurality of parallel steel plates 410 having aligned slots 420 that allow the chevron-shaped sleeve panel 150 to slide therethrough if it is within specification. It should be pointed out that part of the novelty of this technology is the flex-ability of the process to manufacture chevrons to meet PWR or BWR or any other fuel manufacturer fuel storage requirements. Chevrons have been manufactured with legs from 4.250″-8.900″ width, gauge thickness for 0.065″-0.120″ T, apex radii from 0.375-0.625 inches, and lengths from 130-172″ L. It appears from initial fabrications that the process is very scalable and is capable of meeting all known spent fuel storage applications. Referring now to FIGS. 16-23, an alternative embodiment of a neutron absorbing insert 500 (and a method of installing the same in a fuel rack) according to the present invention is disclosed. The neutron absorbing insert 500 is similar to the neutron absorbing insert 100 described above in material, specification and manufacture of the sleeve portion. Thus, only those details of the neutron absorbing insert 500 that differ from the neutron absorbing insert 100 will be described in detail below with the understanding that the discussion above is fully applicable. Referring first to FIGS. 16, 17A and 17B concurrently, the neutron absorbing insert 500 generally comprises a sleeve 510. Unlike the neutron absorbing insert 100, the neutron absorbing insert 500 does not have a reinforcement block or structure at the top of the sleeve 510. Instead, the tops of the walls 511, 512 of the sleeve 510 comprise flanges 513, 514 that are formed by bending the walls 511, 512 The flanges 513, 514 extend from the walls 511, 512 outwardly away from the central axis E-E of the neutron absorbing insert 500 so as to allow a fuel assembly to move freely along axis E-E without obstruction from the flanges 513, 514. This allows the fuel assembly to be loaded into and unloaded from a cell within the fuel rack that utilizes the neutron absorbing insert 500 without the need to remove the neutron absorbing insert 500 during such procedures. The flanges 513, 514 are preferably inclined upward and away from the axis E-E, thereby forming a funnel structure for guiding the fuel assembly into proper position during a loading procedure. The inclined nature of the flanges 513, 514 also minimizes the horizontal space in which the flanges 513, 514 extend, thereby minimizing the possibility of interfering with other neutron absorbing inserts 500 located in adjacent cells in the fuel rack. In other embodiments, the flanges may be bent at a 90 degree angle to the walls 511, 512 if desired. Furthermore, while the flanges 513, 514 are preferably formed by bending the top ends of the walls 511, 512, the flanges 513,514 may, of course, be omitted all together or can be connected as separate structures in other embodiments. Moreover, a reinforcement block or structure can also be utilized if desired. In such a scenario, the reinforcement structure is preferably located on the outside surface of the walls 511, 512 so as to avoid obstructing free movement of the fuel assembly along axis E-E. Holes 515 are provided in the flanges 513, 514 so as to provide a simple mechanism by which the neutron absorbing insert 500 can be lifted and lowered within the fuel pool by a hook or other grasping tool. Of course, the holes 515 could be provided in the walls 511, 512 or can be omitted all together so long as some structure or surface arrangement is provided for facilitating movement of the neutron absorbing insert 500. The neutron absorbing insert 500 also comprises flanges 516, 517 located at the bottom end of the sleeve 510. The flanges 516, 517 extend inwardly toward the axis D-D of the neutron absorbing insert 500. As will be discussed in grater detail below, this allows the neutron absorbing insert 500 to be adequately secured to the fuel rack at its bottom end and in a manner that does not interfere with loading and/or unloading the fuel assembly along axis E-E. The flanges 516, 517 are preferably formed at an approximate 90 degree angle to the walls 511, 512 but the invention is not so limited. Furthermore, while the flanges 513, 514 are preferably formed by bending the bottom ends of the walls 511, 512, the flanges 513, 514 may, of course, be connected as separate structures in other embodiments. The radius of curvature discussed above for the crease can be used for the bottom flanges. Referring now to FIG. 18, a hold-down plate 600 is illustrated. The hold-down plate 600 comprises a plate-like body 601 formed of aluminum or other non-corrosive material. The plate 601 is of sufficient thickness to be adequately rigid so as not to deflect when performing its anchoring function discussed below. A central hole 605 is provided in the plate 601. A plurality of bendable pins or barbs 602 are attached to the plate 601 about the periphery of the central hole 605 in a circumferentially spaced apart arrangement. The barbs 601 extend beyond and protrude from the bottom surface of the plate 601. The barbs 602 are movable between an open position in which the barbs 601 can pass through a flow hole in the floor of a cell in the fuel rack and a locking position in which the barbs 601 engage the floor of a cell in the fuel rack. While the securing structure is illustrated as bendable barbs, the neutron absorbing insert 500 can be secured to the fuel rack in a variety of ways, including resilient tangs, a conical ridge that forms a tight-fit with the hole in the floor, fasteners, clamps, and/or combinations thereof. In one embodiment, rotatable cams may be used. Referring to FIGS. 19A and 19B concurrently, the hold-down plate 600 is shown in its installed position wherein it is securing the neutron absorbing insert 500 in place within the cell of the fuel rack. The walls of the fuel rack are illustrated in phantom for ease of illustration. The installation of the neutron absorbing insert 500 into a cell of a fuel rack will now be discussed. During installation of the neutron absorbing insert 500 into a cell of a fuel rack, the cell is initially empty (i.e., it does not contain a fuel assembly). In an initial step, the neutron absorbing insert 500 is coupled to a crane by using a hook that engages the holes 515 on the flanges 513, 514 of the sleeve 510. The neutron absorbing insert 500 is then aligned above the empty cell of the fuel rack and is lowered into the cell with its bottom end leading the way. The neutron absorbing insert 500 is lowered until the bottom flanges 516, 517 contact and rest atop the floor 700 of the fuel rack via a surface contact. Once the neutron absorbing insert 500 is in place within the fuel cell, the hold-down plate 600 is then lowered/inserted into the fuel cell with an appropriate tool. At this stage, the barbs 601 of the hold-down plate are in an open position (i.e., bent toward the axis of the central hole 605. The hold-down plate 600 continues to be lowered until it contacts the upper surfaces of the bottom flanges 516, 517 of the neutron absorbing insert 500. At this time, the barbs 601 insert into the hole 705 of the floor 700 of the fuel rack in the open position (the barbs are in the closed position in FIGS. 19A-19B). The central hole 605 of the hold-down plate 600 is substantially aligned with the hole 705 of the floor 700 of the fuel rack. This allows the cooling water within the pool to freely flow into the fuel cell as needed and in an unimpeded manner. As can be seen the, bottom flanges 516, 517 of the neutron absorbing insert 500 are located between (i.e. sandwiched) the floor 700 of the fuel rack and the hold-down plate 600 at this time. Referring now to FIGS. 20-22 concurrently, once the hold-down plate 600 is in position, a plunger tool 800 is inserted into the fuel cell. A head 801 of the plunger tool 800 comprises a chamfered disc 802 that is inserted into the holed 605, 705. As the chamfered disc 802 slides through the holes 605, 705, the barbs 601 are bent outward (away from a central axis of the holes 605, 705). The barbs 601 are bent outward until their head portions slide under the floor 700 of the fuel rack and their elongated body portions contact the side walls of the holes 605, 705. As a result, the barbs 601 lock the hold-down plate 600 in place, thereby securing the neutron absorbing insert 500 in place within the fuel cell by compressing the bottom flanges 516, 517 between the floor 700 and the plate 600. Of course, other tools and locking mechanisms can be used. Once the neutron absorbing insert 500 is secured in place, the fuel assembly 900 can be lowered safely into the fuel rack (FIG. 23). As discussed above, in order for a neutron absorbing insert to be used in existing fuel racks, the sleeve needs to be sufficiently thin so that it can fit within the small space formed between a fuel assembly housing and the walls of the fuel cell. During further development of a commercially viable neutron absorbing insert, it was discovered that while the aforementioned manufacturing process could be used to successfully bend a single sheet of boron carbide aluminum matrix composite material into a chevron shape, the walls of the chevron shaped sleeve were experiencing undesirable degrees of waviness and/or curvature (i.e., non-planarity) within each wall. The non-planarity of the walls of the chevron shaped sleeve can present serious issues with respect to the sleeve of the neutron absorbing insert properly fitting within the small space formed between a fuel assembly housing and the walls of the fuel cell. Moreover, even if the chevron shaped sleeve could be fit into the space, non-planarity of the walls could impede the sleeve and/or fuel assembly housing from being subsequently slid in and out of the fuel rack during loading and/or unloading operations. It has been discovered that creating a plurality of holes along the desired crease line, prior to or after bending, eliminates the non-planarity within the walls of the resulting sleeve. Surprisingly, these openings do not present a significant pathway for neutron radiation escape through the sleeve and do not appear to pose any substantial threat of criticality arising between fuel assemblies housed within adjacent cells of the fuel rack. It is to be understood that the principles described above with respect to FIGS. 1-23 for the sleeves 110, 510 (and the manufacturing process) are applicable to this alternative embodiment of the sleeve 110B (and the associated manufacturing and fuel rack loading processes). Furthermore, this alternative embodiment of the sleeve 110B (and the associated manufacturing process) can be incorporated into the neutron absorbing insert 100, formed into the desired configuration for use in the neutron absorbing insert 500, or used with other neutron absorbing apparatus. Thus, only those significant aspects of the sleeve 110B that differ from the sleeve 110 will be discussed. With reference to FIGS. 24A-G, the sleeve 110B and the process for manufacturing the sleeve 110B will now be described. Again, the hot manufacturing processes discussed above with respect to FIGS. 9A-9D are generally applicable to the creation of the sleeve 110B and, thus, the discussion will not duplicated with the understanding that the same basic processing steps, machines and parameters are used. Beginning with FIG. 24A, a rough hot panel is sheared to the desired final length and width. The necessary skew is sheared into the bottom edge of the panel, resulting in the panel 150B shown at FIG. 24A. The panel 150B is a single sheet of a metal matrix composite having neutron absorbing particulate reinforcement. Preferably, the panel 150B is a single sheet of boron carbide aluminum matrix composite material. The gauge thickness of the panel 150B is preferably 0.04 to 0.10 inches, more preferably 0.06 to 0.08 inches, and most preferably 0.07 inches. In one embodiment, the panel 100B is constructed of an aluminum boron carbide metal matrix composite material that is preferably 15% to 35% by weight boron carbide, 20% to 30% by weight boron carbide, and most preferably between 24% to 25% by weight boron carbide. Of course, the invention is not so limited and other percentages may be used. As shown in FIG. 24B, a V-shaped notch 105B is cutout of the top edge of the panel 150B and the dowel holes 103B are punched therein. Of course, the formation of the V-shaped notch 105B and/or the dowel holes 103B can be performed at a subsequent stage of the processing or can be omitted all together. Referring now to FIGS. 24C-24D, a plurality of spaced-apart holes 160B are formed into the flat panel 150B in a linear arrangement along the intended crease line D-D. The spaced-apart holes 160B form through holes in the panel 150B, forming passageways through the panel 150B. The spaced-apart holes 160B extend the entire length of the panel 150B, from at or near the bottom edge of the panel 150B to at or near the top edge of the panel 150B. In the exemplified embodiment, the spaced-apart holes 160B are in the form of elongated slits having rounded edges. Preferably, the elongated slits 160B cover between 50% to 70% of the entire length of the crease D-D. The invention, however, is not so limited. It is nonetheless preferred that a sufficient amount of the holes 160B be provided along the crease D-D to substantially eliminate (or reduce to an acceptable tolerance of 0.25 inches) waviness in the first and second walls. The elongated slits 160B are preferably formed by a water jet cutting tool. Of course, other cutting techniques may be used, including without limitation punching, pressing, milling, and torching. The elongated slits 160B may be formed by creating circular pierce holes at the desired distance apart and then connecting these pierce holes by forming a slit that extends between the pierce holes with the water jet cutter. The elongated slits 160B have a major axis and a minor axis. The major axis of the elongated slits 160B are coextensive with the desired crease line D-D. The minor axis of the elongated slits 160B are substantially perpendicular to the major axis and, thus, extend perpendicular to the desired crease line D-D. While the spaced-apart holes 160B are exemplified as elongated slits, the invention is not so limited in all embodiments. In other embodiments, the spaced apart holes may be circular, rectangular, or any other shape. Moreover, alternative arrangements may be used, such as perforations, score lines, or other pre-weakening techniques. The major axis of the elongated slits 160B have a length LMAJ and the minor axis of the elongated slits 160B have a length LMIN. Comparatively, in one embodiment, the length LMAJ of the major axis is preferably between 50 to 100 times longer than the length LMIN of the minor axis, more preferably between 60 to 80 times longer than the length LMIN of the minor axis, and most preferably 75 times longer than the length LMIN of the minor axis. In one embodiment, length LMAJ is preferably between 4 to 8 inches, and more preferably 6 inches. In such an embodiment, the length LMIN preferably between 0.05 to 0.1 inches, and more preferably 0.08 inches. Of course, the invention is not limited to any specific length or ratio in all embodiments, and may be determined on case-by-case basis. Furthermore, in alternative embodiments, the lengths LMAJ and/or LMIN may vary from hole to hole. Adjacent elongated slits 160B in the linear arrangement are separated by a distance d. The distance d is preferably shorter than the length LMAJ of the major axis of the elongated slits 160B. Comparatively, the distance d is preferably between 50% to 75% of the length LMAJ of the major axis, and more preferably 66% of the length LMAJ of the major axis. In one embodiment, the distance d is preferably between 2 to 6 inches, and more preferably 4 inches. Of course, the invention is not limited to any specific length or ratio in all embodiments, and may be determined on case-by-case basis. Furthermore, in alternative embodiments, the distance d may vary along the length of the crease D-D. Referring now to FIGS. 24E-24F, once the spaced-apart holes 160B are formed, the flanges 117B, 118B are bent into the panel 150B by bending the panel 150B along line C-C. With reference to FIGS. 24F-24G, the panel 150B is then bent along the crease line D-D, thereby forming the sleeve 110B. Bending of the panel along the crease line D-D results in the crease 113B of the resulting sleeve 110 to comprise the elongated slits 160B. The presence of the elongated slits 160B in the crease 113B allows the sheet 150B to remain in a bent arrangement without the creation of stresses that create waviness within each of the longitudinal walls 111B, 112B formed. The crease 113B connects the non-coplanar longitudinal walls 111B, 112B together. In the illustrated embodiment, the non-coplanar longitudinal walls 111B, 112B are in a chevron shape. It should be noted that the bending of the panel 150B may result in the elongated slits 160B becoming visibly minimized and/or eliminated from the final sleeve 110B. Preferably, the resulting sleeve 110B has an inner radius of curvature along the crease 113B from 0.15 to 0.25 inches, and more preferably 0.22 inches. Of course, the invention is not so limited. Finally, while the invention is described wherein the formation of the elongated slits 160B in the panel 150B takes place prior to the panel 150B being bent, it is possible for the elongated slits 160B to be formed into the panel 150B at a subsequent or preceding step in the process. Furthermore, in some embodiments, the elongated slits 160B may be formed into the crease 113B of the sleeve 110B after the panel 150B has been bent into the chevron-shape to eliminate built-up stresses. In this manner, pre-existing neutron absorbing inserts, such as neutron absorbing insert 100A, can be processed to eliminate undesired non-planarity in the walls. The present invention has been described in relation to the accompanying drawings; however, it should be understood that other and further modifications, apart from those shown or suggested herein, may be made within the spirit and scope of the present invention. It is also intended that all matter contained in the foregoing description or shown in the accompanying drawings shall be interpreted as illustrative rather than limiting.
039980570	claims	1. A nuclear power plant, comprising: a. an inner, generally cylindrical pressure vessel; b. a high-temperature reactor contained within a generally centrally oriented cavity within said inner vessel; c. a gas tubine assembly located in a horizontally oriented chamber positioned in said inner vessel beneath said reactor; d. a plurality of heat exchanger means positioned in a plurality of vertically oriented pods spaced radially about said reactor in the wall of said inner vessel, there being a plurality of first such heat exchanger means and at least one second heat exchanger means; e. a plurality of hot gas conduit means interconnecting said reactor and said turbine assembly, each of said hot gas conduit means including a horizontal connection to said reactor and to said turbine and a vertical section between said connections; f. a combined turbine exhaust gas distribution/cooled gas collection system horizontally positioned in the bottom wall of said inner vessel beneath said gas turbine assembly, said distribution/collection system comprising distribution means including inner conduit means for transporting exhaust gas from said turbine to each of a plurality of said first heat exchanger means and collection means including outer conduit means, coaxial to said inner conduit means, for transporting cooled gas from each of said first heat exchanger means to a collection point, said conduit means being comprised of horizontally and vertically oriented sections, whereby said distribution means and said collection means are disposed coaxially with respect to one another; and g. means for transporting cooled gas from said collection point to said reactor, including at least as many vertically extending cool gas conduit means as the number of said hot gas conduit means, one of said vertically extending cool gas conduit means coaxially surrounding the vertical section of each of said hot gas conduit means, whereby each of said conduits carrying gas at an elevated temperature is surrounded by cooler gas flowing in a coaxially arranged conduit. 2. The nuclear power plant as defined by claim 1, wherein said first heat exchanger means comprise a recuperator coupled with a pre-cooler, said recuperator being positioned vertically with respect to said pre-cooler in said vertical pod. 3. The nuclear power plant as defined by claim 2, wherein there are six of said first heat exchanger means and wherein said pre-cooler is in each instance positioned beneath said recuperator. 4. The nuclear power plant as defined by claim 2, wherein said vertically oriented pods are spaced symmetrically in a circle about said reactor. 5. The nuclear power plant as defined by claim 1, wherein said vertically extending cool gas conduit means are arranged as pods spaced about said reactor. 6. The nuclear power plant as defined by claim 4, wherein said gas turbine assembly comprises a gas turbine, a high-pressure compressor and a low-pressure compressor, said turbine assembly being removable as a single unit. 7. The nuclear power plant as defined by claim 6, wherein said means for transporting cooled gas from said collection point to said reactor comprises, interconnected in series between said collection point and said reactor, said low-pressure compressor, at least one of said second heat exchanger means, said high-pressure compressor, and a parallel connection to each of said recuperators, wherein said vertically extending coal gas conduit means are positioned between said high-pressure compressor and each of said recuperators. 8. The nuclear power plant as defined by claim 7, wherein said means for transporting cooled gas from said collection point to said reactor further comprises conduit means for transporting re-heated cool gas from each of said recuperators to said reactor, a substantial portion of the length of said re-heated gas conduit means being positioned within one of said vertically extending cool gas conduit means, whereby said conduit means carrying re-heated gas are surrounded over said portion of length by cool gas in said vertically extending conduit means. 9. The nuclear power plant as defined by claim 1, further comprising vertically oriented conduit means connecting each of said first heat exchanger means with said distribution means and, coaxially surrounding said vertically oriented conduit means, outer vertically extending conduit means connecting each of said first heat exchanger means with said collection means. 10. The nuclear power plant as defined by claim 7, further comprising a single vertical gas transporting conduit connecting said collection point and said low-pressure compressor. 11. The nuclear power plant as defined by claim 6, wherein said gas turbine, high- and low-pressure compressors are connected to a single shaft which is rigidly coupled to a generator. 12. The nuclear power plant as defined by claim 1, wherein the high-temperature reactor is equipped at its base with four radially placed connectors each of which leads into a vertically placed gas conduit, and wherein a horizontal conduit leads from each vertically placed conduit to a symmetrically arranged turbine-intake connector. 13. A nuclear power plant as defined by claim 12, wherein said reactor comprises graphite packing thereabout and said packing extends along said radially placed connectors to said vertically placed gas conduits. 14. The nuclear power plant as defined by claim 6, wherein there are two of said second heat exchanger means in said vertically oriented pods and wherein each comprises an intermediate cooler. 15. The nuclear power plant as defined by claim 14, wherein each of said two vertically oriented pods comprises two intermediate coolers arranged one above the other, and further comprising a coaxial gas transfer conduit leading from said low-pressure compressor to each of said intermediate cooler-containing pods and back to said high-pressure compressor, whereby gas flow toward said intermediate coolers takes place in the exterior path of said coaxial conduit and gas flow to said high-pressure compressor takes place in the interior path of said coaxial conduit. 16. The nuclear power plant as defined by claim 1, wherein said vertically oriented pods containing said heat exchanger means comprise gas-tight steel liners provided with insulation and water-cooling means. 17. The nuclear power plant as defined by claim 1, further comprising a secondary heat absorption means positioned inside said inner vessel, said device comprising a blower. 18. The nuclear power plant as defined by claim 17, wherein said secondary heat absorption means further comprises a recuperator and a cooling means. 19. The nuclear power plant as defined by claim 17, wherein one of said secondary heat absorption means is arranged in each of four vertically oriented pods spaced symmetrically around said reactor. 20. The nuclear power plant as defined by claim 1, further comprising control means for flow of said gas, all of said control means being located in vertically arranged chambers for easy accessibility from outside said inner vessel. 21. The nuclear power plant as defined by claim 1, further comprising a pressure tight safety vessel surrounding said inner vessel and all components carrying said gas. 22. The nuclear power plant as defined by claim 21, wherein said safety vessel comprises a cylindrical housing for holding a generator, said housing including pressure- and gas-tight sealing means.
055263879	abstract	A spacer for use with a fuel bundle in a nuclear reactor includes a matrix of ferrules for surrounding individual fuel rods within a bundle; a band surrounding the matrix and defining a peripheral wall of the spacer, the band having an upper edge; and a plurality of laterally spaced flow tabs extending upwardly from the upper edge, each flow tab having a lower substantially vertical portion and an upper inclined portion extending away from the vertical portion. The vertical portion and the inclined portion are formed with centrally located creases which define reverse bends in the upper and lower portions of the tab.
	claims	1. A method for verifying whether a post-trip inspection of a vehicle has been performed, comprising the steps of:(a) detecting that the vehicle has completed a trip;(b) after the trip has been completed, transmitting a signal indicating that a person has moved through the vehicle to a predefined location within the vehicle, while nominally conducting the post-trip inspection of the vehicle; and(c) determining whether the signal has been received before a predefined event occurs, and if not, determining that the person has not yet completed the post-trip inspection of the vehicle, wherein if it is determined that the person has not yet completed the post-trip inspection of the vehicle, performing at least one of the following functions:(i) storing data indicating that the person has not yet completed the post-trip inspection of the vehicle;(ii) displaying a status message indicating that the person has not yet completed the post-trip inspection of the vehicle; and(iii) producing an alarm. 2. The method of claim 1, wherein the alarm includes at least one of the following:(a) an audible alarm that is audible outside the vehicle; and(b) a visible alarm that is visible outside the vehicle. 3. The method of claim 1, wherein the step of detecting that the vehicle has completed a trip includes the step of uniquely identifying the vehicle. 4. The method of claim 1, wherein the step of uniquely identifying the vehicle comprises the step of remotely reading a token on the vehicle, said token being uniquely associated with the vehicle. 5. The method of claim 1, wherein the step of transmitting the signal occurs in response to at least one of the steps of:(a) reading a token that is disposed in the predefined location;(b) actuating a switch that is disposed in the predefined location, said switch being actuated by the person upon reaching the predefined location; and(c) reading a unique identification code that is disposed proximate the predefined location, with a sensor. 6. The method of claim 5, wherein the person carries a portable device when moving to the predefined location through the vehicle, farther comprising the step of using the portable device to read the unique identification code that is disposed at the predefined location. 7. The method of claim 6, further comprising the step of displaying at least one prompt to the person on the portable device regarding the post-trip inspection. 8. The method of claim 1, wherein the predefined event comprises at least one of:(a) a lapse of a predefined interval of time since detecting that the vehicle completed the trip;(b) a lapse of a predetermined time after powering off the vehicle; and(c) activating a switch that is external to the vehicle, where activation of the switch is intended to indicate that at least the post-trip inspection has been completed. 9. The method of claim 1, wherein the step of detecting that the vehicle has completed the trip comprises the step of sensing the vehicle arriving at a location that corresponds to an end of the trip. 10. A system for verifying whether a post-trip inspection of a vehicle has been performed, comprising:(a) a detector that detects when the vehicle has completed a trip, by producing a first signal indicative thereof(b) a sensor that produces a second signal indicating that a person has reached a predefined location within the vehicle, said predefined location being accessible only by moving through an interior of the vehicle while nominally completing a post-trip inspection of the vehicle; and(c) a monitor that receives the first signal from the detector and the second signal from the sensor, said monitor producing an indication that the person cannot yet have performed the post-trip inspection of the vehicle, if the second signal has not been received by the monitor before a predefined event occurs after the first signal was received by the monitor. 11. The system of claim 10, further comprising:(a) a transmitter for transmitting the second signal produced by the sensor; and(b) a receiver that receives the second signal, producing an output in response thereto, the output signal being conveyed to the monitor. 12. The system of claim 10, wherein the detector comprises one of the following:(a) a pressure sensor disposed at a location corresponding to an end of the trip, said pressure sensor responding to a weight of the vehicle by producing the first signal;(b) a light sensor disposed at a location corresponding to an end of the trip, said light sensor detecting passage of the vehicle past the light sensor, interrupting light received by the light sensor from a source, producing the first signal;(c) a video camera disposed at a location corresponding to an end of the trip, said video camera producing an image of at least a portion of the vehicle that is indicative of the vehicle being at the location, causing the first signal to be produced;(d) a radio frequency (RF) source and an RF detector, one of the RF source and the RF detector being disposed on the vehicle, and the other of the RF source and the RF detector being disposed at a location corresponding to an end of the trip, said RF detector responding to a radio signal from the RF source when the vehicle completes the trip, producing the first signal; and(e) a token reading device that responds to a token disposed on the vehicle that is read by the token reading device when the vehicle has completed the trip, producing the first signal. 13. The system of claim 10, wherein the first signal is conveyed to the monitor over at least one of:(a) a wireless communication link; and(b) a wired communication link. 14. The system of claim 10, wherein the second signal is conveyed to the monitor over at least one of:(a) a wireless communication link; and(b) a wired communication link. 15. The system of claim 10, wherein the sensor comprises a responder that responds by producing the second signal when the responder is proximate a token, one of the token and the responder being disposed at the predefined location within the vehicle, and the other of the token and the responder being portable and carried by a person moving to the predefined location within the vehicle. 16. The system of claim 15, wherein the responder includes a display on which at least one prompt regarding the post-trip inspection is displayed to a person. 17. The system of claim 10, wherein the detector comprises a responder that responds by producing the first signal when the responder is proximate a token, one of the token and the responder being disposed proximate an area where the detector detects when the vehicle has completed a trip, and the other of the token and the responder being portable and carried by the vehicle. 18. The system of claim 10, further comprising an optically encoded identifier, wherein the sensor comprises an optical reader for reading the optically encoded identifier, at least one of the optical reader and the optically encoded identifier being disposed at the predefined location within the vehicle and the other of the optical reader and the optically encoded identifier being carried by a person moving to the predefined location within the vehicle. 19. The system of claim 10, wherein the sensor comprises a switch that is actuated by a person arriving at the predefined location, causing the second signal to be produced. 20. The system of further comprising a transmitter, wherein the switch actuates the transmitter, causing it to transmit the second signal. 21. The system of claim 10, wherein one of the first signal and the second signal uniquely identifies the vehicle. 22. The system of claim 10, wherein the indication is an alarm condition and comprises at least one of:(a) a status message that is displayed on the monitor;(b) an audible sound; and(c) a visible light. 23. The system of claim 10, wherein the predefined event comprises at least one of:(a) lapse of a predefined interval of time since detecting that the vehicle completed the trip;(b) lapse of a predetermined time after powering off the vehicle; and(c) activation of a switch that is external to the vehicle, wherein activation of the switch is intended to indicate that at least the post-trip inspection has been completed. 24. A method for verifying whether at least one of a pre-trip inspection and a post-trip inspection of a vehicle has been performed, comprising the steps of:(a) detecting a triggering condition, the triggering condition indicating at least one of the following:(i) the vehicle is about to start at least a segment of a trip; and(ii) the vehicle has completed at least the segment of the trip;(b) after the triggering condition has been detected, transmitting a signal indicating that a person has been proximate at least one predefined location associated with the vehicle, while nominally conducting at least one of the pre-trip inspection and the post-trip inspection of the vehicle; and(c) determining whether the signal has been received before a predefined event occurs, and if not, determining that the person cannot yet have completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle, wherein if it is determined that the person cannot have completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle, performing at least one of the following functions:(i) storing data indicating that the person cannot have completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle;(ii) displaying a status message indicating that the person cannot have completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle; and(iii) producing an alarm. 25. The method of claim 24, wherein the alarm includes at least one of:(a) an audible alarm that is audible outside the vehicle; and(b) a visible alarm that is visible outside the vehicle. 26. The method of claim 24, wherein the step of detecting the triggering condition comprises the step of uniquely identifying the vehicle. 27. The method of claim 26, wherein the step of uniquely identifying the vehicle comprises the step of remotely reading a token on the vehicle, said token being uniquely associated with the vehicle. 28. The method of claim 24, wherein the step of transmitting the signal occurs in response to at least one of the steps of:(a) reading a token that is disposed proximate the at least one predefined location;(b) wherein the at least one predefined location comprises a plurality of predefined locations, reading a plurality of tokens, each of which is disposed proximate a different one of the plurality of predefined locations, such that the signal is not transmitted until the plurality of tokens have been read;(c) actuating a switch that is disposed proximate the at least one predefined location, said switch being actuated by the person upon reaching the at least one predefined location;(d) wherein the at least one predefined location comprises a plurality of predefined locations, actuating a plurality of switches, each of which is disposed proximate a different one of the plurality of predefined locations, each switch being actuated by the person upon reaching the predefined location to which the switch is proximate, such that the signal is not transmitted until the plurality of switches are activated;(e) reading a unique identification code that is disposed proximate the at least one predefined location, with a sensor; and(f) wherein the at least one predefined location comprises a plurality of predefined locations, reading a plurality of unique identification codes, each of which is disposed proximate a different one of the plurality of predefined locations, with a sensor, such that the signal is not transmitted until the plurality of unique identification codes are read. 29. The method of claim 28, wherein the person carries a portable device when performing the at least one of the pre-trip inspection and the post-trip inspection of the vehicle, farther comprising the step of using the portable device to read each unique identification code that is disposed proximate the predefined locations. 30. The method of claim 29, farther comprising the step of displaying at least one prompt to the person on the portable device regarding the at least one of the pre-trip inspection and the post-tip inspection. 31. The method of claim 24, wherein the predefined event comprises at least one of:(a) a lapse of a predefined interval of time since detecting that the vehicle is about to begin at least the segment of the trip;(b) a lapse of a predetermined time after powering on the vehicle;(c) activating a switch that is external to the vehicle, where activation of the switch is intended to indicate that at least the pre-trip inspection has been completed;(d) a lapse of a predefined interval of time since detecting that the vehicle completed at least the segment of the trip;(e) a lapse of a predetermined time after powering off the vehicle; and(f) activating a switch that is external to the vehicle, where activation of the switch is intended to indicate that the at least one of the pre-trip and the post-trip inspection has been completed. 32. The method of claim 24, wherein the step of detecting a triggering condition comprises at least one of the steps of:(a) sensing that the vehicle is disposed at a location that corresponds to a beginning of at least the segment of the trip;(b) sensing that the vehicle has departed from a location that corresponds to the end of a previous trip;(c) sensing that the vehicle has been powered on; and(d) determining that a predefined period of time has elapsed since a previous post-trip inspection. 33. The method of claim 24, wherein the step of detecting a triggering condition comprises at least one of the steps of:(a) sensing that the vehicle has arrived at a location that corresponds to an end of at least the segment of the trip;(b) sensing that the vehicle has been powered off; and(c) determining that a predefined period of time has elapsed since a previous pre-trip inspection. 34. A system for producing an indication that at least one of a pre-trip inspection and a post-trip inspection of a vehicle has not been performed, comprising:(a) a detector that detects a triggering condition and produces a first signal indicative thereof, the triggering condition indicating at least one of the following:(i) the vehicle is about to start at least a segment of a trip; and(ii) the vehicle has completed at least the segment of the trip;(b) a sensor configured to produce a second signal indicating that a person has reached at least one predefined location associated with the vehicle, the at least one predefined location generally corresponding to an area to be inspected during the at least one of the pre-trip inspection and the post-trip inspection; and(c) a monitor that receives the first signal from the detector and the second signal from the sensor, said monitor producing an indication that the person cannot yet have performed the post-trip inspection of the vehicle, if the second signal has not been received by the monitor before a predefined event occurs after the first signal was received by the monitor. 35. The system of claim 34, further comprising:(a) a transmitter for transmitting the second signal produced by the sensor; and(b) a receiver that receives the second signal, producing an output in response thereto, the output signal being conveyed to the monitor. 36. The system of claim 34, wherein the detector comprises one of the following:(a) a pressure sensor disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said pressure sensor responding to a weight of the vehicle by producing the first signal;(b) a light sensor disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said light sensor detecting passage of the vehicle past the light sensor, interrupting light received by the light sensor from a source, producing the first signal;(c) a video camera disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said video camera producing an image of at least a portion of the vehicle that is indicative of the vehicle being at the location, causing the first signal to be produced;(d) a radio frequency (RF) source and an RF detector, one of the RF source and the RF detector being disposed on the vehicle, and the other of the RF source and the RF detector being disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said RF detector responding to a radio signal from the RF source when the vehicle completes the trip, producing the first signal; and(e) a token reading device that responds to a token disposed on the vehicle that is read by the token reading device when the vehicle has completed the trip, producing the first signal. 37. The system of claim 34, wherein the detector comprises one of the following:(a) a detector configured to determine that the vehicle has been powered up; and(b) a detector configured to determine that the vehicle has been powered down. 38. The system of claim 34, wherein one of the first signal and the second signal uniquely identifies the vehicle. 39. The system of claim 34, wherein the predetermined event comprises at least one of:(a) lapse of a predetermined time after powering up the vehicle;(b) lapse of a predetermined time after powering down the vehicle; and(c) activation of a switch that is external to the vehicle, wherein activation of the switch is intended to indicate that the at least one of the pre-trip and the post-trip inspection has been completed. 40. The system of claim 34, wherein the sensor comprises a responder that responds by producing the second signal when the responder is proximate a token, one of the token and the responder being disposed at the at least one predefined location associated with the vehicle, and the other of the token and the responder being portable and carried by a person moving to the at least one predefined location associated with the vehicle. 41. The system of claim 34, further comprising an optically encoded identifier, wherein the sensor comprises an optical reader for reading the optically encoded identifier, one of the optical reader and the optically encoded identifier being disposed at the at least one predefined location associated with the vehicle and the other of the optical reader and the optically encoded identifier being carried by a person moving to the at least one predefined location associated with the vehicle. 42. The system of claim 34, wherein the sensor comprises a switch that is actuated by a person arriving at the at least one predefined location, causing the second signal to be produced. 43. A method for verifying whether at least one of a pre-trip inspection and a post-trip inspection of a vehicle is likely to have been performed by a person, comprising the steps of:(a) detecting a triggering condition, the triggering condition indicating at least one of the following:(i) the vehicle is about to start at least a segment of a trip; and(ii) the vehicle has completed at least the segment of the trip;(b) waiting for a predefined event to occur;(c) determining whether the person has been proximate at least one predefined location associated with the vehicle, while nominally conducting at least one of the pre-trip inspection and the post-trip inspection of the vehicle, based upon an indication provided when the person is proximate the at least one predefined location; and(d) if it has been determined that the person has not been proximate at least one predefined location associated with the vehicle, then transmitting a signal indicating that the person has not yet completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle. 44. The method of claim 43, wherein a plurality of predefined locations is associated with the vehicle, and the step of transmitting the signal comprises the step of identifying those predefined locations for which no indication exists. 45. The method of claim 43, wherein the step of determining that the person has been proximate the at least one predefined location associated with the vehicle comprises the step of determining if a switch disposed adjacent to the at least one predefined location has been activated, activation of the switch comprising the indication. 46. The method of claim 45, wherein the person carries a portable device, further comprising the step of using the portable device to activate the switch disposed at the at least one predefined location. 47. The method of claim 43, wherein the predefined event comprises at least one of:(a) a lapse of a predefined interval of time since detecting the triggering condition; and(b) activating a switch that is external to the vehicle, where activation of the switch is intended to indicate that the at least one of the pre-trip and the post-trip inspection has been completed. 48. A method for verifying whether at least one of a pre-trip inspection and a post-trip inspection of a vehicle has been performed, comprising the steps of:(a) detecting a triggering condition, the triggering condition indicating at least one of the following:(i) the vehicle is about to start at least a segment of a trip; and(ii) the vehicle has completed at least the segment of the trip;(b) waiting for a first predefined event to occur;(c) determining if a second predefined event has occurred, and if not, determining that the person has not yet completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle; and(d) transmitting a signal indicating that the person has not yet completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle. 49. The method of claim 48, wherein the second predefined event comprises the activation of a switch disposed proximate the at least one predefined location associated with the vehicle. 50. The method of claim 49, wherein the person carries a portable device, further comprising the step of using the portable device to activate the switch disposed proximate the at least one predefined location associated with the vehicle. 51. The method of claim 48, wherein the person carries a portable device, and the second predefined event comprises the person using the portable device to read a token that is disposed proximate the at least one predefined location associated with the vehicle. 52. The method of claim 48, wherein the person carries a portable device, and the second predefined event comprises the person using the portable device to read a unique identification code that is disposed proximate the at least one predefined location associated with the vehicle. 53. The method of claim 48, wherein the first predefined event comprises at least one of:(a) a lapse of a predefined interval of time since detecting the triggering condition; and(b) activating a switch that is external to the vehicle, where activation of the switch is intended to indicate that the at least one of the pre-trip and the post-trip inspection has been completed. 54. A system for producing an indication when at least one of a pre-trip inspection and a post-trip inspection of a vehicle has not been performed, comprising:(a) a detector that detects a triggering condition and produces a first signal indicative thereof, the triggering condition indicating at least one of the following:(i) the vehicle is about to start at least a segment of a trip; and(ii) the vehicle has completed at least the segment of the trip; and(b) a sensor configured to receive the first signal, to produce a second signal, and to transmit the second signal, the second signal being a wireless communication not limited by a line of sight, the second signal being produced only if each of the following conditions are met:(i) the first signal has been received;(ii) a predefined event has occurred; and(iii) the sensor has not detected that a person has been proximate at least one predefined location associated with the vehicle, the at least one predefined location generally corresponding to an area to be inspected during the at least one of the pre-trip inspection and the post-trip inspection, the second signal thus indicating that a person has not yet completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle. 55. The system of claim 54, further comprising a wireless receiver configured to receive the second signal, the wireless receiver being disposed at a location that is remote from the sensor, and producing an indication that a person has not yet completed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle. 56. The system of claim 54, further comprising a monitor that receives the second signal from the sensor, said monitor producing an indication that a person has not yet performed the at least one of the pre-trip inspection and the post-trip inspection of the vehicle. 57. The system of claim 54, wherein the detector comprises one of the following:(a) a pressure sensor disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said pressure sensor responding to a weight of the vehicle by producing the first signal;(b) a light sensor disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said light sensor detecting passage of the vehicle past the light sensor, interrupting light received by the light sensor from a source, producing the first signal;(c) a video camera disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said video camera producing an image of at least a portion of the vehicle that is indicative of the vehicle being at the location, causing the first signal to be produced;(d) a radio frequency (RF) source and an RF detector, one of the RF source and the RF detector being disposed on the vehicle, and the other of the RF source and the RF detector being disposed at a location corresponding to at least one of an end of a segment of the trip and a beginning of a segment of the trip, said RF detector responding to a radio signal from the RF source when the vehicle completes the trip, producing the first signal; and(e) a token reading device that responds to a token disposed on the vehicle that is read by the token reading device when the vehicle has completed the trip, producing the first signal. 58. The system of claim 54, wherein the first signal is conveyed to the sensor over at least one of:(a) a wireless communication link; and(b) a wired communication link. 59. The system of claim 54, wherein the sensor is disposed proximate the at least one predefined location associated with the vehicle. 60. The system of claim 59, wherein the sensor comprises:(a) a switch that can be activated by a person to verify a person was proximate the at least one predefined location associated with the vehicle;(b) a wireless transmitter configured to transmit the second signal; and(c) a processor logically coupled to the switch and the wireless transmitter, the processor determining if the conditions are met, and if so, controlling the wireless transmitter to transmit the second signal. 61. The system of claim 60, wherein the switch comprises at least one of a token and a responder, and the other of the token and the responder being portable and carried by a person moving to at least one predefined location associated with the vehicle, the switch being activated when the token and the responder are placed in proximity to each other. 62. The system of claim 61, wherein the responder is portable and includes a display on which at least one prompt regarding the at least one of the pre-trip and post-trip inspection is displayed to a person. 63. The system of claim 54, further comprising an optically encoded identifier, wherein the sensor comprises an optical reader for reading the optically encoded identifier, at least one of the optical reader and the optically encoded identifier being disposed proximate the at least one predefined location associated with the vehicle and the other of the optical reader and the optically encoded identifier being carried by a person moving to the predefined location within the vehicle. 64. The system of claim 54, further comprising an RF source, wherein the sensor comprises an RF detector, one of the RF source and the RF detector being disposed proximate the at least one predefined location associated with the vehicle and the other of the RF source and the RF detector being carried by a person moving to the predefined location within the vehicle. 65. The system of claim 54, wherein the sensor is portable and configured to be carried by a person performing the at least one of the pre-trip inspection and the post-trip inspection. 66. The system of claim 54, wherein the second signal uniquely identifies the vehicle. 67. The system of claim 54, wherein the second signal uniquely identifies the at least one predefined location associated with the vehicle that a person has not been proximate, and thus cannot yet have inspected. 68. A method for verifying whether an inspection has likely been performed, comprising the steps of:(a) detecting a triggering condition, the triggering condition corresponding to the beginning of a period during which the inspection should be performed;(b) after the triggering condition has been detected, transmitting a signal indicating that a person has been proximate at least one predefined location associated with the inspection, while nominally performing the inspection; and(c) determining whether the signal has been received before a predefined event occurs, the predefined event corresponding to the end of a period during which the inspection should be, and if not, determining that the person cannot yet have completed the inspection, wherein if it is determined that the person cannot have completed the inspection, performing at least one of the following functions:storing data indicating that the person cannot have completed the inspection;(ii) displaying a status message indicating that the person cannot have completed the inspection; and(iii) producing an alarm condition. 69. A system for verifying whether an inspection has likely been performed, comprising:(a) a detector that detects a triggering condition, the triggering condition corresponding to the beginning of a period during which the inspection should be performed, the detector producing a first signal indicative thereof;(b) a sensor that produces a second signal indicating that a person has reached a predefined location associated with the inspection, while nominally performing the inspection; and(c) a monitor that receives the first signal from the detector and the second signal from the sensor, said monitor producing an indication that the person cannot yet have performed the inspection, if the second signal has not been received by the monitor before a predefined event occurs after the first signal was received by the monitor, the predefined event corresponding to the end of a period during which the inspection should be performed. 70. A method for verifying whether an inspection has likely been performed, comprising the steps of:(a) detecting a triggering condition, the triggering condition corresponding to the beginning of a period during which the inspection should be performed;(b) waiting for a predefined event to occur, the predefined event corresponding to the end of a period during which the inspection should be performed;(c) determining whether a person has been proximate at least one predefined location associated with the inspection, while nominally performing the inspection, based upon an indication provided when the person is proximate the at least one predefined location; and(d) if it has been determined that a person has not been proximate at least one predefined location, based upon a lack of the indication, then transmitting a signal indicating that the person has not yet performed the inspection. 71. A method for verifying whether an inspection has likely been performed, comprising the steps of:(a) detecting a triggering condition, the triggering condition corresponding to the beginning of a period during which the inspection should be performed;(b) waiting for a predefined event to occur, the predefined event corresponding to the end of a period during which the inspection should be performed;(c) determining if a second predefined event has occurred, and if not, determining that a person has not yet completed the inspection; and(d) transmitting a signal indicating that the person has not yet performed the inspection. 72. The method of claim 71, wherein the second predefined event comprises at least one of:(a) the activation of a switch disposed proximate at least one predefined location associated with the inspection; and(b) wherein the person carries a portable device, and the second predefined event comprises the person using the portable device to read a unique identification code that is disposed proximate at least one predefined location associated with the inspection. 73. A system for producing an indication when an inspection has not been performed, comprising:(a) a detector that detects a triggering condition, the triggering condition corresponding to the beginning of a period during which the inspection should be performed, the detector producing a first signal indicative thereof; and(b) a sensor configured to receive the first signal, to produce a second signal, and to transmit the second signal, the second signal being a wireless communication not limited by a line of sight, the second signal being produced only if all of the following conditions are met:(i) the first signal has been received;(ii) a predefined event has occurred, the predefined event corresponding to an end of a period during which the inspection should be performed; and(iii) the sensor has not detected that a person has been proximate at least one predefined location generally corresponding to an area to be inspected during the inspection, the second signal thus indicating that a person has not yet completed the inspection. 74. The system of claim 73, wherein the sensor is disposed proximate the at least one predefined location, the sensor comprising:(a) a switch that can be activated only when a person is proximate the at least one predefined location;(b) a wireless transmitter configured to transmit the second signal; and(c) a processor logically coupled to the switch and the wireless transmitter, the processor determining if the conditions are met, and if so, controlling the wireless transmitter to transmit the second signal. 75. The system of claim 73, further comprising a token, and wherein the sensor comprises a reader for reading the token, at least one of the reader and the token being disposed proximate the at least one predefined location, and the other of the reader and the token being carried by a person assigned to perform the inspection. 76. The system of claim 73, further comprising a token disposed proximate the at least one location, and wherein the sensor comprises a reader for reading the token, the reader being configured to be carried by a person assigned to perform the inspection. 77. The system of claim 76, wherein the sensor further comprises:(a) a wireless transmitter configured to transmit the second signal; and(b) a processor logically coupled to the switch and the wireless transmitter, the processor determining if the conditions are met, and if so, controlling the wireless transmitter to transmit the second signal.
039403114	claims	1. In a nuclear reactor having a pressure vessel with a removable closure head and fuel assemblies mounted within the vessel between a lower core plate and an upper core plate, in combination, means supporting the lower core plate and the fuel assemblies, an upper support plate mounted in the vessel, support tubes extending between the upper support plate and the upper core plate to support the upper core plate, rectilinearly movable control rods for preselected fuel assemblies, drive shafts attached to the control rods, adapter tubes extending through the closure head, said drive shafts entering the vessel through the adapter tubes, a plurality of control rod guide tubes located in each of said support tubes and extending the length thereof, and at least one of said control rod drive shafts and its attached control rod in each of said guide tubes. 2. The combination defined in claim 1, wherein the support tubes are generally cylindrical, each drive shaft is attached to at least two control rods, and the drive shafts for the control rods for each preselected fuel assembly are arranged in a circle about the center line of the support tube containing the drive shafts and control rods for that fuel assembly. 3. The combination defined in claim 2, wherein one half of the guide tubes in each support tube are generally triangular in cross section and the other half are generally oblong in cross section and disposed between the triangular guide tubes. 4. The combination defined in claim 2, wherein each guide tube has generally circular portions formed integrally therewith for receiving and guiding the drive shaft and the control rods. 5. The combination defined in claim 2, including generally ring-shaped support plates spaced along the length of each support tube and attached to the guide tubes therein to provide a guide tube assembly and maintain the relative lateral position of the guide tubes within the support tube. 6. The combination defined in claim 5, wherein each ring-shaped support plate has angularly spaced slots in its outer periphery, and including key means attached to the support tube and cooperating with said slots to align the guide tube assembly in the support tube. 7. The combination defined in claim 6, wherein the support tube has holes in its wall having the same angular spacing as the slots in the ring-shaped support plate, and said key means being disposed in said holes. 8. The combination defined in claim 7, wherein key means includes an insert attached thereto and extending through the hole in the support tube into the slot in the ring-shaped support plate. 9. The combination defined in claim 8, including spring means on the key insert engaging a wall of the slot in the ring-shaped support plate. 10. The combination defined in claim 9, wherein the spring means is a cantilever spring formed integrally with the insert. 11. The combination defined in claim 1, including upwardly extending beams on the upper support plate, a top plate mounted on said beams, an end plate attached to the top plate, and said guide tubes being secured to said end plate and extending to the upper core plate of the reactor. 12. The combination defined in claim 11, including means for aligning the end plate on the top plate, and additional means for aligning the support tubes on the upper core plate. 13. The combination defined in claim 12, including a base plate attached to the end plate, a support column secured to the base plate and extending into an adapter tube, and guide plates spaced on the support column for aligning and guiding the drive shafts in the adaptor tube.
	description	Field The present disclosure relates to the emergency (and planned) shutdown of a boiling water nuclear reactor (BWR) via the insertion of negative reactivity into the core. Description of Related Art In response to a scram signal during the operation of a boiling water nuclear reactor (BWR), a hydraulic control unit shuts down the reactor by inserting neutron-absorbing control rods into the core. In a conventional hydraulic control unit, a control rod is driven into the core by pressurized water that is applied to a metal piston and includes a seal formed by an elastomeric O-ring. However, the conventional piston/O-ring approach may be prone to leaks. As a result, regular maintenance (e.g., changing the elastomeric O-ring) must be performed to ensure proper functionality during a scram operation. A hydraulic control unit for a nuclear reactor may include a scram accumulator, a gas supply vessel, and/or a scram valve. The scram accumulator may have a first end with an inlet and an opposing second end with an outlet. The scram accumulator defines a chamber therein and contains bellows within the chamber. The bellows are configured to hold a scram liquid. The bellows have a fixed end and a moveable end. The fixed end of the bellows may be secured to the second end of the scram accumulator. The bellows are configured to transition between an expanded state and a compressed state via the moveable end. The gas supply vessel may be connected to the first end of the scram accumulator. The gas supply vessel is configured to hold a scram gas under pressure and in fluid communication with the chamber of the scram accumulator through the inlet so as to exert a compressive force on the moveable end of the bellows in a form of stored energy. The scram valve may be connected to the second end of the scram accumulator and configured to withstand the compressive force exerted through the bellows to contain the stored energy. The scram valve is configured to open in response to a scram signal to release the stored energy and allow the bellows to yield to the compressive force and transition to the compressed state. A method of shutting down a nuclear reactor may include compressing a scram gas that is in fluid communication with a scram accumulator. The scram accumulator defines a chamber therein and contains bellows within the chamber. The bellows are configured to hold a scram liquid in isolation of the scram gas. The scram gas exerts a compressive force on the bellows in a form of stored energy. The method may additionally include releasing the stored energy in response to a scram signal such that the scram gas expands into the chamber of the scram accumulator to compress the bellows and expel the scram liquid from the scram accumulator to insert control rods into a core of the nuclear reactor. It should be understood that when an element or layer is referred to as being “on,” “connected to,” “coupled to,” or “covering” another element or layer, it may be directly on, connected to, coupled to, or covering the other element or layer or intervening elements or layers may be present. In contrast, when an element is referred to as being “directly on,” “directly connected to,” or “directly coupled to” another element or layer, there are no intervening elements or layers present. Like numbers refer to like elements throughout the specification. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It should be understood that, although the terms first, second, third, etc. may be used herein to describe various elements, components, regions, layers and/or sections, these elements, components, regions, layers, and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region, layer, or section from another region, layer, or section. Thus, a first element, component, region, layer, or section discussed below could be termed a second element, component, region, layer, or section without departing from the teachings of example embodiments. Spatially relative terms (e.g., “beneath,” “below,” “lower,” “above,” “upper,” and the like) may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It should be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the term “below” may encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. The terminology used herein is for the purpose of describing various embodiments only and is not intended to be limiting of example embodiments. As used herein, the singular forms “a,” “an,” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “includes,” “including,” “comprises,” and/or “comprising,” when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. Example embodiments are described herein with reference to cross-sectional illustrations that are schematic illustrations of idealized embodiments (and intermediate structures) of example embodiments. As such, variations from the shapes of the illustrations as a result, for example, of manufacturing techniques and/or tolerances, are to be expected. Thus, example embodiments should not be construed as limited to the shapes of regions illustrated herein but are to include deviations in shapes that result, for example, from manufacturing. For example, an implanted region illustrated as a rectangle will, typically, have rounded or curved features and/or a gradient of implant concentration at its edges rather than a binary change from implanted to non-implanted region. Likewise, a buried region formed by implantation may result in some implantation in the region between the buried region and the surface through which the implantation takes place. Thus, the regions illustrated in the figures are schematic in nature and their shapes are not intended to illustrate the actual shape of a region of a device and are not intended to limit the scope of example embodiments. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which example embodiments belong. It will be further understood that terms, including those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and will not be interpreted in an idealized or overly formal sense unless expressly so defined herein. FIG. 1 is a perspective view of a hydraulic control unit according to an example embodiment of the present disclosure. Referring to FIG. 1, a hydraulic control unit 100 for a nuclear reactor includes a scram accumulator 102, a gas supply vessel 104, and a scram valve 106. Although not shown, it should be understood that, when the hydraulic control unit 100 is implemented in a nuclear reactor, the scram valve 106 will be operatively connected to a control rod to facilitate the insertion of the control rod into the reactor core during a scram operation. The scram accumulator 102 has a first end with an inlet and an opposing second end with an outlet. In FIG. 1, the first end of the scram accumulator 102 is at the bottom, and the second end of the scram accumulator 102 is at the top, although example embodiments are not limited thereto. The scram accumulator 102 defines a chamber therein and contains bellows within the chamber. The bellows will be discussed in further detail in connection with FIGS. 2-4. The bellows is configured to hold a scram liquid. The bellows has a fixed end and a moveable end. The fixed end of the bellows is secured to the second end of the scram accumulator. The bellows is configured to transition between an expanded state (e.g., FIG. 2) and a compressed state (e.g., FIG. 3) via the moveable end. The gas supply vessel 104 is connected to the first end of the scram accumulator 102. The gas supply vessel 104 is configured to hold a scram gas under pressure and in fluid communication with the chamber of the scram accumulator 102 through the inlet so as to exert a compressive force on the moveable end of the bellows in a form of stored energy. In particular, the gas supply vessel 104 is configured to hold a quantity of the scram gas that is sufficient to transition the bellows within the scram accumulator 102 from the expanded state to the compressed state upon opening of the scram valve 106. For instance, the gas supply vessel 104 may be configured to hold the scram gas such that the pressure within is at least 10 MPa (e.g., at least 15 MPa) prior to a scram operation. The gas supply vessel 104 may also be configured to hold an inert gas as the scram gas. The inert gas may be nitrogen (N2), although example embodiments are not limited thereto. The scram valve 106 is connected to the second end of the scram accumulator 102 and is configured to withstand all forces (e.g., the compressive force) exerted through the bellows (by the scram gas) to contain the stored energy. The scram valve 106 is a fast opening valve and is configured to open quickly in response to a scram signal to release the stored energy (and, thus, expel the scram liquid from the scram accumulator 102) and allow the bellows in the scram accumulator 102 to yield to the compressive force and transition to the compressed state. FIG. 2 is a cross-sectional view of a scram accumulator of a hydraulic control unit wherein the bellows therein are in an expanded state according to an example embodiment of the present disclosure. Referring to FIG. 2, the scram accumulator 200 has a first end 202 with an inlet 206 and an opposing second end 222 with an outlet 220. The scram accumulator 200 also includes an attachment structure 204 at the first end 202 and the second end 222 to facilitate the connection with the gas supply vessel (e.g., gas supply vessel 104 in FIG. 1) and the scram valve (e.g., scram valve 106 in FIG. 1), respectively. The scram accumulator 200 defines a chamber therein and contains bellows 214 within the chamber. The bellows 214 is configured to hold a scram liquid 216. In an example embodiment, the scram liquid 216 is water. The scram accumulator 200 is configured such that the scram liquid 216 is isolated from the scram gas 208 via the bellows 214 so that the scram liquid 216 and the scram gas 208 do not contact each other or intermingle within the chamber. The bellows 214 has a fixed end and a moveable end. The fixed end of the bellows 214 is secured (e.g., welded) to the second end 222 of the scram accumulator 200. The scram accumulator 200 also includes a stop structure 218 within the chamber at the second end 222. The moveable end of the bellows 214 includes a diffuser plate 210 that is configured to uniformly distribute the compressive force exerted by the scram gas 208 on the moveable end. Additionally, the moveable end of the bellows 214 includes a stop plate 212 that is configured to halt against the stop structure 218 during the compressed state. In an example embodiment, the scram accumulator 200 may have a cylindrical body that defines a cylindrical chamber therein. In such an embodiment, the bellows 214 may resemble a cylindrical accordion in order to more effectively occupy the chamber within the scram accumulator 200. Additionally, the stop structure 218 may be a pipe-like structure. As a result, the cylindrical body of the scram accumulator 200 and the stop structure 218 may define an annular space therebetween. The annular space may be adjusted to a size that is just sufficient to accommodate the bellows 214 (when expanded and compressed) in order to maximize the available space for the scram liquid 216. The diffuser plate 210 and the stop plate 212 may also be flat, circular structures that resemble a disk in order to correspond to the inner walls of the scram accumulator 200 that define the cylindrical chamber within. Although not shown, the scram accumulator 200 may further include an anti-rotation device or arrangement to maintain a desired alignment during the compression of the bellows 214 during a scram operation. For instance, the inner walls of the scram accumulator 200 that define the cylindrical chamber may be provided with one or more linear ridges that extend from the first end 202 to the second end 222. In addition, the diffuser plate 210 and the stop plate 212 may be provided with one or more grooves that mate with the one or more linear ridges on the inner walls defining the chamber so as to provide a guided track when the bellows 214 transitions from the expanded state to the compressed state. Conversely, the inner walls of the scram accumulator 200 that define the cylindrical chamber may be provided with one or more linear grooves, while the diffuser plate 210 and the stop plate 212 may be provided with one or more ridges that mate with the one or more linear grooves on the inner walls defining the chamber so as to provide a guided track when the bellows 214 transitions from the expanded state to the compressed state. The bellows 214 is configured to transition between an expanded state and a compressed state via the moveable end. The scram accumulator 200 is configured such that the moveable end of the bellows 214 is closer to the first end 202 than the second end 222 of the scram accumulator 200 during the expanded state. For instance, the scram accumulator 200 may be configured such that the bellows 214 occupies 80% or more of a volume of the chamber during the expanded state. In an example embodiment, depending on the size of the bellows 214 and the amount of the scram liquid 216 therein, the bellows 214 may fully occupy the volume of the chamber such that the diffuser plate 210 abuts the inlet 206. During the expanded state prior to a scram operation, the scram gas 208 from the gas supply vessel (e.g., gas supply vessel 104 in FIG. 1) exerts pressure on the bellows 214, which in turn exerts pressure on the scram liquid 216 therein which is retained via the scram valve (e.g., scram valve 106 in FIG. 1). The hydraulic control unit, particularly the scram accumulator and scram valve, is designed to be strong enough to contain the forces generated therein and to hold them in equilibrium in the form of stored energy until the proper time for release (e.g., during a scram operation). FIG. 3 is a cross-sectional view of a scram accumulator of a hydraulic control unit wherein the bellows therein are in a compressed state according to an example embodiment of the present disclosure. Referring to FIG. 3, the scram accumulator 200 is configured such that the scram liquid 216 is expelled from the outlet 220 at the second end 222 when the bellows 214 transitions from the expanded state to the compressed state. The scram accumulator 200 is configured such that the moveable end of the bellows 214 is closer to the second end 222 than the first end 202 of the scram accumulator 200 during the compressed state. Although not shown in FIG. 3, it should be understood that the stop plate 212 of the moveable end of the bellows 214 will be pressed against the stop structure 218 when the bellows 214 has fully transitioned to the compressed state. An O-ring may also be disposed on the rim of the stop structure 218 so as to form a seal when contacted by the stop plate 212. The size (e.g., length) of the stop structure 218 may be adjusted to allow for a full compression (or nearly a full compression) of the bellows 214 in order to maximize the amount of the scram liquid 216 expelled from scram accumulator 200. For instance, the scram accumulator 200 may be configured (e.g., adjusting length of stop structure 218) such that the bellows 214 occupies 30% or less (e.g., 20% or less) of a volume of the chamber during the compressed state. FIG. 4 is an enlarged and partial cross-sectional view of the bellows contained within a scram accumulator of a hydraulic control unit according to an example embodiment of the present disclosure. Referring to FIG. 4, the bellows is formed by welding a plurality of annular structures 402 together, wherein the plurality of annular structures 402 may be formed of a suitable metal. The plurality of annular structures 402 resemble flattened rings or thin, gasket-like structures, although the plurality of annular structures 402 may not be completely flat but instead provided with a slight curvature or other shaping. In any event, each of the plurality of annular structures 402 will have an outer diameter edge and an inner diameter edge. In the partial cross-sectional view of FIG. 4, the left side of the drawing corresponds to the inner diameter edges of the annular structures 402 (as well as the bellows), while the right side of the drawing corresponds to the outer diameter edges of the annular structures 402 (as well as the bellows). To form the bellows, adjacent annular structures 402 are alternately welded by their inner and outer diameters. In particular, referring to FIG. 4, one annular structure 402 is first arranged so as to be a mirror image of an adjacent annular structure 402. The outer diameters of the adjacent annular structures 402 are then welded to form a first weld 404. Next, another annular structure 402 is arranged so as to be a mirror image of one of the two welded annular structures 402. The inner diameters of the adjacent annular structures 402 are then welded to form a second weld 406. Afterwards, another annular structure 402 is arranged so as to be a mirror image of one of the three welded annular structures 402. The outer diameters of the adjacent annular structures 402 are then welded to form a first weld 404. Subsequently, another annular structure 402 is arranged so as to be a mirror image of one of the four welded annular structures 402. The inner diameters of the adjacent annular structures 402 are then welded to form a second weld 406. This approach is repeated as needed to form a bellows of the desired size. When completed, the bellows will resemble a cylindrical accordion. Notably, the bellows will have a double-welded structure, which is not only durable but is also able to compress/collapse to a greater degree than conventional bellows which are obtained by molding or bending. As mentioned above, the present hydraulic control unit may be used to shut down a nuclear reactor during an emergency or for a planned outage. Referring back to FIGS. 2-3, a method of shutting down a nuclear reactor may include compressing a scram gas 208 that is in fluid communication with a scram accumulator 200. The compressing may include filling a gas supply vessel (e.g., gas supply vessel 104 in FIG. 1) with the scram gas 208, wherein the gas supply vessel is in fluid communication with the chamber of the scram accumulator 200. In addition, the compressing may be performed to a pressure of at least 10 MPa (e.g., at least 15 MPa). The scram accumulator 200 defines a chamber therein and contains bellows 214 within the chamber. The bellows 214 are configured to hold a scram liquid 216 in isolation of the scram gas 208. In particular, the bellows 214 form a fluid-tight partition between the scram gas 208 and the scram liquid 216. As a result, during proper operation, the scram gas 208 will not exit through the outlet 220, and the scram liquid 216 will not exit through the inlet 206. Because the scram gas 208 is pressurized, the scram gas 208 exerts a compressive force on the bellows 214 in a form of stored energy. The method additionally includes releasing the stored energy in response to a scram signal such that the scram gas 208 expands into the chamber of the scram accumulator 200 to compress the bellows 214 and expel the scram liquid 216 from the scram accumulator 200 to insert control rods into a core of the nuclear reactor. The releasing may include opening a scram valve (e.g., scram valve 106 in FIG. 1) that is attached to the scram accumulator 200 so that the bellows 214 is able to yield to the compressive force exerted by the scram gas 208. In particular, the releasing may include the scram gas 208 entering the scram accumulator 200 from a first end 202, while the scram liquid 216 exits from an opposing second end 222 of the scram accumulator 200. As a result, the bellows 214 will collapse along a lengthwise direction of the chamber towards the second end 222. In an example embodiment, the scram gas 208 will decrease a volume of the bellows 214 by 70% or more (e.g., 80% or more). While a number of example embodiments have been disclosed herein, it should be understood that other variations may be possible. Such variations are not to be regarded as a departure from the spirit and scope of the present disclosure, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims.
	description	The application claims benefit of priority to U.S. Provisional Patent Application No. 62/210,609, entitled “FUEL ELEMENT WITH MULTI-SMEAR DENSITY FUEL” and filed on Aug. 27, 2015, which is specifically incorporated herein for all that it discloses and teaches. The present disclosure relates to nuclear fission reactors and fuel elements, particularly for fast reactors, such as breed and burn reactors. In particular, the present disclosure relates to tubular fuel elements in which a ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to total area of the interior volume in the cross-section of the tubular fuel element (or a smear density) varies with position along the longitudinal axis of the fuel element. This longitudinal variation can improve one or more of the operational profile, strain distribution, and burn-up distribution relative to the burn-up limit. In the discussion that follows, reference is made to certain structures and/or methods. However, the following references should not be construed as an admission that these structures and/or methods constitute prior art. Applicant expressly reserves the right to demonstrate that such structures and/or methods do not qualify as prior art against the present invention. In conventional nuclear reactors, fuel loading in the fuel elements is typically uniform along a longitudinal direction of the fuel assemblies. During irradiation in the reactor, the fuel expands due to, for example, the production of fission products and, in particular, fission products in the form of gas. The expanded fuel expands within the available space of the inner diameter of a cladding of an individual fuel element. However, over time and at higher burn-up values, the expansion of the fuel can strain the cladding, particularly where gas retention occurs and when fission products (gas or solid) begin to fill voids within the fuel. At this point, cladding strain may become proportional to burn-up and cladding strain begins to increase, quickly. This strain ultimately limits the life of fuel elements in the reactor core as expansion of the fuel cladding leads to decreased (sometimes non-uniform) coolant flow areas external to the cladding. The rate of strain is increased by the constant effect of radiation on the structural material (cladding and fuel assembly ducts). The fuel elements can expand enough to impart further strain on the duct wall of their associated fuel assemblies, which may become ‘jammed’ together due to the swelling and/or cause bowing of the fuel assembly. The fuel element swelling may sometimes cause cracks in the cladding which can lead to uncontrolled release of fission products and/or coolant interaction with the fuel. At least in part due to the resulting strain, the maximum burn-up of any particular fuel element can set the useful lifetime of a fuel element and/or the entire fuel assembly. Strain imparted on a cladding and fuel assembly influences the upper burn-up limit of fuel elements. A high temperature and high radiation environment can cause fuel to swell, which imparts a strain on the cladding over time. The time, pressure, and temperature conditions associated with cladding failure create a strain limit. Strain is sometimes also analogized with “burn-up” because fuel burn-up creates high temperatures and conditions that contribute to swell of the fuel. Limits on strain and/or burn-up are typically set by the fuel experiencing the greatest neutron flux. Flux typically takes on a shape (longitudinally and radially) that is based at least in part on fuel and poison distribution within a fuel element. Fuel distributions through the body of the fuel elements have traditionally been uniform, which produces an inverted cosine or a Gaussian flux distribution shape across the uniform distribution of fuel in the longitudinal direction. However, the uniform longitudinal distribution of fuel with attendant longitudinal distribution of flux underutilizes fuel at the longitudinal ends of the fuel element. The disclosed technology provides precise adjustments to the local conditions of fuel density to equalize fuel performance along the longitudinal length of the fuel elements, increase the average actual burn-up, increase the effective lifetime of the fuel element in the reactor core, equalize the fuel strain, and/or to provide more neutronic contribution (and achieve an increased number of fission reactions or “higher burn-up”) from traditionally under-utilized portions of the fuel resulting in increased fuel utilization efficiency. Adding more fuel at other than the longitudinal location of expected maximum burn-up (or strain) in a fuel element (for example, at the longitudinal ends) can reduce uneven burn-up and decrease the fuel strain. Adding more fuel between the longitudinal ends and the location of the maximum burn-up and/or strain can flatten and shift a burn-up distribution longitudinally within the fuel element which can, overall, generally even-out the burn-up distribution of a fuel element by increasing average burn-up and decreasing the fuel strain. This process is affected by varying the relative amount of fuel per unit volume area as longitudinal location in the fuel element varies along an axis oriented longitudinally, e.g., between a first end and a second end of the fuel element. The variation of this ratio (or smear density) with position along the longitudinal (or axial) axis can be repeated and/or varied among the various fuel elements constituting the fuel assembly and, ultimately the reactor, until the burn-up distribution is longitudinally spread out while maintaining peak core power, criticality of the reactor, etc. The net effect increases the average burn-up of the fuel and increases efficiency of the reactor design and/or operation. An example embodiment of a fuel element for a nuclear fission reactor comprises a fissionable composition comprising a fissionable nuclear fuel and may contain one or more non-fuel materials. Fissionable fuel may include any fissile fuel and/or fertile fuel and may be mixed, combined or included within the interior volume of the fuel element with other suitable materials, including neutron absorbers, neutron poisons, neutronically transparent materials, etc. Fissile material may include any suitable material for nuclear fission or neutron flux production including uranium, plutonium and/or thorium. Fertile fuel may be any appropriate fertile fuel including natural uranium, un-enriched uranium, etc. which can be bred up to fissile fuel. The fuel element is generally tubular and has a longitudinal axis. The fissionable composition occupies at least a portion of an interior volume of the tubular fuel element. The fissionable composition may be in thermal transfer contact with an interior surface of the fuel element. A ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to total area of the interior volume in the cross-section of the tubular fuel element varies with position along the longitudinal axis. In one aspect, the ratio can be expressed as a smear density percentage defined by: Smear ⁢ ⁢ Density ⁡ ( % ) = Area Fuel Area Interior ⁢ ⁢ Cross - Section × 100 where: AreaFuel=area of fissionable nuclear fuel in a cross-section of the fuel element perpendicular to the longitudinal axis of the fuel element, and AreaInterior Cross-Section=area of the interior of the cladding in the cross-section of the fuel element perpendicular to the longitudinal axis of the fuel element. While the fuel elements and related structures and methods disclosed herein will be particularly described with respect to a Traveling Wave Reactor (TWR), the fuel elements and related structures and methods disclosed herein are generally applicable to many types of solid fuel nuclear reactors and may be used in any appropriate nuclear reactor with solid fuel, such as a breed and burn reactor. As used herein, TWR means a type of breed-and-burn and/or breed-and-burn equilibrium nuclear reactor in which waves that breed and then burn may travel relative to the fuel and includes, without limitation, standing wave reactors. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, the use of similar or the same symbols in different drawings typically indicates similar or identical items, 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. 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 as limiting. The present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Disclosed below are a number of techniques for increasing fuel efficiency in a nuclear reactor core. Although some of these techniques are particularly useful in breed and burn reactors, the disclosed technology is also extendable to other types of solid-fueled fission reactors including without limitation fast reactors, breed and burn reactors, sodium cooled reactors, light-water reactors, heavy-water reactors, etc. As used herein, the term “burn-up,” also referred to as “% FIMA” (fissions per initial heavy metal atom) refers to a measure (e.g., a percentage) of fission that occurs in fissile fuel. For example, a burn-up of 5% may indicate that 5% of the fuel underwent a fission reaction. Due to a number of factors, burn-up may not occur evenly along the length of each individual fuel element in a fuel assembly. A fuel element is considered exhausted when a region of the fuel element has undergone enough burn-up to reach a burn-up limit, also sometimes referred to as a “peak burn-up” or “maximum burn-up.” When any one location reaches the burn-up limit, the entire fuel element is considered discharged even though only a portion of the fuel within that element has actually reached the discharge limit. Therefore, more energy can be obtained from an individual fuel element when average burn-up is higher along the full length of the fuel element than when average burn-up is lower. In contrast to peak or maximum burn-up, the term “actual burn-up” is used herein to refer to an amount of burn-up that has occurred within a defined area of the fuel assembly at the time when the fuel element is considered discharged because at least a portion of the fuel within the fuel element has reached the burn-up limit. Burn-up for a given amount of fuel can be influenced by several factors including, for example, neutron flux distribution (e.g., areas of high flux correspond to areas of high burn-up), poison distribution, temperature, coolant flow, heat rate, etc. Thus, a number of implementations disclosed herein generally relate to methods for increasing the average actual burn-up observed within each individual fuel element of a fuel assembly. Although the following description describes a fast sodium cooled reactor, this is for example purposes only and any solid fueled fission reactor may be used as appropriate. FIG. 1A illustrates an example nuclear fission reactor 130 with a nuclear reactor core 132. The nuclear reactor core 132 is disposed in a reactor vessel 140 surrounded by a guard vessel. The nuclear reactor core 132 typically contains a coolant, such as a pool of coolant (e.g., liquid sodium) or loops through which coolant may flow throughout the nuclear fission reactor 130. In addition, the nuclear reactor core 132 includes a number of elongated fuel assemblies (e.g., a fuel assembly 136) that may each include multiple fuel elements. In various implementations and/or operational states, the fuel assemblies may include fissile nuclear fuel assemblies (e.g., a set of starter fuel assemblies, etc.) and fertile nuclear fuel assemblies (e.g., a set of feed fuel assemblies, etc.), and/or fuel assemblies that includes both fertile and fissile fuel. Neutron distribution in the nuclear reactor core 132 is describable by a neutron flux profile 104. The neutron flux profile plot 104 illustrates example neutron flux magnitude as a function of vertical position within a fuel region 110 of the nuclear reactor core 132. Specific characteristics of this profile may vary dramatically from one implementation to another based on characteristics of the nuclear fission reactor 130 and local conditions within the nuclear reactor core 132. In FIG. 1A, neutron flux is greatest in a central region 112 and tapers off with distance from the central region 112 toward end regions 114 and 116 of the fuel region 110. For example, end regions 114 and 116 may correspond to opposing ends of a fuel element or a group of fuel elements within the nuclear reactor core 132. In the illustrated example, the effect of increased neutron flux in the central region 112 is attributable to one or more factors influencing local conductions such as flux and/or temperature. For example, neutron-absorbing control rods, such as control rod 118 may, in operation, be inserted into the fuel region 110 from the direction of an upper end of the nuclear reactor core 132 toward the central region 112. This may have the effect of reducing neutron flux in the upper end of the fuel region 110, as shown. In other implementations, control rods may be positioned differently to influence neutron flux in a different manner via the same or similar principles. The neutron flux profile 104 also illustrates unequal average flux and unequal rates of increasing flux in end regions 114 and 116. In one implementation, this effect is due to dissimilar local temperatures in the end regions 114 and 116, respectively. In general, power output is proportional to a mass flow rate multiplied by a difference in temperature between fuel and coolant. Therefore, if a coolant entrance point is, for example, proximal to the end region 116 and a coolant exit point is proximal to the end region 114, a difference in temperature may be larger at the coolant entrance point than the coolant exit point, while the mass flow rate remains approximately uniform at both locations. Consequently, neutron flux (e.g., power output) may be greater near the end region 116 (e.g., the coolant entrance point) and/or may increase more quickly within the end region 116 than within the end region 114. This unequal flux distribution contributes to unequal strain on the individual fuel elements and may cause localized peak burn-up in the regions with greatest neutron flux and/or strain. Consequently, the average burn-up across the full length of each fuel element may be significantly less than the peak burn-up. If, however, the effects of unequal flux and/or strain can be negated or reduced, the average burn-up across the length of each fuel element is closer to the peak burn-up and higher fuel efficiency is realized. In one implementation, this is accomplished by manipulating a smear density (e.g., a ratio of fissionable material to cross-sectional area) internal to each fuel element to increase average burn-up at a plurality of locations along the longitudinal axis. Techniques for manipulating smear density for improved fuel efficiency are described in detail with respect to the following figures. Fuel elements are typically long, slender bodies including a thin-walled, outer jacket (also called cladding) and a fissionable composition (including fissionable nuclear fuel) within the cladding. Depending on the design of the nuclear reactor, multiple fuel elements are typically co-located into a fuel bundle or fuel assembly, and multiples of fuel assemblies are included in the nuclear reactor. The geometric shape of the fuel element can be any suitable shape designed for the physical and design constraints of the fuel assembly and the nuclear reactor. Two example embodiments of fuel elements are illustrated in FIGS. 1B and 1C, in perspective view. A first embodiment of a fuel element 10 has a general tubular shape (FIG. 1B) with an elongated longitudinal axis relative to the radial axis. The tubular shape may be solid throughout, as shown, or may include gaps between the cladding and fissionable composition, gaps or interstitial spaces between the bodies of the fissionable composition, and/or one or more hollow voids within the fissionable composition body (voids). In some cases, the fuel element 10 has a cross-sectional shape, in a plane perpendicular to the longitudinal axis 20 that is rectilinear. In FIG. 1B, the rectilinear shape is a regular polygon such as a square, but the cross-sectional shape can be any rectilinear shape or appropriate polygonal shape including triangular, hexagonal, octagonal, dodecagonal, etc. The longitudinal axis 20 is oriented along the major axis of the general tubular shape and extends longitudinally a distance L1. Typically, the longitudinal axis will extend between a first wall or end 30 and a second wall or end 40 of the cladding. In some examples, side walls 50 join the first end 30 to the second end 40 to define an interior volume and form the general tubular shape of the fuel assembly 10. A second embodiment of a fuel element 10′ has a general tubular shape (FIG. 1C) with an elongated longitudinal axis relative to the radial axis. The tubular shape may be solid throughout, as shown, or may include gaps between the cladding and fissionable composition, gaps or interstitial spaces between the bodies of the fissionable composition, and/or one or more hollow voids within the fissionable composition body (voids). In some cases, the fuel element 10′ has a cross-sectional shape, in a plane perpendicular to the central, longitudinal axis 20′, that is curved, either concave or convex or combinations thereof. In FIG. 1C, the curved shape is a circle, but the cross-sectional shape can be any ellipse with an eccentricity, e, that varies between 0 and less than 1, i.e., 0≤e<1, alternatively e is equal to or less than 0.5, or any complex curved shape, regular curved shape such as a hyperboloid or a shape approximating a hyperboloid. The longitudinal axis 20′ is oriented along the major axis of the general tubular shape and extends longitudinally a distance L2. Typically, the longitudinal axis will extend between a first wall or end 30′ and a second wall or end 40′ of the cladding. In some examples, side wall 50′ joins the first end 30′ to the second end 40′ to define an interior volume and form the general tubular shape of the fuel assembly 10′. Although the examples shown in FIGS. 1B and 1C show solid walls and ends enclosing the interior volume, it should be appreciated that in some cases the cladding or other structure defining the interior volume of the fuel element may be partial enclosures, solid portions of the fuel element, etc. The cladding of the fuel element can be any suitable material and may reduce corrosion of the fuel and/or the release of fission products while also having a low absorption cross section for neutrons. In one embodiment, the cladding layer may include at least one material chosen from a metal, a metal alloy, and a ceramic. In one embodiment, the cladding may contain a refractory material, such as a refractory metal including at least one element chosen from Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, Nd, and Hf. In another embodiment, the cladding may contain a metal alloy, such as steels including steels of different compositions and microstructures. Non-exhaustive examples of suitable materials for the cladding include ceramics, such as silicon-carbide (SiC) and titanium-metal or metalloid-carbides (Ti3AlC and Ti3 SiC2) and metals, such as aluminum, stainless steel, and zirconium alloys. Other materials suitable for cladding may be used as appropriate. The cladding may be constructed in two or more layers or regions or include one or more liners of different materials. Examples of appropriate cladding including liners are described in U.S. patent application Ser. No. 13/794,633, filed Mar. 11, 2013, titled “Nuclear Fuel Element”, which is incorporated herein by reference. The fissionable composition can include fissionable nuclear fuel and non-fuel materials. The fissionable nuclear fuel comprises at least about 88 wt % fissionable fuel, e.g., at least 94 wt %, 95 wt %, 98 wt %, 99 wt %, 99.5 wt %, 99.9 wt %, 99.99 wt %, or higher of fissionable fuel. Lower smear densities may be appropriate in other circumstances and within the scope of the invention of selectively varying the smear density along the longitudinal length of the fuel element. Fissionable nuclear fuel includes fissile nuclear fuel, fertile nuclear fuel (which can be ‘bred up’ to fissile fuel) and/or a mixture of fissile and fertile nuclear fuel compositions. The fissionable nuclear fuel of the fissionable composition can include any nuclide that is capable of undergoing fission after capturing either high-energy (fast) neutrons or low-energy thermal (slow) neutrons. Fissionable fuel may be of a metal form, an oxide form, or a nitride form and may contain a metal and/or metal alloy. Examples of suitable fissionable nuclear fuel include one or more radioactive elements (including their isotopes) selected from the group comprising uranium, plutonium, thorium, americium and neptunium and isotopes and alloys thereof. In one embodiment, the fuel may include at least about 88 wt % U, e.g., at least 94 wt %, 95 wt %, 98 wt %, 99 wt %, 99.5 wt %, 99.9 wt %, 99.99 wt %, or higher of U. The non-fuel material (if present) of the fissionable composition can include any suitable non-fuel material which may have any feature desirable within the design of the fuel element, including bonding of the nuclear fuel material, neutron transparent or absorbent characteristics, coolant properties, thermal transmission properties, etc. For example, the non-fuel material may include a refractory material, which may include at least one element chosen from Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, and Hf. In one embodiment, the non-fuel material may include additional burnable poisons, such as boron, gadolinium, or indium. Fuel elements containing a fissionable composition utilizing fissionable nuclear fuel can have a ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to of the interior of the cladding in the cross-section of the tubular fuel element that varies with position along the longitudinal axis. Such ratio can be selectively varied at multiple different locations along the longitudinal axis to selectively design a strain profile for the fuel element that is achieved by targeting expected strain and/or burn-up for the expected flux and fuel assembly movement over the fuel cycles of the fuel assembly lifetime. The strain profile may account for or consider any one or more of the effects of nuclear reactor operations, such as material strain, temperature and pressure effects, production of reaction by-products and gas release fractions, within the design envelope of the fuel element, fuel assembly and nuclear reactor and additionally may also account for or maintain selected reactor core characteristics which may include any one or more of core criticality (keff>1), peak core power, etc. Despite burn-up alone not being sufficient to determine strain definitively without additional information including operational temperature, reaction by-product, geometries of components, etc., burn-up is highly correlated to strain, and is a potential proxy for strain. One of skill in the art, however, will understand that burn-up limit is an appropriate proxy for strain in some cases, and strain is a possible proxy for burn-up limits in some cases. Some examples below are described using burn-up limits and other examples are described using strain profiles. One of skill in the art will recognize that either burn-up limits or strain limits, or any other suitable parameter or reactor core characteristic may be used as a threshold and/or profile, such as measure of fission products, measure of deformation, temperature, etc. As noted above, selective adjustment of the smear density over the length of the fuel element can tailor local areas of the fuel element to expected local conditions, which may include one or more of burn-up, heat rate, coolant flow and/or temperature, displacement per atom (DPA), DPA history and/or rate, etc. along the longitudinal length of the fuel elements. Increasing the average actual burn-up can increase the effective lifetime of the fuel element in the reactor core, allow for efficient use of nuclear fuel, and/or allow the fuel element to have more neutronic contribution (and achieve higher burn-up) as compared to traditional reactors with under-utilized portions. The ratio of fissionable material to cross-sectional area can vary with position along the longitudinal axis of the fuel element, which as noted above can reduce detrimental effects (burn-up and/or strain) and may increase the utilization of fuel over time along the longitudinal axis. The ratio can be expressed as a smear density of the fuel element. For example, the smear density percentage can be defined by: Smear ⁢ ⁢ Density ⁡ ( % ) = Area Fuel Area Interior ⁢ ⁢ Cross - Section × 100 where: AreaFuel=area of fissionable nuclear fuel in a cross-section of the fuel element perpendicular to the longitudinal axis of the fuel element, and AreaInterior Cross-Section=area of the interior of the cladding in the cross-section of the fuel element perpendicular to the longitudinal axis of the fuel element. In the above definition of smear density, fissionable nuclear fuel means any material that is capable of undergoing fission including fissile nuclear material and fertile nuclear material but not including non-fuel material. Thus, the smear density is a function or ratio of the amount of fissionable material in a defined area which can be affected by the presence of non-fissionable (non-fuel) material including bonding materials, voids in the fuel element which can contain substantial vacuums, gaseous non-fissionable material, etc. Smear density in the above equation is expressed as a percentage (%); smear density can alternatively be expressed mathematically as a fraction, in which case it is called fractional smear density. In some cases, smear density can vary substantially continuously, such that the smear density at any one location along the longitudinal length may vary from the immediate cross-sectional area adjacent to the smear density being calculated. In some cases it may be appropriate to indicate some sectional or zonal volume averaged smear density with an indication of rate of change of smear density, rather than representing smear density at any one particular cross-sectional area. FIGS. 2-6 show cross-sectional, schematic views taken along sections A-A to E-E, respectively, of the fuel element 10 shown in FIG. 1A. The cross-sectional views 200, 300, 400, 500, 600 are perpendicular to the longitudinal axis 20 of the fuel element 10 and schematically illustrate an example distribution of fissionable composition which is shown as pellets. The interstitial spacing between pellets can optionally contain non-fuel material such as bonding material, a substantial vacuum, gaseous materials, or any suitable material. The cross-sectional views also graphically illustrate the area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis and the total area of the interior volume in the cross-section of the tubular fuel element that can be used to determine the ratio of these two areas that varies with position along the longitudinal axis. The ratio, which can be expressed as a smear density, at each of the five indicated locations along the longitudinal axis 20 of the fuel element 10 are shown in the noted figures. For illustrative purposes, sections A-A and E-E are located toward longitudinal ends of the tubular fuel element and section C-C is located proximate the longitudinal center of the tubular fuel element; sections B-B and D-D are located approximately midway between the respective longitudinal end section and the longitudinal center section. The use of five sections is illustrative and non-limiting and a tubular fuel element can have an increasingly large number of cross-sections having discrete ratios (or smear densities) as the locations along the longitudinal axis are parsed to thinner and thinner sections, which in the aggregate produce a step-wise varying or substantially continuously varying change in the ratio along the longitudinal axis. In the schematic views in FIGS. 2-6, the fissionable composition including the fissionable nuclear fuel is formed as pellets. FIG. 7 is a cross-sectional schematic view of an example embodiment of a fuel pellet 700. The fuel pellet 700 includes a fissionable composition 705 including fissionable nuclear fuel. The fissionable composition of the fuel pellet can optionally contain non-fuel material. The fuel pellet 700 can optionally include one or more of an outer coating 710 and/or filler material 715. The outer coating may be any suitable material and in some cases may include a material to limit diffusion or interaction between the fuel composition and either or both the coolant, the cladding, etc. Example outer coating materials are described further below with reference to liners of the cladding. The filler material (as a form of non-fuel material) differs from the fissionable nuclear fuel. The filler material can fill a void space in a structure of the fuel material and/or be distributed or intermixed with the fuel material and can include a gas, liquid, solid, vacuum, bonding material, neutronic poison, or any suitable non-fuel material for the nuclear fuel element. Returning to the cross-sectional views in FIGS. 2-6, a plurality of fuel pellets 210, 310, 410, 510, 610 are distributed in the interior volume of the tubular fuel element. The fuel pellets are in thermal contact with each other and with the interior surface of the cladding. Interstices formed between the pellets and between the pellets and the cladding provide a volume to accommodate, at least initially, fission products and, in particular, fission products in the form of gas, that form during irradiation. As a result, the interstices contribute to reduce mechanical forces, in particular tensile forces, exerted on the cladding. The fuel pellets are in heat transfer contact with the cladding. For example, a mechanical contact bond allows heat generated from nuclear reactions to be transferred to the cladding, primarily by conduction either directly or through non-fuel materials. Additionally or alternatively, a medium, such as a solid or fluid (including for example liquid metal or a gas), can be located in the spaces and can provide an additional conduction path for heat transfer from the pellet to the cladding via a solid or even a fluid such as a liquid metal bond or gas bond. In additional or alternative embodiments, the fuel element cladding can include a liner on the interior surface of the cladding which separates the fuel (whether in pellet form, sponge form or other form) from the outer cladding material. In some cases, particularly at high burn-ups, the contents of the fuel and the cladding may tend to diffuse, thereby causing undesirable alloying and/or degrading the material of the fuel and the cladding (e.g., by de-alloying of the contents of the fuel and/or cladding layer or forming a new alloy with degraded mechanical properties). A liner may serve as a barrier layer between the fuel and the cladding material to mitigate such interatomic diffusion. For example, one or more liners of the cladding may be employed to mitigate interatomic diffusion between the elements in the fissionable composition of the fuel and the cladding material to avoid, for example, degradation of the fuel and/or cladding material. The liner may contain one layer or multiple layers. In the case where the liner contains multiple layers, these layers may contain the same or different materials and/or have the same or different properties. For example, in one embodiment, at least some of the layers may include the same material while some include different materials. In one example, the outer cladding material may comprise at least one material chosen from a metal, a metal alloy, and a ceramic. In one embodiment, the cladding may contain a refractory material, such as a refractory metal including at least one element chosen from Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, Nd, and Hf. A metal alloy in a cladding layer may be, for example, steel. The steel may be chosen from a martensitic steel, an austenitic steel, a ferritic steel, an oxide-dispersed steel, T91 steel, T92 steel, HT9 steel, 316 steel, and 304 steel. The steel may have any type of microstructure. For example, the steel may include at least one of a martensite phase, a ferrite phase, and an austenite phase. The cladding may include a first liner adjacent to the outer cladding material and may comprise a material selected to reduce interatomic diffusion and/or interaction with the outer cladding and a second liner. The second liner, adjacent to the first liner and forming the interior surface for containing the fuel composition, may comprise a material selected to reduce interatomic diffusion or interaction with the first liner and the fuel composition. An example of a suitable material for use as a liner includes one or more alloys including Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, Nd and Hf. The liner of the cladding may be deposited on the interior surface of the outer cladding material, may be deposited on the outer, cladding facing surface of the fuel composition, and/or may be a layer of material disposed between the outer cladding material and the fuel composition and/or in any suitable manner. In the three-dimensional fuel element, the plurality of fuel pellets occupy at least a portion of the interior volume defined by the cladding. However, in the two-dimensional cross-sectional views in FIGS. 2-6, the plurality of fuel pellets 210, 310, 410, 510, 610 occupy a portion of the interior area defined by the inner surfaces 220, 320, 420, 520, 620 of the walls 230, 330, 430, 530, 630 that constitute the cladding 240, 340, 440, 540, 640 and an interstitial space 250, 350, 450, 550, 650 is formed between the fuel pellets themselves and between the fuel pellets and the inner surface of the cladding. Typically, the interstitial space can be occupied by a non-fuel material such as a bonding and/or cooling material, such as a solid or fluid which in some cases may comprise a liquid metal or a gas forming, respectively, a liquid metal bond or a gas bond. A further characteristic of the illustrated fuel element is that there is thermal conduction contact (one form of thermal transfer) between at least some of the fuel pellets 210, 310, 410, 510, 610 and walls 230, 330, 430, 530, 630 both directly by fuel pellet to cladding (or fuel pellet to liner) contact as well as indirectly through the bonding material in the spaces 250, 350, 450, 550, 650. In any one cross-sectional view, the portion of the area occupied by the fissionable material of the plurality of fuel pellets 210, 310, 410, 510, 610 is the value for AreaFuel in the smear density equation and the area defined by the inner surfaces 230, 330, 430, 530, 630 of the walls 240, 340, 440, 540, 640 that constitute the cladding 250, 350, 450, 550, 650 is the value for AreaInterior Cross-Section in the smear density equation. If a liner is present in the cladding, then the value for AreaInterior Cross-Section is determined using the innermost surface of the liner, i.e., the surface exposed to the interior volume. The ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to area of the interior of the cladding in the cross-section of the tubular fuel element can vary with position along the longitudinal axis by any suitable means and can be selectively designed or formed based on the presence and amount of any non-fissionable material including any one or more of the interstitial spaces between the fuel pellets, voids within the pellets, other non-fuel material in the pellet, etc. or in any other suitable manner. For example, the longitudinal variation can be achieved by varying the fuel loading of the fissionable composition. The fissionable composition, such as in each pellet, can vary (by, for example, varying the relative amounts of one or more of fissionable material and non-fuel material). The non-fuel material includes one or more of interstitial spaces, voids within the fuel composition, filler material, bonding material, coolant, etc. In this example, the amount of non-fuel material within the fuel composition can be selectively varied to achieve the desired fuel loading. The pellets of varying composition can then be disposed longitudinally in the tubular fuel element to establish the desired longitudinal variation in the ratio (or smear density). In an alternative or additional example, the longitudinal variation can be achieved by varying the size of the bodies containing the fissionable composition. The geometric size of the fuel pellet (or other body shape of the fissionable composition) can change and can take the form of any one or more of powders, particulates, slugs, etc. The size and shape of the pellet may be designed and manufactured based on the desired smear density including designing both the volume and cross-sectional area of fissionable material and non-fuel material including voids within the pellet, interstitial spaces between the pellets, fuel loading, etc. given the fuel element cladding interior volume, shape and size. Such pellets can then be loaded at a higher or lower cross-sectional density at longitudinal locations in the tubular fuel element to establish the desired longitudinal variation in the ratio (or smear density). Also, depending on shape of the pellet, larger pellets may reduce the loading density as a result of the contact between pellets leaving a larger area or a smaller area for the interstitial space between the pellets. In this regard, at a close packed density for a given spherical pellet size, a smaller spherical size will have smaller spaces between contacting pellets than for a larger spherical size. As an alternative or additional example, a mixture of sizes and/or shapes can further selectively reduce the spaces between contacting pellets as a result of suitably sized smaller pellets occupying the spaces between contacting larger pellets. Compacting or settling of the pellets (whether of the same or different sizes and/or shapes) within the cladding can be accomplished through any suitable method including, but not limited to, mechanical (including tamping, pressure, air blowing), vibration, etc. In additional or alternative examples, non-fuel material may be intermixed or distributed with the fissionable material within the fuel pellet and/or non-fuel material may be provided in separate pellets. The non-fuel material pellets may be intermixed with the fissionable composition pellets to provide the desired smear density for that particular location along the longitudinal length of the fuel element. Additionally or alternatively, voids (whether filled with vacuum or other nonfuel material) within the fuel pellets can provide a mechanism to vary fuel loading in the fuel element. In a further example, the longitudinal variation can be achieved by distributing the pellets in a medium. A structure or material in the fuel element can provide a medium for distribution of the pellets or other bodies containing the fissionable composition. A sufficiently dense bonding material or non-gaseous bonding material can be used to fill the space between pellets. Such a bonding material can be used in suitable quantity to vary the distance between pellets (and between pellets and the inner surfaces of the walls) and thereby obtain a higher or lower cross-sectional density at longitudinal locations in the tubular fuel element to establish the desired longitudinal variation in the ratio (or smear density). Any of the above methods or combinations of one or more of the above methods to achieve the variation can be used to achieve different and/or more complex gradations in the longitudinal distribution of the ratio or smear density. For example, varying the fuel loading of the pellet can be combined with varying the size of the bodies containing the fissionable composition. FIGS. 2-6 schematically present one example in which smear density can vary with position along the longitudinal axis within a single fuel element, e.g., fuel element 10 of FIG. 1B. As observable from FIGS. 2-6, the smear density varies across section A-A (FIG. 2), section B-B (FIG. 3), section C-C (FIG. 4), Section D-D (FIG. 5) and section E-E (FIG. 6) and is illustrative of the ratio (or smear density) varying with position along the longitudinal axis of the fuel element. Overall, in example embodiments, the smear density can vary with position along the axis between a minimum of 30%, alternatively 50%, and a maximum of 100%, alternatively 75% or 70%. As an additional or alternative modification, the smear density at each longitudinal end of the tubular fuel element can be greater than the smear density at a longitudinal center of the tubular fuel element. FIGS. 2-6 illustrate such an example embodiment in that comparing FIG. 2 (a cross-sectional view at a first longitudinal end) and FIG. 6 (a cross-sectional view at a second longitudinal end) with FIG. 4 (a cross-sectional view at a longitudinal center), one can readily ascertain that the area occupied by the fuel pellets 210, 610 at the first longitudinal end and second longitudinal end is greater than the area occupied by the fuel pellets 410 at a longitudinal center. In an additional modification, the smear density at a first longitudinal end of the tubular fuel element is greater than the smear density at a second longitudinal end of the tubular fuel element. FIGS. 2-6 also illustrate this example embodiment in that comparing FIG. 2 (a cross-sectional view at a first longitudinal end) with FIG. 6 (a cross-sectional view at a second longitudinal end), one can ascertain that the area occupied by the fuel pellets 210 at the first longitudinal end is greater than the area occupied by the fuel pellets 610 at the second longitudinal end. FIGS. 8-10 illustrate, in a side cross-sectional view of a portion of the fuel element, variation of the density of fuel pellets (fuel pellets per volume) along the longitudinal length of the fuel element by changing the relative amount of fuel pellets (and their corresponding amount of fuel composition with optional non-fuel materials) and non-fuel components (whether bond material, voids, or other non-fissionable material) used to fill the interior volume during the manufacturing of the fuel element. In the cross-sectional schematic views, fuel pellets 810 are located within the interior volume of the fuel element 800 between walls 820 of the cladding 830. To achieve a lower smear density, the fuel pellets can be distributed within the interior volume with the majority of fuel pellets being discrete fuel pellets (meaning that the fuel pellet does not directly, mechanically contact another fuel pellet) although some fuel pellets can be in contact with other fuel pellets, as seen in FIG. 8. Also as seen in FIG. 8, fuel pellets 810 may either be in contact with the interior surfaces 840 of the walls 820 or be at a distance from the interior surfaces 840 of the walls 820 or a mixture of such positioning. A bond material 850, such as helium gas or liquid sodium or solid, and/or any other suitable non-fuel material are located in the space surrounding and between fuel pellets 810 and is located between fuel pellets 810 and the interior surfaces 840 of the walls 820. The bond material can provide, at least in part, thermal transfer contact between the fuel pellets and the cladding. To achieve a higher smear density, the fuel pellets can be distributed within the interior volume with a higher proportion of fuel regions to non-fuel regions. For example, the majority of fuel pellets or alternatively all of the fuel pellets can be in contact with an adjacent fuel pellet, although there can be some spaces separating adjacent fuel pellets to accommodate the packing arrangement of the fuel pellets within the interior volume, as seen in the fuel element 800′ in FIG. 9. Also as seen in FIG. 9, fuel pellets 810 may be in contact with the interior surfaces 840 of the walls 820 although some fuel pellets 810 may be at a distance from the interior surfaces 840 of the walls 820 and there is a mixture of such positioning. Contact between adjacent fuel pellets 810 can be promoted by, for example, vibrational techniques, mechanical techniques (such as tamping, fluid flow, etc.), although packed fuel pellets are not necessarily in a close-packing arrangement. In FIG. 9, contact between adjacent fuel pellets in a first area 860 is in close-packing arrangement and contact between adjacent fuel pellets in a second area 870 is in packed, but not in close-packing arrangement. A bond material 850 which can be any suitable solid or fluid bonding material such as a liquid metal cooling media, such as liquid sodium, is located in the space surrounding and between fuel pellets 810 and is located between fuel pellets 810 and the interior surfaces 840 of the walls 820. In some cases, the size and/or shape of the fuel pellet and/or non-fuel material pellet may be selected to ensure regular, predictable fitting within the given dimension (size and shape) of the cladding interior. For example, some spherical and/or regular polygons may have a predictable and selected nesting or structure when introduced to or disposed within the interior of the cladding, such as that illustrated in FIG. 10. In an additional or alternative example, the amount of fissionable nuclear fuel per unit volume area, i.e., the fissionable nuclear fuel loading, can vary and the longitudinal variation in smear density can be achieved by using pellets of higher or lower fissionable nuclear fuel loading. The fissionable nuclear fuel loading can be determined and selected to vary longitudinally along the length of the fuel element on a quantized basis or on a continuous basis, depending on the manner in which pellets of different fissionable nuclear fuel loading were manufactured. Using such pellets, a variation in smear density can be achieved by placing fuel pellets with different fissionable nuclear fuel loading in different locations or zones along the longitudinal length of the fuel element. For example, pellets with higher fissionable nuclear fuel loading can be located at longitudinal locations which are to have higher smear densities and pellets with lower fissionable nuclear fuel loading can be located at longitudinal locations which are to have lower smear densities. FIG. 10 schematically illustrates another embodiment of a fuel element 800″ with fuel pellets 880a, 880b of different fissionable nuclear fuel loading that have been located in different strata of the interior volume to thereby form quantized zones A, B along the longitudinal length of the fuel element. In the FIG. 10 embodiment, fuel pellets 880a have a lower fissionable nuclear fuel loading than fuel pellets 880b and zone A has a lower smear density than zone B. Although described herein is some instances as a pellet, a pellet is not limited to a spherical geometric shape, but can be any volumetric shape including oblong, ovoid, cylindrical, rod-like, conical, rectilinear or any other closed, volumetric shape as long as the fissionable composition in each such geometric shape can be quantified and, if desired, selectively modified to any suitable shape. For example and in additional or alternative embodiments, the fissionable composition may be contained within a body having the shape of an annular slug. FIGS. 11A and 11B illustrate such a shape in a top view (FIG. 11A) looking along the longitudinal axis and in a side view (FIG. 11B) looking along a radial axis. The annular slug 900 includes a body 905 having outer walls 910 connecting longitudinally separated top surface 915 and bottom surface 920. The annular slug 900 has a center passage 925 from the top surface 915 to the bottom surface 920. In the embodiment shown in FIGS. 11A-B, the diameter of the outer walls is substantially constant along the longitudinal length and the amount of fissionable material along the longitudinal length may be varied in any suitable manner by varying fuel loading, amount of non-fuel material, and/or adjusting voids or interstitial spacing. In additional or alternative embodiments, such as shown in FIGS. 12A and 12B, the diameter of the outer walls can vary. FIGS. 12A and 12B show one option for varying the outer wall of the body. In FIGS. 12A and 12B, the outer wall 910′ of the body 905′ of the annular slug 900′ has, relative to the longitudinal center C′, a diameter D′ that decreases as the outer wall 910′ extends longitudinally from the top surface 915′ and bottom surface 920′ toward a middle region 930. Such outer walls 910′ can be linear or have a concavity or convexity or have a combination of such surfaces. In addition, the outer walls 910′ with decreasing diameter D′ in the middle region 930 can impart a shape to the body 905′ along the longitudinal length of each slug. For example, the body may have an exterior surface with a reduced central diameter in a central portion compared to a first diameter of the body at a first end portion and compared to a second diameter at a second end portion. Such a shape can, in some embodiments, be the shape of a hyperboloid or of a hyperbolic paraboloid. In additional or alternative examples, the exterior surface can vary not only along the longitudinal length of each body but, additionally or alternatively, the composite effect of multiple slugs arranged in the fuel element can impart such a shape along the longitudinal length of the fuel element. Variation of the diameter or width of the fissionable material in the slug provides another mechanism of varying fissionable material fuel loading along the longitudinal length of the fuel element by utilizing slugs of varying shapes selectively located along the longitudinal length of the fuel element to impart a desired shape, such as the shape of a hyperboloid or of a hyperbolic paraboloid. Although the passages 925, 925′ shown in FIGS. 11A-B and 12A-B are cylindrical and located symmetrically about the longitudinal axis of the body, the passage can be any suitable shape and located at any position within the body of the annular slug, either disposed off-center of the longitudinal axis in a radial direction, disposed as multiple passages, and/or disposed in some longitudinal volume of the slug and not present in other longitudinal volumes of the slug. It is to be appreciated that any one or more of the preceding may be used as appropriate. Additionally or alternatively, the radius of the passage 925, 925′ can vary similarly and/or differently from the outer wall 910, 910′ of the slug. The fissionable composition is disposed within the body 905,905′ of the annular slug 900,900′ and may be of uniform composition throughout the body 905,905′ or can vary in either or both the longitudinal and radial directions. Multiple annular slugs 900,900′ can be positioned, e.g., stacked, in the interior volume of the tubular fuel element. The passages 925, 925′ of the slugs 900, 900′ when stacked into a fuel element can be selectively aligned or misaligned as may be appropriate. The dimensions of the diameter (D,D′) of the outer diameter walls and the dimensions of the diameter (d,d′) of the passage can vary. Further, varying one or more of these dimensions varies the volume of the body of the slug and, by varying the volume amongst a plurality of slugs in a tubular fuel element, such variation in body volume can be used to adjust the fuel loading for at least a portion or all of the desired longitudinal variation in the ratio (or smear density). Alternatively or additionally, the longitudinal variation in the ratio (or smear density) can be established by varying the composition within one or more bodies or by varying the dimensions of the diameters (D,d or D′,d′) in one or more bodies. Although the slugs shown in FIGS. 11A-B and 12A-B show the ends of the slugs as substantially perpendicular to the longitudinal axis, it is to be appreciated that either or both ends of the slug may be non-perpendicular to the longitudinal axis, and/or may be any appropriate shape to either selectively fit with adjacent slugs and/or create selected interstitial spaces to selectively adjust the smear density. The bodies having the shape of an annular slug can be arranged within the fuel element in any suitable manner. FIGS. 13A-B are schematic cross-sectional views of a portion of a fuel element showing the annular slug bodies, such as those in FIGS. 11A-B and FIGS. 12A-B. In the illustrated embodiment, the bodies 905,905′ are in stacked longitudinal relationship within the walls 930,930′ of the cladding of the fuel element. One or more gaps 935,935′ may be present between the bodies, e.g., the fuel material, and the cladding, though gaps are not required. In some embodiments, the gap is filled with a non-fuel material. In one embodiment the non-fuel material is a pressurized atmosphere, such as a pressurized helium atmosphere; in another embodiment, the gap is filled with a liquid metal, such as sodium. In some cases, the smear density may differ between different fuel pellets or slugs. Although the following example is discussed with respect to slugs, as a form of pellets, one of skill in the art will recognize that any form of pellet may utilize the below techniques. For example, the smear density of slug 905b of FIG. 13A may be greater than the smear density of slug 905a. In some cases, if the difference between the smear density of adjacent slugs exceeds a smear density difference threshold, an additional amount of strain may occur at or in the region of the point or the plane of contact between the two different slugs (e.g., location 940 in FIG. 13A). In some cases a minimum smear density threshold may be used to limit the different smear densities in adjacent slugs. Alternatively and/or additionally, a localized area of either or both slugs 905a, 905b proximate the other slug may have their smear densities slighter greater (as in the case of smear density of slug 905a) and/or slightly less (as in the example of slug 905b). Additional slugs may be added in between slugs 905a, 905b to affect the alternate smear densities, and/or the smear density within the bodies of either or both slugs 905a, 905b may be locally adjusted either continuously and/or in a segment of the body of the slug proximate the localized contact with the other slug. Such localized adjustment may be used as appropriate to limit the rate of change of the smear densities and/or limit a localized difference between smear densities. Manufacturing of slugs and/or pellets can be performed in any suitable manner, including any one or more of extrusion, cutting or crushing larger solids forming smaller solids, molding, powder consolidation with or without sintering, removing selected areas of formed materials (such for voids, shape, size, annular passages, etc.) such as by mechanical means for example, drilling, machining, grinding, etc., chemical means, for example, dissolving, reacting, converting, decomposing, etc., or combinations thereof. FIGS. 14-18 show cross-sectional, schematic views taken along sections A′-A′, B′-B′, C-C′, D′-D′, and E′-E′, respectively, of the fuel element 100 shown in FIG. 1C. The cross-sectional views 1000, 1100, 1200, 1300, 1400 are perpendicular to the longitudinal axis 120 of the fuel element 100 and schematically illustrate the distribution of fissionable composition and void space. The cross-sectional views also graphically illustrate the area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis and the total area of the interior volume in the cross-section of the tubular fuel element that can be used to determine the ratio of these two areas that varies with position along the longitudinal axis. The ratio, which can be expressed as a smear density, at each of the five indicated locations along the longitudinal axis 120 of the fuel element 100 is shown in FIGS. 14-18. For illustrative purposes, sections A′-A′ and E′-E′ are located toward longitudinal ends of the tubular fuel element and section C′-C′ is located proximate the longitudinal center of the tubular fuel element; sections B′-B′ and D′-D′ are located approximately midway between the respective longitudinal end sections and the longitudinal center section. The use of five sections is illustrative and non-limiting and a tubular fuel element can have an increasingly large number of cross-sections having discrete ratios (or smear densities) as the locations along the longitudinal axis are parsed to thinner and thinner sections, which in the aggregate produce a step-wise varying or continuously varying change in the ratio along the longitudinal axis. It is also to be appreciated that the number of sections of similar smear densities may include one or more individual bodies (such as slugs) of similar smear densities; it is also to be appreciated that the plurality of bodies (such as slugs) in a section of similar smear density can achieve the smear density in any single or combination of techniques described above such as any one or more of fuel loading, amount of non-fissionable material, interstitial spaces, voids in the fuel, width of the slug, etc. In the schematic views in FIGS. 14-18, the fissionable nuclear fuel is in a form of a metal sponge. The metal sponge 1010, 1110, 1210, 1310, 1410 includes a matrix 1020, 1120, 1220, 1320, 1420 of the fissionable nuclear fuel (comprising a fissionable nuclear fuel and a non-fuel material) and a plurality of non-fuel regions 1030, 1130, 1230, 1330, 1430 in the matrix 1020, 1120, 1220, 1320, 1420. The non-fuel regions are typically non-fuel solid material and/or voids that are a substantial vacuum or gas filled, such as with helium. In the three-dimensional fuel element, the metal sponge occupies at least a portion of the interior volume enclosed by the cladding. However, in the two-dimensional cross-sectional views in FIGS. 14-18, the metal sponge 1010, 1110, 1210, 1310, 1410 occupies a portion of the interior area defined by the inner surfaces 1040, 1140, 1240, 1340, 1440 of the wall 1050, 1150, 1250, 1350, 1450 provided by the cladding 1060, 1160, 1260, 1360, 1460. A further characteristic of the illustrated fuel element is that there is thermal conduction contact (one form of thermal transfer) between at least a portion of the metal sponge 1010, 1110, 1210, 1310, 1410 and the wall 1050, 1150, 1250, 1350, 1450 (or with a liner of the cladding or other coating layer, where present). Such thermal conduction contact can be present over greater than 50%, alternatively greater than 70%, greater than 80%, greater than 90% up to 95%, 98% or 99%, of the perimeter of the metal sponge 1010, 1110, 1210, 1310, 1410. Where the metal sponge does not make mechanical contact, the space between the metal sponge 1010, 1110, 1210, 1310, 1410 and the wall 1050, 1150, 1250, 1350, 1450 (or with a liner of the cladding or other coating layer, where present) can be occupied by a non-fuel material such as a liquid metal or a gas forming, respectively, a liquid metal bond or a gas bond. Although FIGS. 14-18 are shown at constant diameter with no annular passage, it is to be appreciated that any one or a combination of these features and other features may be additionally employed with the metal sponge to establish the desired smear density variation along the longitudinal length of the fuel element. In any one cross-sectional view, the amount of fissionable nuclear fuel material in the fuel composition in the portion of the area occupied by the matrix 1020, 1120, 1220, 1320, 1420 of the metal sponge 1010, 1110, 1210, 1310, 1410 is the value for AreaFuel in the smear density equation and the area defined by the inner surfaces 1040, 1140, 1240, 1340, 1440 of the wall 1050, 1150, 1250, 1350, 1450 formed by the cladding 1060, 1160, 1260, 1360, 1460 is the value for AreaInterior Cross-Section in the smear density equation. As observable from FIGS. 14-18, the smear density varies between that in any of section A′-A′ to section E′-E′ and is illustrative of the smear density varying with position along the longitudinal axis of the fuel element. FIGS. 14-18 also illustrate the modification where the smear density at each longitudinal end of the tubular fuel element is greater than the smear density at a longitudinal center zone of the tubular fuel element. Comparing FIG. 14 (a cross-sectional view at a first longitudinal end) and FIG. 18 (a cross-sectional view at a second longitudinal end) with FIG. 16 (a cross-sectional view at a longitudinal central zone), one can readily ascertain that the area occupied by the matrix 1020, 1420 of fissionable nuclear fuel at the first longitudinal end and second longitudinal end is greater than the area occupied by the matrix 1220 of the fissionable nuclear fuel at a longitudinal center. In addition, FIGS. 14-18 illustrate the additional or alternative modification where the smear density at a first longitudinal end of the fuel element is greater than the smear density at a second longitudinal end of the fuel element, with both ends having a smear density greater than the central zone. Comparing FIG. 14 (a cross-sectional view at a first longitudinal end) with FIG. 18 (a cross-sectional view at a second longitudinal end), one can readily ascertain that the area occupied by the matrix 1020 of fissionable nuclear fuel at the first longitudinal end is greater than the area occupied by the matrix 1420 of fissionable nuclear fuel at the second longitudinal end. In embodiments where the fissionable nuclear fuel is located either or both within a matrix of the metal sponge or as part of the metal sponge and located in the interior volume of the fuel element, modifications of smear density can be achieved by suitable means. For example, the metal sponge can be formed with a specified balance of matrix and void in which the matrix has, nominally, a uniform amount of fissionable nuclear fuel. In such a case, a variation in smear density can be achieved by varying the balance between matrix and void space as different sections of the interior volume along the longitudinal length of the fuel element are filled. To achieve a lower smear density, the amount of matrix is reduced and the amount of void is increased. Similarly, to achieve a higher smear density, the amount of matrix is increased and the amount of void is decreased. One manufacturing method to control the relative amounts of matrix and void in a given area is to use a gas injection technique in which matrix material is passed through a blower or other injector and entrains a gaseous medium. The gaseous medium can be an inert gas or a gas that can function as a gas bond. The amount of gaseous medium entrained per a unit volume of matrix can be controlled and adjusted to produce the longitudinal variation in the ratio of area of fissionable nuclear fuel to total area of interior volume. Other manufacturing techniques that can be additionally or alternatively utilized include powder consolidation with a sintering process to consolidate and form inter-particle metallurgical bonds (which may be optionally accompanied by an external pressure) and one or more sintering processes can remove portions of the original sintered powder due to differences in melting points, or a powder pack process, molds with void forms, drilling out voids within a preformed solid or foam, hydride/dehydriding and/or a 3D printing process. Overall, the smear density can vary with position along the axis between a minimum of 30% and a maximum of 75%. Alternatively, the smear density is 50% to 70%. Higher smear density fuel can be used to offset the long term effects of poison on burn-up distribution. Neutron poisons tend to reduce flux while higher smear density fuels breed more fissile fuel and tend to increase flux. In addition, the reactor coolant first entering the nuclear reactor core tends to be cooler (and more dense) than the coolant exiting the reactor core, leading to higher neutronic absorption and lower flux near the coolant entry point (typically at one end of the fuel element) and lower neutronic absorption and higher flux near the coolant exit point (typically at the opposing end of the fuel element). Also, since the coolant exit point is generally at a higher temperature than the coolant entry point, thermal creep and other temperature effects on the fuel element can be more significant and thus may require an even lower flux to manage those effects. Additionally, control rods generally are ‘pushed down’ from the direction of one longitudinal end of the fuel element towards the central area of the fuel element. These effects can be considered in determining the expected neutron flux and also considered in engineering and selecting the smear density for the fuel at various locations along the length of the fuel element to compensate or influence the expected flux distribution along the length of the fuel element. Based on these principles, the multi-smear fuel element may be provided with various selected smear densities at selected locations along the length of the fuel element to favor breeding toward one end of the fuel element thereby pulling the flux distribution towards or away one or the other longitudinal ends of the fuel element and/or reducing the rate of change or increase the flattening of the flux distribution over the length of the fuel element). Many variables contribute to strain in fuel elements, such as burn-up, displacements per atom (DPA) history, DPA rate history, fission gas release, smear density, time history, cladding material, cladding dimensions, and operational temperatures such as linear heat rate, heat transfer coefficient, and even smear density point differences. In the various embodiments disclosed herein, the fuel can be selectively determined based on the strain limit for the target burn-up in that location as well as in consideration of thermal creep, flux, and/or burn-up. For the embodiment with discrete fuel bodies, whether in fuel pellet form or fuel slug form or other body shape, different diameters (both inner and outer diameters in relevant cases) can be used. For example considering fuel pellets or rods, fuel diameters could be large at a first longitudinal end (i.e., up to the strain limit or the target burn-up in that location), narrower in the middle (region with highest burn-up) of the longitudinally oriented fuel element, and then expand at a second longitudinal end (i.e., up to the strain limit or the target burn-up in that location). Thermal creep is larger at the end of the fuel element that functions as the exit for cooling media due to higher temperatures in this region, so for longer term operation, the actual burn-up and design considerations such as operational creep need to be computationally iterated to find the optimum. Similar considerations can be made for fuel forms which are annular, foamed fuel, or packed pellets with a desired porosity, fuel loading, etc. Also, initial fuel can be loaded with varying porosity (in addition to or alternative to loading full density fuel). Metal fuel, such as in the TWR, expands due to solid and gaseous fission products. The metal fuel tends to fill the volume it is contained within, within the initial couple percent burn-up (early in life). In order to maintain transport of fission gas within the fuel element, the initial smear density, porosity, and burn-up characteristic of the fuel may be considered within design strain limits of the form of the fuel, the aggregated fuel assemblies, and the overall core. FIG. 19 illustrates an example method for decreasing the strain limit and/or increasing fuel burn-up limits by determining the desired distribution of flux and the correlated distribution of smear density. Although the method is presented as a sequence of steps for illustrative purposes, this sequence does not limit the scope of the claimed methods, and those of ordinary skill in the art will be aware of modifications and variations that may be made to the order, timing, etc. of the sequence. The figure is a series of steps or flowcharts depicting implementations. For ease of understanding, the steps or flowcharts are organized such that the initial steps or flowcharts present implementations via an example implementation and thereafter the following steps or flowcharts present other implementations and/or expansions of the initial steps or flowchart(s) as either sub-component operations or additional component operations building on one or more earlier-presented steps or flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a step(s) or flowchart(s) presenting an example implementation and thereafter providing additions to and/or further details in subsequent step(s) or flowchart(s)) 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. Determining the neutronic flux and smear density can be performed using any suitable method. Generally, commercially or government available software can provide the baseline analytics needed to simulate the reactor core and determine the flux. For example, a neutron transport simulator can be used to provide information regarding where the neutrons are within the core and fuel elements; a nuclide transmutation simulator can be used to determine how quickly fissionable atoms are breeding up (from fertile to fissile material) and burning (fission of the fissionable material); a neutron kinetics module can provide the distribution of heat generation along with a thermal hydraulics simulator can provide the heat conduction of the fuel element and temperature profile and determine the maximum strain and maximum allowable burn-up for a particular component such as the fuel element, and a mechanical interaction simulator and fuel performance feedback software can optionally be used to determine other mechanical interactions (such as fuel bowing and interaction with ducts which may affect the neutronic and thermal hydraulics of the system). Various available computer modules may be used including, but not limited to SERPENT available from http://montecarlo.vtt.fi/, MCNp6 available from https://mcnp.lanl.gov/, REBUS available from http://www.ne.anl.gov/codes/rebus/, ERANOS available from https://www.oecd-nea.org/tools/abstract/detail/nea-1683, DRAGON available from http://www.polymtl.ca/nucleaire/en/logiciels/index.php, and the like. As shown in the example method 1500 of FIG. 19, an initial smear density distribution over the length of fuel element is selected 1502. The initial smear density distribution may be any suitable initial starting point which may approximate a uniform smear density or crude or coarse distribution along the length of the fuel element. At 1504, the operational characteristics of the nuclear core are determined based at least in part on the smear density of the fuel element, using known and future methods such as those discussed above with respect to commercially, government, and openly available simulation and analysis tools. Operational characteristics include those variables such as neutronics, thermal hydraulics, fuel performance feedback, mechanical interactions, neutron kinetics and neutron flux distribution, etc. which may include burn-up, displacements per atom (DPA) history, DPA rate history, fission gas release, smear density, time history, cladding material, cladding dimensions, and operational temperatures such as linear heat rate, heat transfer coefficient, and even smear density point differences. Based on the determined operational characteristics, a profile of the fuel strain along the longitudinal axis of the fuel element can be determined 1506. For example, the determination operation 1506 may entail determining strain on selected zones of the fuel element based on the initial smear density distribution and determined nuclear core characteristics. A corrective smear density profile can then be modeled 1508 based on the profile of the fuel strain. In one implementation, the corrective smear density profile is designed to offset and “flatten out” high strain regions. For example, strain may be computed for a homogenous smear density and the smear density may then be selectively increased in areas of low strain and decreased in areas of high strain to flatten out the corrective fuel strain profile. As a result, the corrective smear density profile 1508 may include regions of increased smear density corresponding to areas of low strain in the original strain profile and areas of decreased smear density in regions corresponding to areas of high strain in the original strain profile. In one implementation, the corrective smear density profile is constructed based on a zone-by-zone assessment of the modeled strain. For example, the strain for each zone of the fuel element can be determined (e.g., based on nuclear core characteristics specific to the zone) and this localized strain can then be compared to one or more thresholds, such as a threshold maximum and/or minimum. Based on the comparison of the threshold(s) to the modeled strain, the initial smear density for the associated zone can be modified 1510 (e.g., increased or reduced) to arrive at the corrective smear density for the zone. This corrective smear density forms a portion of the corrective smear density profile of the fuel element. This process can be repeated for each individual zone to generate, by iteration, a smear density profile that offsets the effects of fuel strain and/or uneven flux, resulting in a higher average burn-up rate across the length of the fuel element. As mentioned above, construction of the corrective smear density profile 1508 may entail comparing localized fuel strain to a strain threshold or set of thresholds with respect to each of the different zones. In some implementations, a uniform strain threshold or uniform set of strain thresholds are commonly used for assessment of corrective strain to be applied within each individual zone. In other implementations, the strain threshold(s) are individually set for the different zones based on the structures, materials, and/or other core components proximate to each of the zones of the fuel element. The strain threshold(s) can be any suitable threshold to meet the design characteristics of the system (which can be approximated with a flux distribution, burn-up limit, etc.). When strain in a selected zone of the fuel element violates a strain threshold (either higher than the maximum strain threshold or below the minimum threshold), the initial smear density for the associated zone can be modified, e.g., reduced or increased based on the degree of the threshold violation. For example, smear density may be reduced dramatically in a region with very high strain in excess of a maximum strain threshold; likewise, smear density may be increased slightly in a region with low strain that is just less than a minimum strain threshold. As a result, the corrective smear density profile has the effect of flattening out the flux distribution (e.g., reducing strain across the length of the fuel element). The above process of determining smear density (e.g., the initial smear density), modeling strain, and adjusting smear density (e.g., the corrective smear density) may be repeated multiple times in each zone until the corrective strain profile is determined to meet some suitable stopping criteria. Other considerations within the iteration of determining the corrective smear density within each zone may include any one or more of ensuring keff>1 (ensure criticality of the core by ensuring sufficient fissionable material to maintain criticality), peak core power needed to be retained, etc. Any one or more of these considerations can be further constraints on iterating and potentially optimizing the fuel smear density profile along the length of the fuel element. As noted above in discussing the illustrative embodiments, the smear density can vary with position along the longitudinal axis between a minimum of 30% and a maximum of 100%. Alternatively, the smear density is 50% to 70% or 50% to 75%. In one embodiment, the first longitudinal end of the fuel element has a 50% smear density and a second portion of the fuel element, such as the second longitudinal end, has a different (e.g., greater) smear density (e.g., 55%, 65%, 70%, 75%, etc.). In another embodiment, the second longitudinal end of the fuel element has a 50% smear density and a second portion of the fuel element, such as the first longitudinal end, has a different (e.g., greater) smear density (e.g., 55%, 65%, 70%, 75%, etc.). In a still further embodiment, the central longitudinal portion of the fuel element has a 50% smear density and a second portion of the fuel element, such as both the first longitudinal end and the second longitudinal end, has a different (e.g., greater) smear density (e.g., 55%, 65%, 70%, 75%, etc.). Notably, fuel strain can be influenced by both smear density and cladding thickness. Therefore, some implementations may achieve the effect of flattening-out the fuel strain profile by manipulating cladding thickness rather than (or in addition to) smear density. For example, some implementations may model a corrective cladding thickness profile in lieu of or in addition to the corrective smear density profile described above. Modifications to a fuel element can then be implemented to cladding thickness based on the corrective cladding thickness profile. Such modifications may be made in isolation or combination with modifications to fuel density based on a corrective smear density profile, as described above. FIG. 20 is a graph of smear density (%) as a function of longitudinal location (in arbitrary units) for a first example embodiment of a fuel element. In the FIG. 20 graph, the longitudinal locations are zoned ranging from one to 12. The smear density varies substantially continuously from 70% in the first zone to 52% in zone 6 and 7 and then to 65% in zones 11 and 12. Varying the smear density in this manner can flatten out the expected distribution of flux (typically a substantially Gaussian distribution if the fuel smear density is not varied) that the fuel element will encounter, and allow for higher burn-up of the fuel overall due to the higher burn-up achieved at the end zones 1-3 and 10-12. FIG. 21 is a graph of fractional smear density as a function of longitudinal location (in arbitrary units) for a second example embodiment of a fuel element. In the FIG. 21 graph, the longitudinal locations include twelve zones 1601, 1602, 1603, 1604, 1605, 1606, 1607, 1608, 1609, 1610, 1611, 1612. Although each of the zones in FIG. 21 is shown to be of equal size, it is to be appreciated that the zones may differ in longitudinal length of the fuel element as compared to other zones. The fractional smear density varies in stepwise fashion from 0.70 in the first end zone 1601 to 0.53 in central zones 1606 and 1607 and then to 0.65 in second end zones 1611 and 1612. The graphs in FIGS. 20 and 21 are representative of a longitudinal variation in smear density in a fuel element where the smear density at a first longitudinal end, e.g., zone 1601, is greater than the smear density at a second longitudinal end, e.g., zone 1612. This can help flatten out the expected flux variations along the longitudinal axis due many core factors including, for example, coolant entering at zone 1 and exiting at zone 12. The graphs in FIGS. 20 and 21 are also representative of a longitudinal variation of smear density in a fuel element where the smear density at both the first longitudinal end, e.g., zone 1601, and the second longitudinal end, e.g., zone 1612, is greater than the smear density at a longitudinal center of the elongated fuel element, e.g., zones 1606 and 1607. FIGS. 22A and 22B are graphs of smear density (fraction) as a function of longitudinal location (in arbitrary units or zones) for several other example embodiments of a fuel element. In the FIGS. 22A and 22B graphs, the longitudinal locations include twelve zones and the smear density varies in stepwise fashion. The baseline (solid line 1700) represents the longitudinal variation in smear density from FIG. 21. Two variations of this baseline are illustrated—a first variation (V_1 in FIG. 21A indicated by the long dashed line 1710) and a second variation (V_2 in FIG. 21B indicated by the short dashed line 1750). The first variation (V_1) 1710 overlays the baseline 1700 at longitudinal locations in zones 1711, 1712, 1713, 1714, 1715, 1716, 1717, but in zones 1718, 1719, 1720, 1721, 1722 the smear density of the first variation (V_1) 1710 is lower than that for the same zone in the baseline 1700 and is less than the corresponding smear values in the first end in zones 1711, 1712, 1713, 1714, 1715 of V_1 variation 1710. In addition, in the first variation (V_1) 1710, the minimum in smear density occurs at a longitudinal center of the elongated fuel element that includes zones 1716, 1717, 1718. In general, the smear densities in zones 1719, 1720, 1721 in the first variation (V_1) 1710 are the same as in the corresponding one lower zone in the baseline 1700 (Smear DensityV_1, zone n=Smear Densitybaseline, zone (n-1)) by which, in the first variation (V_1) 1710, the smear density distribution in the zones 1719, 1720, 1721, 1722 of the fuel element has been modified to match a lower value in the second longitudinal end relative to the baseline 1700 (the smear density of a particular location in the second end matches the smear density of a corresponding segment of the baseline variation more centrally located relative the particular segment in the second variation) and, as a result, tends to further distribute the neutron flux toward the second end, and tends to shift the neutron flux towards the first end overall. The second variation (V_2) 1750 of FIG. 21B overlays the baseline 1700 at some longitudinal locations, for example, at zones 1751, 1753, 1754, 1756 and 1757, but in the other zones the smear density of the second variation (V_2) 1750 is lower than that for the same zone in the baseline 1700. In addition, in the second variation (V_2) 1750, the minimum in smear density occurs at a location off-centered from the longitudinal center of the elongated fuel element toward the second longitudinal end. In the second variation (V_2) 1750 example, the minimum smear density is 52% (or fractional smear density of 0.52) in zone 1758. In general, the variations in smear densities in the second variation (V_2) 1750 compared to the baseline 1700 are such that the smear densities in the first longitudinal zones (1751, 1752, 1753) and in the second longitudinal zones (1759, 1760, 1761, 1762) include smear densities that are the same as or lower than in the corresponding zones in the baseline 1700. Further, in the second end longitudinal zones, the smear densities in zones 1760, 1761, 1762 are the same as in the corresponding two lower zones in the baseline (Smear DensityV_2, zone n=Smear Densitybaseline, zone (n-2)) by which the smear density distribution in the second end zones (1759-1762) of the fuel element has been shifted to a lower value in the second longitudinal end relative to the same segment in the baseline 1700 and, as a result, tends to further distribute the neutron flux toward the second end, and tends to shift the flux towards the first end overall. A smear density fraction of 50% (or fractional smear density of 0.50) is shown in FIGS. 22A and 22B as a reference. The improvements and benefits of longitudinal variation in smear density in a fuel element as compared to fuel elements having a uniform smear density can be demonstrated using the software simulation and analysis tools described above. In a first simulation (shown graphically in FIG. 23), a fuel element had a uniform smear density fraction as a function of longitudinal location. The fuel element in this first simulation included a substantially uniform smear density of 50% (smear density fraction of 0.5) with a burn-up limit of approximately 30%. In the second simulation (shown graphically in FIG. 24), a fuel element had a smear density fraction that varied as a function of longitudinal location. In FIG. 24, the smear density fraction 1930 is presented as a pseudo curve fit of the segmented variation of smear density along the longitudinal length of the fuel element and varies from approximately 70% in longitudinal location 1 to approximately 50% around longitudinal location 4. The material, geometries, etc. and other characteristics of the fuel elements in the first simulation (shown in FIG. 23) and in the second simulation (shown in FIG. 24) were substantially the same and were operated at substantially the same operating conditions and boundary conditions and varied only with respect to the longitudinal variation of smear density fraction. Accordingly, the change in burn-up limit and burn-up for each of the fuel elements in the simulations, as well as the difference in fuel utilization efficiency, can be attributed to the difference in longitudinal variation of smear density fraction between the two fuel elements in the two simulations. FIGS. 23 and 24 illustrate the burn-up limit, burn-up and smear density fraction as a function of longitudinal position (in arbitrary units or zones) for the two simulations (the first simulation in FIG. 23 and the second simulation in FIG. 24). FIG. 23 illustrates an equilibrium cycle distribution plot of fuel nuclides relating burn-up (in % FIMA (fissions per initial metal atom)) to longitudinal location for conventional nuclear reactors in which the smear density is nominally constant with position along the longitudinal axis. Profile 1830 illustrates the smear density fraction as a function of longitudinal position. Profile 1800 is the distribution of burn-up over the longitudinal length of the fuel element and has a shape substantially of that of an inverted cosine. The profile 1800 is a function of the physics of the system with a constant smear density and time integrated flux on the fuel, which produces a distribution with a peak toward the longitudinal center. Reference profile 1810 illustrates the maximum burn-up limit for the system. The difference between the burn-up in profile 1800 and the burn-up limit in reference profile 1810 (see shaded area 1820a and 1820b between profile 1800 and profile 1810) represents the amount of fuel available, but not used and not achieving its potential burn-up limit at the locations proximate the longitudinal ends due to the fuel material in the longitudinal central portion of the fuel element reaching its burn-up limit and thus, requiring the need to retire the fuel element despite available fuel in the longitudinal end portions. This ‘early’ retirement of a fuel element in FIG. 22 relates to the lower efficiency of the uniform smear density (as represented by the larger the area between the burn-up and the burn-up limit, indicating the system is less efficient with regards to fuel utilization). FIG. 24 illustrates an equilibrium cycle distribution plot of fuel nuclides relating burn-up (in % FIMA (fissions per initial metal atom)) to longitudinal location for a nuclear reactor in which the smear density in the fuel elements varies with position along the longitudinal axis. Profile 1900 displays a cosine-like distribution of actual burn-up over the longitudinal length of the fuel element with a maximum of 25% FIMA at about longitudinal position 5.5. This distribution of burn-up over longitudinal length reflects the varied smear density over the same longitudinal length with zones 1-3 having a higher smear density than the smear density of zones 9-12, and zones 9-12 having a higher smear density than the central zones 4-8. Compared to the reference profile 1800 of the conventional reactor with uniform smear density, the profile 1900 has a reduced maximum burn-up (25% FIMA compared to 30% FIMA) and a reduced global maximum overall. Also compared to the reference profile 1800 of FIG. 23, the profile 1900 of FIG. 29 is shifted toward the first longitudinal end (1940 in FIG. 24 and is located at longitudinal location 1; corresponding first longitudinal end in FIG. 23 is 1840 and located at longitudinal location 1), and the burn-up at the first longitudinal end is increased and the burn-up at the second longitudinal end (1950 in FIG. 24 and located at longitudinal location 12; corresponding second longitudinal end in FIG. 23 is 1850 and is located at longitudinal location 14) is decreased and as a result, the overall burn-up of the fuel element is increased and better meets the overall potential burn-up through the entire fuel element. Reference profile 1810 of FIG. 23 illustrates the potential or maximum allowable burn-up limit for the fuel element. The difference between the burn-up in profile 1800 and the burn-up limit in reference profile 1810 is represented by areas 1820a and 1820b and represents the amount of fuel available, but not being used, and relates to the efficiency of the conventional nuclear reactor (the smaller the area between the burn-up and the burn-up limit, the more efficient is the system). In FIG. 24, the difference between the burn-up in profile 1900 and the burn-up limit in reference profile 1910 is represented by area 1920. Area 1920 is reduced (relative to areas 1820a and 1820b illustrated in FIG. 23), particularly in the portion of the fuel element from the first longitudinal end to past the longitudinal center of the elongated fuel element. A reduced difference between the burn-up in profile 1900 and the burn-up limit in reference profile 1910 indicates that the nuclear reactor in which the smear density in the fuel elements varies with position along the longitudinal axis is more efficient in its fuel utilization than the conventional nuclear reactor with uniform smear density. It is to be appreciated that the difference between the actual burn-up and the burn-up limit may be designed through smear density to substantially match in all zones, be optimized globally over all zones, and/or be optimized in a subset of zones, e.g., the first longitudinal end and central zones while leaving some unmet burn-up limit at the second longitudinal end due to other design constraints or desires, such as safety margin in the higher temperature second longitudinal end due to proximity to coolant exit, etc. It is to be appreciated that selecting the smear density to match certain desired and selected design characteristics of the fuel element and/or nuclear core while meeting other design constraints can be utilized using the general methods described herein including optimizing for peak power, peak temperature, criticality, burn-up limit, strain, cladding stability, etc. FIG. 24 also illustrates that, by adding more fuel (e.g., increasing the smear density) above and below the expected peak burn-up location relative to the central location in conventional systems (for example, at or towards the respective longitudinal ends), the peak burn-up is reduced and the burn-up distribution is flattened and shifted. The process of longitudinally varying smear density as disclosed herein can be used to longitudinally spread out the burn-up distribution, reduce the strain on the fuel element and fuel assembly, and/or increase the overall burn-up of a fuel element allowing for more efficient use of fuel material within a nuclear reactor. Further, the variation between profile 1900 and reference profile 1910 demonstrates that longitudinal variation in smear density meaningfully increases fuel utilization efficiency, preferably increases fuel utilization efficiency by 10% or greater. As an additional benefit, longitudinal variation of smear density contributes to reactor design benefits. For example, the overall longitudinal length in the fuel element represented by FIG. 24 is less than the overall longitudinal length in the fuel element represented by FIG. 23. However, although the fuel element represented by FIG. 24 is shorter than the fuel element represented by FIG. 23, the fuel utilization efficiency is greater in the fuel element represented by FIG. 24 than in the fuel element represented by FIG. 23. In one example, the height of the fuel column decreased from 2.4 meters to 2.0 meters while at the same time the fuel mass decreased by 2%. Thus, FIGS. 23 and 24 illustrate an attendant benefit to reactor design from incorporating longitudinal smear density variation—shorter fuel elements which are more efficient with respects to fuel utilization and that can contribute to more compact and efficient reactor designs without reducing power output of the reactor. Also, a decreased fuel column height (in addition to the mechanical advantage of smaller reactor) can provide reduce pressure drop in associated coolant over less length. FIG. 25 illustrates the strain limit, strain and smear density fraction as a function of longitudinal position (in normalized units or zones) in a fuel element which varies smear density with longitudinal length. Profile 2000 illustrates the smear density fraction as a function of longitudinal position and is illustrated as a pseudo-curve fit of the segmented smear density in each zone of the normalized longitudinal location. Profile 2000 of the smear density fraction varies from between about 45% and 50% in a central portion (at longitudinal location 0.5) to about 70% at the end portions (at longitudinal location 0 and 1.0) and has the shape that approximates an inverted Gaussian distribution. Profile 2010 illustrates a distribution of strain over the longitudinal length of the fuel element and has a shape substantially of that of an inverted cosine. The profile 2010 is a function of the physics of the system and the selectively varied smear density and time integrated flux on the fuel, which produces a distribution with a peak strain toward the longitudinal center. Reference profile 2020 illustrates the maximum strain limit for the system. The difference between the strain in profile 2010 and the strain limit in reference profile 2020 (see shaded area 2030a and 2030b between profile 2010 and profile 2020) represents the amount of strain safety margin at the locations proximate the edges due to the fuel material in the central portion of the fuel element reaching its strain limit. Although the figure shows the example as a single longitudinal location of the fuel column hitting the maximum strain, one or more sections of the fuel column may hit the strain limit and may have other curves of strain or even variable strain thresholds along the length of the fuel element. FIG. 25 illustrates a maximum strain limit for the system (reference profile 2020) with a constant value as a function of longitudinal location. However, the constant strain limit shown in the figures is only an example and other strain limits, including variable strain limits, are possible to allow higher strain in some locations and lower strains at other locations. One of skill in the art can recognize that the multi-smear fuel disclosed herein can be tailored to accommodate a variable strain limit and allow more strain at the associated location of the higher strain limits. FIG. 26 illustrates an example method 2100 of manufacturing a fuel element with varied smear density along the longitudinal length of the fuel element. The selected smear density at each zone of the fuel element may be determined 2110 in any suitable manner such as that described with reference to FIGS. 19, 22 and 23 above. The cladding may be provided 2112 and defines an interior volume. As noted above, the cladding may include one or more liners and may comprise any suitable material for containing a fuel composition and may include tubular shaped walls. Although the following method is described with reference to 12 zones along the longitudinal length of the fuel element and three general sections (first end, central section, and second end), it is to be appreciated that any number of zones and/or sections may be used as appropriate to achieve the desired manufacturing and/or operational fidelity characteristics desired. Moreover, each zone and/or section may be of similar length and/or size and shape as the other zones and/or sections along the longitudinal length of the fuel element, and it is to be appreciated that the zones and/or sections may be of different lengths relative to other zones and/or sections within the same fuel element as appropriate for the determined and selected distribution profile of the smear density. A first fissionable composition containing a fissionable nuclear fuel material may be disposed 2114 in a first section of the interior volume of the cladding proximate a first longitudinal end of the nuclear fuel element. The first fissionable composition may be in thermal transfer contact with an interior surface of the cladding. The first fissionable composition may have a first smear density. Within the first section, the smear density can vary continuously, vary step-wise, be non-variant or be a combination thereof as a function of longitudinal position along the length of the fuel element to achieve the determined smear density for that section. Disposing the fissionable composition in the cladding may be accomplished with any of the techniques described above which may include, without limitation, inserting a pellet (or other body shape of the fissionable composition, for example powder, particulate, slug, etc.) and may include additional steps of compacting or consolidating the fissionable composition such as sintering, vibration, blowing, tamping, etc. A second fissionable composition, also containing a fissionable nuclear fuel material, which may be the same as or different from the first fissionable nuclear fuel material, may be disposed 2116 in a second section of the interior volume of the cladding proximate the central section. The second section of the interior volume of the cladding may be adjacent and/or spaced from the first section. The second fissionable composition may be in thermal transfer contact with the cladding and may have a second smear density. The second smear density may be different from the first smear density. To flatten out the burn-up profile of the fuel element, the second smear density may be less than the first smear density, and in some cases, may compensate for difference in coolant temperature (e.g., if the first longitudinal end is proximate a coolant entry point of the fuel assembly and the second longitudinal end is proximate the coolant exit point of the fuel assembly). Within the second section, the smear density can vary continuously, vary step-wise, be non-variant or be a combination thereof as a function of longitudinal position along the length of the fuel element to achieve the determined smear density for that section. A third fissionable composition, also containing a fissionable nuclear fuel material, which may be the same as or different from the first and/or second fissionable nuclear fuel materials, may be disposed 2118 in a third section of the interior volume of the cladding proximate a second longitudinal end of the nuclear fuel element. The third section of the interior volume of the cladding may be adjacent and/or spaced from the second central section. The third fissionable composition may be in thermal transfer contact with the cladding and may have a third smear density. The third smear density may be different from the first and/or second smear density. To flatten out the burn-up profile of the fuel element, the third smear density may be greater than the second smear density. In some cases, it may be appropriate for the third smear density to be less than the first smear density. Within the third section, the smear density can vary continuously, vary step-wise, be non-variant or be a combination thereof as a function of longitudinal position along the length of the fuel element to achieve the determined smear density for that section. The first section proximate the first longitudinal end may be divided into any number of zones as appropriate for the selected fidelity of smear density distribution determined 2110. In one example, the first section may contain two zones. In this example, the first zone of the first section is proximate the first longitudinal end of the fuel element cladding may contain a fourth fissionable composition having a fourth smear density. The fourth smear density may be different from or the same as the first smear density of the first section. A second zone of the first section may be adjacent to the first zone and may have disposed therein a fifth fissionable composition having a fifth smear density. The fifth smear density may be the same as or different from the first smear and/or fourth smear densities, and in some cases, the fifth smear density may be less than the fourth smear density. The volume average of the fourth and fifth smear densities may be equal to the first smear density of the first section overall. Additionally and/or alternatively the third section, proximate the second end, may be divided into any number of zones as appropriate for the selected fidelity of smear density distribution determined 2110. In one example, the third section may contain two zones—and as appropriate, the number of zones in the third section may be the same as, greater than or less than the number of zones in the first and/or second sections. In this example the fourth zone of the third section (if the first section has first and second zone and the second section has one zone) is proximate the second section of the fuel element cladding and may contain a sixth fissionable composition having a sixth smear density. The sixth smear density may be different from or the same as the third smear density of the third section. A fifth zone of the third section may be adjacent to the fourth zone and may have disposed therein a seventh fissionable composition having a seventh smear density. The seventh smear density may be the same as or different from the third smear and/or sixth smear densities, and in some cases, the seventh smear density may be greater than the sixth smear density. The volume average of the sixth and seventh smear densities may be equal to the third smear density of the third section overall. To provide higher fidelity in distribution or variation of smear density along the length of the fuel element, additional zones and/or sections may be implemented. In this example, a fuel element with three sections and twelve zones is described. The smear densities of the first, second and third sections may be as described above with respect to FIG. 26. The first section proximate the first longitudinal end may be divided into any number of zones as appropriate for the selected fidelity of smear density distribution determined 2110. In this example, the first section may contain five zones. In this example, the first zone of the first section is proximate the first longitudinal end of the fuel element cladding may contain a fourth fissionable composition having a fourth smear density. The fourth smear density may be different from or the same as the first smear density of the first section. A second zone of the first section may be adjacent to the first zone and may have disposed therein a fifth fissionable composition having a fifth smear density. The fifth smear density may be the same as or different from the first smear and/or fourth smear densities, and in some cases, the fifth smear density may be less than the fourth smear density. A third zone of the first section may be adjacent the second zone and may have disposed therein a sixth fissionable composition having a sixth smear density. The sixth smear density may be the same as or different from the first, fourth, and/or fifth smear densities, and in some cases, the sixth smear density may be less than the fifth smear density. A fourth zone of the first section may be adjacent the third zone and may have disposed therein a seventh fissionable composition having a seventh smear density. The seventh smear density may be the same as or different from the first, fourth, fifth, and/or sixth smear densities, and in some cases, the seventh smear density may be less than the sixth smear density. A fifth zone of the first section may be adjacent the fourth zone and may have disposed therein an eighth fissionable composition having an eighth smear density. The eighth smear density may be the same as or different from the first, fourth, fifth, sixth, and/or seventh smear densities, and in some cases, the eighth smear density may be less than the seventh smear density. The volume average of the fourth, fifth, sixth, seventh and eighth smear densities may be equal to the first smear density of the first section overall. The second central section may be divided into any number of zones as appropriate for the selected fidelity of smear density distribution determined 2110. In this example, the second section may contain two zones. In this example the sixth zone of the fuel element in the second section is proximate the first section of the fuel element and may contain a ninth fissionable composition having a ninth smear density. The ninth smear density may be different from or the same as the second smear density of the second section. A seventh zone of the second section may be adjacent to the sixth zone and may have disposed therein a tenth fissionable composition having a tenth smear density. The tenth smear density may be the same as or different from the second and/or ninth smear densities, and in some cases, the tenth smear density may be the same as the ninth smear density. The volume average of the ninth and tenth smear densities may be equal to the second smear density of the second section overall. Additionally and/or alternatively the third section, proximate the second longitudinal end, may be divided into any number of zones as appropriate for the selected fidelity of smear density distribution determined 2110. In one example, the third section may contain five zones—and as appropriate, the number of zones in the third section may be the same as, greater than or less than the number of zones in the first and/or second sections. In this example, the eighth zone of the fuel element in the third section is proximate the second section of the fuel element and may contain an eleventh fissionable composition having an eleventh smear density. The eleventh smear density may be different from or the same as the third smear density of the third section, and in some cases may be greater than the second smear density of the second central section and/or the tenth smear density of the eighth zone. A ninth zone of the third section may be adjacent to the eighth zone and may have disposed therein a twelfth fissionable composition having a twelfth smear density. The twelfth smear density may be the same as or different from the third and/or eleventh smear densities, and in some cases, the twelfth smear density may be greater than the eleventh smear density. A tenth zone of the third section and adjacent ninth zone, may have disposed therein a thirteenth fissionable composition having a thirteenth smear density. The thirteenth smear density may be the same as or different from the third, eleventh, and/or twelfth smear densities, and in some cases, the thirteenth smear density may be greater than the twelfth smear density. An eleventh zone of the third section and adjacent the tenth zone, may have disposed therein a fourteenth fissionable composition having a fourteenth smear density. The fourteenth smear density may be the same as or different from the third, eleventh, twelfth, and/or thirteenth smear densities, and in some cases, the fourteenth smear density may be greater than the thirteenth smear density. A twelfth zone of the third section may be adjacent the eleventh zone and proximate the second longitudinal end of the fuel element and may have disposed therein a fifteenth fissionable composition having a fifteenth smear density. The fifteenth smear density may be the same as or different from the third, eleventh, twelfth, thirteenth, and/or fourteenth smear densities, and in some cases, the fifteenth smear density may be greater than the fourteenth smear density. The volume average of the eleventh, twelfth, thirteenth, fourteenth, and fifteenth smear densities may be equal to the third smear density of the third section overall. In this example, within any one or more of the first to twelfth zones, the smear density can vary continuously, vary step-wise, be non-variant or be a combination thereof as a function of longitudinal position along the length of the fuel element to achieve the determined smear density for that zone. FIG. 27 illustrates an example arrangement 2700 including a fuel assembly duct 2706 holding a number of individual fuel elements (e.g., a fuel element 2702). Each of the fuel elements includes a cladding (e.g., cladding 2704) defining a tubular interior volume that stores a fissionable composition (e.g., fissionable composition 2710) in thermal transfer contact with an interior surface of the cladding. In one implementation, each fuel element is defined by a smear density profile that selectively varies with position along a longitudinal axis of the fuel element. As discussed with respect to other implementations above, the selectively variable smear density may be designed to offset points of natural strain in the fuel element. For example, regions of locally increased smear density may be deliberately positioned to coincide with regions of the fuel element 2702 that would otherwise exhibit locally decreased strain. Likewise, regions of locally decreased smear density may be deliberately positioned to coincide with regions of the fuel element 2702 that would otherwise exhibit locally increased strain. In still other implementations, the fuel element 2702 has a smear density profile along the longitudinal axis designed to provide points of high strain at points that generally mirror points of high strain naturally occurring along the fuel assembly duct 2706. For example, the fuel assembly duct 2706 may have natural points of strain at various heights along the longitudinal axis, such as at an example high strain location 2712. For example, a pressure differential between interior and exterior walls of the fuel assembly duct 2706 may create a driving force that causes the wall of the fuel assembly duct 2706 to be in tension and swell away from the internal fuel elements proximal to one or more regions along the longitudinal axis (e.g., near the example high strain location 2712). This localized swelling can cause the walls of the fuel assembly duct 2706 to bow outward, widening a gap between the fuel elements and the walls of the fuel assembly duct. This effect permits coolant to bypass the center fuel elements rather than flowing between them, causing a thermal hydraulic penalty. In one implementation, the fuel element 2702 has one or more points of locally enhanced strain at a height along its longitudinal axis that corresponds to a high strain point of the fuel assembly duct 2706. In general, points of increased strain in the fuel assembly duct 2706 can be identified by analyzing flux and pressure distributions. For example, high strain points may correlate with regions where the multiple of flux and pressure is highest. Once high strain regions of the fuel assembly duct 2706 are identified, corresponding high strain regions can be created in the fuel element 2702 by selectively varying smear density and/or cladding thickness along the longitudinal axis of the duct assembly 2706. Although the present invention has been described in connection with embodiments thereof, it will be appreciated by those skilled in the art that additions, deletions, modifications, and substitutions not specifically described may be made without departure from the spirit and scope of the invention as defined in the appended claims. 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,” “configured by,” “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. Those skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims.
	description	The present disclosure relates generally to ion implantation and, more particularly, to techniques for plasma injection for space charge neutralization of an ion beam. Ion implanters are widely used in semiconductor manufacturing to selectively alter the conductivity of materials. In a typical ion implanter, ions generated from an ion source are transported downstream through a series of beamline components which may include one or more analyzer and/or collimator magnets and a plurality of electrodes. The analyzer magnets may be used to select desired ion species and filter out contaminant species or ions having undesirable energies. The collimator magnets may be used to manipulate the shape of the ion beam or otherwise adjust the quality of the ion beam before it reaches a target wafer. Suitably shaped electrodes can be used to modify the energy and the shape of the ion beam. After the ion beam has been transported through the series of beamline components, it may be directed into an end station to perform ion implantation. FIG. 1 depicts a conventional ion implanter system 100. As is typical for most ion implanters, the system 100 is housed in a high-vacuum environment. The ion implanter system 100 may comprise an ion source 102 and a series of beamline components through which an ion beam 10 passes. The series of beamline components may include, for example, an extraction manipulator 104, a filter magnet 106, an acceleration or deceleration column 108, an analyzer magnet 110, a mass slit 112, a scanner 114, and a collimator magnet 116. Much like a series of optical lenses that manipulate a light beam, the ion implanter components can filter and focus the ion beam 10 before steering it towards a target wafer 118. As the semiconductor industry keeps reducing feature sizes of micro-electronic devices, ion beams with lower energies are desirable in order to achieve shallow dopant profiles for forming shallow p-n junctions. Meanwhile, it is also desirable to maintain a relatively high beam current in order to achieve a reasonable production throughput. Such low-energy, high-current ion beams may be difficult to transport within typical ion implanters due to space charge blow-up. To prevent “blow-up” of a positive ion beam, negatively charged particles, such as electrons or negative ions, may be introduced for charge neutralization. One way of sustaining space charge neutralization is through magnetic confinement of negatively charged particles. However, existing magnetic confinement approaches tend to introduce extra magnetic field components that cause ion beam distortion. Moreover, in order to improve low-energy beam transportation caused by space charge limitations, a high-energy ion beam may be decelerated to a desired energy level before reaching a target (e.g., a wafer). In such cases, some ions may go through “charge exchange” with surrounding neutral particles, thus losing their charge prior to deceleration and generating neutral particles having high energy. Neutral particles having high energy fail to be decelerated and may impact the target at a higher energy level than desired, thus negatively impacting implantation results. Low-energy ion beams may be difficult to transport through the beamline to the target due to mutual repulsion between ions having the same charge. High-current ion beams typically include a high concentration of charged ions that tend to diverge due to mutual repulsion. To maintain low-energy, high-current ion beam quality, a plasma may be injected into the ion beam for the purpose of charge neutralization. High-energy ion implantation beams typically propagate through a weak plasma that is a byproduct of beam interactions with residual or background gas. This plasma tends to neutralize the space charge caused by the ion beam, thereby largely eliminating transverse electric fields that would otherwise disperse the ion beam. However, for a low-energy ion beam, the probability of ionizing collisions with background gas is lower compared to a high-energy ion beam. In addition, low-energy ion beam blow-up may occur at much lower transverse electric field strength. In view of the foregoing, it may be understood that there are significant problems and shortcomings associated with current techniques for transporting low-energy ion beams. Techniques for plasma injection for space charge neutralization of an ion beam are disclosed. In one particular exemplary embodiment, the techniques may be realized as a plasma injection system for space charge neutralization of an ion beam. The plasma injection system may comprise a first array of magnets and a second array of magnets positioned along at least a portion of an ion beam path, the first array being on a first side of the ion beam path and the second array being on a second side of the ion beam path, the first side opposing the second side. At least two adjacent magnets in the first array of magnets may have opposite polarity. The plasma injection system may also comprise a plasma source configured to generate a plasma in a region associated with a portion of the ion beam path by colliding at least some electrons with a gas. In accordance with other aspects of this particular exemplary embodiment, at least two adjacent magnets in the second array of magnets may have opposite polarity. In accordance with further aspects of this particular exemplary embodiment, the gas may comprise an inert gas (e.g., argon, nitrogen, xenon, helium, etc.) and/or an electronegative gas. In accordance with additional aspects of this particular exemplary embodiment, the first array of magnets and the second array of magnets may collectively produce cusp magnetic fields to inject the plasma in or near the ion beam path. In accordance with yet another aspect of this particular exemplary embodiment, the plasma source may be embedded in a pole piece along at least a portion of the ion beam path. In accordance with still another aspect of this particular exemplary embodiment, the beamguide may comprise alternating sloped concave portions and sloped convex portions. In accordance with further aspects of this particular exemplary embodiment, the first array of magnets or the second array of magnets may be located at the convex portion of the beamguide. In accordance with additional aspects of this particular exemplary embodiment, the beamguide may further comprise a plurality of apertures located at the concave portion of the beamguide. In accordance with another aspect of this particular exemplary embodiment, at least one of the first array of magnets or the second array of magnets may be a permanent magnet. In accordance with yet another aspect of this particular exemplary embodiment, at least one magnet in the first array or the at least one magnet in the second array may be configured to direct the flow of the plasma in a “cross-B” (x B) loop. In accordance with still another aspect of this particular exemplary embodiment, the “cross-B” (x B) loop may be formed by at least one of diamagnetic drift, E cross B drift (E x B), and curvature drift (R x B). In accordance with further aspects of this particular exemplary embodiment, the first array of magnets may be configured interdigitally. In accordance with additional aspects of this particular exemplary embodiment, the second array of magnets may be configured interdigitally. In accordance with another aspect of this particular exemplary embodiment, the first array of magnets and second array of magnets may be positioned in an analyzer magnet. In accordance with yet another aspect of this particular exemplary embodiment, the first array of magnets and the second array of magnets may be positioned in a collimator magnet. In accordance with still another aspect of this particular exemplary embodiment, the plasma source may comprise a microwave source and a coil. In another particular exemplary embodiment, the techniques may be realized as another plasma injection system for space charge neutralization of an ion beam. The plasma injection system may comprise a first array of magnets and a second array of magnets positioned along at least a portion of an ion beam path, the first array being on a first side of the ion beam path and the second array being on a second side of the ion beam path, the first side opposing the second side. At least two adjacent magnets in the first array of magnets may have opposite polarity. The plasma injection system may also comprise an RF power source coupled to at least one of the first array of magnets and at least one of the second array of magnets, and a plasma source configured to generate a plasma in a region associated with a portion of the ion beam path by colliding at least some electrons with a gas. In accordance with other aspects of this particular exemplary embodiment, the first array of magnets and the second array of magnets may collectively produce cusp magnetic fields to inject the plasma in or near the ion beam path. In accordance with further aspects of this particular exemplary embodiment, the plasma injection system may further comprise a shield member configured to shield at least one of the first array of magnets and at least one of the second array of magnets from the ion beam path. In accordance with additional aspects of this particular exemplary embodiment, the shield member may be made from at least one of metal, silicone, elastomer material, and dielectric material. In accordance with another aspect of this particular exemplary embodiment, the first array of magnets or the second array of magnets may be configured interdigitally. In yet another particular exemplary embodiment, the techniques may be realized as a plasma confinement system for space charge neutralization of an ion beam. The plasma confinement system may comprise a high transparency grid configured to connect to a ground potential and a plate configured to connect to a negative voltage to repel electrons from the ion beam path. In accordance with a further aspect of this particular exemplary embodiment, the plasma confinement system may further comprise an ion scraper configured to protect the high transparency grid from direct impinging ions from the ion beam path. The present disclosure will now be described in more detail with reference to exemplary embodiments thereof as shown in the accompanying drawings. While the present disclosure is described below with reference to exemplary embodiments, it should be understood that the present disclosure is not limited thereto. Those of ordinary skill in the art having access to the teachings herein will recognize additional implementations, modifications, and embodiments, as well as other fields of use, which are within the scope of the present disclosure as described herein, and with respect to which the present disclosure may be of significant utility. Embodiments of the present disclosure overcome inadequacies and shortcomings in existing plasma generation and diffusion techniques used in ion implanters by providing improved techniques for plasma diffusion along an ion beam path in an ion implanter. Instead of diffusing plasma electrons across magnetic dipole field lines, which makes introduction of plasma electrons from an auxiliary source challenging, diffusion of plasma electrons along local magnetic field lines is contemplated. It should be noted that, although the description hereinafter refers to plasma as the subject of cusp coupling into an ion beam path, embodiments of the present disclosure are not limited to plasmas but may be adapted to any cusp-coupled charged particles, including negative and positive ions. Hereinafter, a magnet, whether a permanent magnet, an electromagnet, or otherwise, may sometimes be referred to as a “cusp magnet” if it is used for a magnetic cusp coupling purpose. Moreover, a cusp line may be referred to as a path parallel to and situated midway between two consecutive cusp magnets. Furthermore, plasma in the present disclosure may include a combination of positive ions and electrons. Referring to FIG. 2, there is shown an exemplary sectional view of a cusp coupled plasma injection system 200 in accordance with an embodiment of the present disclosure. The system 200 may comprise a plasma source 201 that may be embedded within a pole piece of bending magnet 202. The plasma source 201 may be located at any magnetized and/or non-magnetized portion of a path of the ion beam 10 between the ion source 102 and the target wafer 118, as shown in FIG. 1. Pole piece of bending magnet 202 may be a magnetic dipole made of metal (e.g., steel, etc.) having a top and a bottom portion. A magnetic coil may be wound around the periphery of the pole piece of the bending magnet 202 to generate a magnetic dipole field. For example, the plasma source 201 may be embedded within the analyzer magnet 110 and/or the collimator magnet 116, as shown in FIG. 1. Plasma source 201 may be, for example, an electron cyclotron resonance (ECR) plasma source and/or a radio frequency (RF) plasma source. An electron cyclotron resonance (ECR) plasma source, as shown in FIG. 2, may include a magnetic coil 203 located around the outer periphery of a plasma chamber 204 for generating a local magnetic field inside the plasma chamber 204. Moreover, magnetic coil 203 may be permanent magnets in order to generate a magnetic field inside the plasma chamber 204. A dielectric window 205 may be placed at a top end of the plasma chamber 204 for allowing microwave energy to be transmitted therethrough. In addition, the dielectric window 205 may provide a vacuum seal to the plasma chamber 204. The plasma chamber 204 may be coupled to a microwave waveguide 206, through the dielectric window 205, for feeding microwave energy from a microwave source 207. The microwave waveguide 206 may be a metal tube having a cross section suitable to allow microwave energy to travel from the microwave source 207 to the plasma chamber 204. For example, microwave waveguide 206 may be a rectangular aluminum tube. When microwave energy is introduced through the dielectric window 205, electrons are accelerated via the electron cyclotron resonance within the interior space of the plasma chamber 204. Gas molecules and/or atomic gas within the plasma chamber 204 may be ionized to produce a plasma 212. An auxiliary gas container (not shown) may be coupled to the plasma chamber 204 for introducing gas which may be ionized to generate the plasma 212. The auxiliary gas container may contain helium, neon, nitrogen, argon, krypton, xenon, radon and/or other electronegative or electropositive gases which may be suitable for plasma generation. The plasma chamber 204 has an aperture 208 extending through an inner wall of the pole piece of the bending magnet 202 and a beamguide 214. Along this inner wall are located plasma extracting magnet arrays 209 for generating a multi-cusped magnetic field. The plasma extracting magnet arrays 209 may be arranged into two arrays, with one array 210 above the path of the ion beam 10 and the other array 211 below the path of the ion beam 10. According to some embodiments, the two plasma extracting magnet arrays 209 may be substantially in parallel with one another. The plasma extracting magnet arrays 209 may be permanent magnets with their magnetic orientation aligned approximately with the propagation direction of the ion beam 10. Also, the plasma extracting magnet arrays 209 may be a coil coupled to a power source in order to generate a multi-cusped magnetic field. Within each array, polarities of individual magnets may be alternated such that multi-cusp magnetic fields may be created in spaces between the magnets (i.e., in or near the beam path of the ion beam 10). As seen in FIG. 2, plasma extracting magnet arrays 209 may be magnetized longitudinally along the propagation direction of the ion beam 10 and may be arranged such that adjacent magnets have opposite polarity poles. For example, an up arrow in FIG. 2 on the plasma extracting magnet arrays 209 may indicate a north pole while a down arrow may indicate a south pole or vice versa. Adjacent magnets of the plasma extracting magnet arrays 209 may have opposite polarities and generate a multi-cusped magnetic field in order to aid the flow of plasma 212 from the plasma chamber 204 along the direction of the ion beam 10. The plasma 212 generated in the plasma chamber 204 may be extracted into the ion beam 10 via aperture 208. FIG. 2 illustrates an initial plasma injection region 213, wherein the plasma tends to congregate near the aperture 208 because the magnetic dipole field is perpendicular to the propagation direction of the ion beam 10. In order to uniformly distribute the plasma along the path of the ion beam 10, plasma extracting magnet arrays 209 produce a multi-cusped magnetic field. The multi-cusped magnetic field, in combination with a plasma density gradient and/or an electric field at the “plasma sheath”, causes a “cross-B” (x B) drift motion and thus may enable uniform distribution of the initial plasma injection region 213 along the path of the ion beam 10. FIG. 3 shows an exemplary planar view of the cusp coupled plasma injection system 200 in accordance with an embodiment of the present disclosure. The plasma source 201 may be embedded at the center portion of the pole piece of the bending magnet 202. The plasma extracting magnet arrays 209 may be configured to form pairs of interdigitated magnets 209A and 209B having opposite polarity. For example, magnets 209A may be north pole magnets while magnets 209B may be south pole magnets. Arrows 301A and 301B illustrate the direction of the multi-cusped magnetic field lines produced by magnets 209A and 209B, respectively. The north pole magnets 209A may generate a multi-cusp magnetic field in the direction 301A coming out of the page, while the south pole magnets 209B may generate a multi-cusped magnetic field in the direction 301B going into the page. A cusp line 302 may be demonstrated by the dotted line between the magnets 209A and 209B. Furthermore, arrows 303 illustrate the “cross-B” (x B) drift motion caused by the plasma extracting magnet arrays 209 which may direct the flow of the plasma along the cusp line 302. Three “cross-B” (x B) drift mechanisms may contribute to the flow of the plasma along the cusp line 302. The first “cross-B” (x B) drift mechanism is a diamagnetic drift which causes the plasma drift in a direction of plasma density gradient (∇n) cross the multi-cusped magnetic field generated by the plasma extracting magnet arrays 209 (B) (∇n x B) . The diamagnetic drift may be caused by a plasma density gradient between a plasma column and surrounding chamber walls. The gradient in plasma density gives rise to a drift in a direction perpendicular to both the plasma density gradient and the magnetic field, and thus directs the flow of plasma 212 along the cusp line 302. The second “cross-B” (x B) drift mechanism may be defined by the electric field (E) cross the multi-cusped magnetic field generated by the plasma extracting magnet arrays 209 (B) (E x B). The electrical field (E) may be caused by a potential variation inside the initial plasma injection region 213 and/or a potential drop from the plasma to the inner wall of a beamguide 214, which may be referred to as a “sheath”. A sheath may be formed in a region between the plasma and the adjacent wall of the beamguide 214 where electron flux to the wall is reduced to a corresponding ion flux. In addition, a sheath region may have a higher electric field than an electric field associated with the plasma. According to the sheath formation mentioned above, the E x B drift mechanism may cause an electric field perpendicular to the inner wall of the beamguide 214. A drift results in a direction perpendicular to both the electric field associated with the sheath formation and the multi-cusped magnetic field generated by plasma extracting magnet arrays 209, thus directing the flow of plasma along the cusp line 302. Finally, the third “cross-B” (x B) drift mechanism is a curvature drift (R x B) which may be defined by a radius of curvature of the multi-cusped magnetic field generated by the plasma extracting magnet arrays 209 (R) cross the multi-cusped magnetic field generated by the plasma extracting magnet arrays 209 (B) (R x B). The curvature drift (R x B) may be caused by the curvature of the magnetic field lines, through a centrifugal force effect. The curvature drift (R x B) may impart a drift motion perpendicular to both the radius of curvature and the magnetic field, and thus direct the flow of the plasma along the cusp line 302. In FIG. 4, there is shown an exemplary detailed sectional view of a cusp coupled plasma injection system 200 in accordance with an embodiment of the present disclosure. After the generation of plasma 212, an initial plasma injection region 213 from the plasma chamber 204 is limited to regions around the aperture 208 because plasma 212 tends to localize along the magnetic field lines. As explained above, the multi-cusped magnetic field generated by the plasma extracting magnet arrays 209 illustrated by curved magnetic field lines 402 causes the plasma to flow along the cusp line 302. A cusp coupled plasma injection region 403 may be formed along the cusp line 302 because of the “cross-B” (x B) drift motion. Hence, the “cross-B” (x B) drift motion may carry the plasma 212 away from the aperture 208 region and thus uniformly distribute the plasma 212 along the path of the ion beam 10. FIG. 5 illustrates an exemplary isometric view of a cusp coupled plasma injection system 200 in accordance with an embodiment of the present disclosure. As mentioned above, microwave energy accelerates electrons via electron cyclotron resonance (ECR) within the interior space of the plasma chamber 204. The electron motion ionizes the gas and thus produces the plasma 212. The plasma 212 may be injected along the path of the ion beam 10 by magnetic dipole field (B) through apertures 208. Beamguide 214 may be configured to have alternating sloping convex portions 501 and sloping concave portions 502. The apertures 208 may be formed at the concave portions 502 of the beamguide 214, while the plasma extracting magnet arrays 209 may be formed at the convex portion 501. FIG. 6 illustrates an exemplary sectional view of the cusp-coupled plasma injection system 200 in accordance with an embodiment of the present disclosure. As mentioned above, “cross-B” (x B) drift mechanisms allow the plasma 212 to propagate along the cusp line 302. The plasma has a higher density at a region around the apertures 208 and thus tends to flow out to the inner wall of the beamguide 214 where the plasma density is lower. The E cross B (E x B) drift mechanism may cause the plasma 212 to flow in the “cross-B” (x B) loop by the potential drop from the plasma to the inner wall of the beamguide 214. Finally, the curvature drift (R x B) mechanism of the “cross-B” (x B) loop facilitates the flow of the plasma through a centrifugal force effect. The combination of the diamagnetic drift, the E cross B (E x B) drift, and the curvature drift (R x B) allows the flow of the plasma in the same direction along the cusp line 302. FIG. 7 illustrates an exemplary sectional view of an RF cusp coupled plasma injection system 700 in accordance with an embodiment of the present disclosure. A plurality of magnet arrays 701 are provided in similar fashion to the plasma extracting magnet arrays 209 of FIGS. 2-6. The magnet arrays 701 may be employed within at least a portion of the ion beam 10. The magnet arrays 701 may be coupled to an electric power source 702, for example, an RF power source, as illustrated in FIG. 7. The RF power source 702 may generate a potential difference between the magnet arrays 701 to thereby generate plasma 705. The magnet arrays 701 operate as an RF electrodes. A cooperative interaction between the magnetic field and the electric field results in the plasma generation in plasma generating region 704. Also, the magnet arrays 701 may generate a multi-cusped magnetic field 703 in order to uniformly distribute plasma within the plasma generating region 704 in the path of the ion beam 10. Electrons moving within plasma generating region 704 continue moving until they collide with a gas, and at least some of the electrons have sufficient energy to ionize a portion of the gas, thereby generating a plasma 705. The RF discharge condition in plasma generating region 704 advantageously provides enhancement of the ion beam 10, whereby beam integrity is improved along the longitudinal length of the beamguide 214. The creation of the RF plasma in one or more plasma generating regions 704 around the ion beam 10 prevents beam “blow-up” by facilitating the transfer of energy to the plasma surrounding the ion beam 10, thereby enhancing the ion beam 10. It will be appreciated that the sizing, orientation, and spacing of magnet arrays 701 may allow the location of the plasma generating region 704 to be generated in accordance with desired ion beam containment goals. For example, the strength of the magnet arrays 701 may be varied by changing the distance between the inner surface of the magnet arrays 701 and the plasma generating region 704. In this manner, the distance may be adjusted according to ion beam size. In addition, the spacing between adjacent magnet arrays 701 may be changed in order to vary the spacing between adjacent plasma generating regions 704. Furthermore, the relative orientations of magnetic pole faces of adjacent magnets 701 may be varied in order to provide additive magnetic field lines between adjacent magnets 701. Many different magnet sizes, orientations, and spacings are possible and contemplated as falling within the scope of the present disclosure. FIG. 8 illustrates an exemplary planar view of the RF cusp coupled plasma injection system 700 in accordance with an embodiment of the present disclosure. The magnet arrays 701 are configured to form a pair of interdigitated magnet arrays 701A and 701B. For example, the RF power source 702 may be applied between magnet arrays 701A and 701B. The gap between magnet arrays 701A and 701B may be small comparing to the physical gap of the beamguide 214. The magnet arrays 701 may take very little space, thus the plasma generator can be made to be a small fraction of the physical gap of the beamguide 214 that may be highly advantageous for beam transportation. As is generally appreciated, plasma experiences a substantial resistance across magnetic field lines, but can readily diffuse along such field lines. Thus, plasma generated along a wall of the beamguide 214 may readily diffuse across to an opposing wall, thereby providing relatively uniform plasma. FIG. 9 illustrates another exemplary sectional view of the RF cusp coupled plasma injection system 700 in accordance with an embodiment of the present disclosure. The plasma injection system 700 may further include a shield member 901 to protect the magnet arrays 701. Without such shield member 901, any stray particles from the ion beam 10 may impinge on the magnet arrays 701. The stray particles from the ion beam 10 may damage and/or contaminate the magnet arrays 701. Thus, the shield member 901 protects the magnet arrays 701 from damages and/or contaminations from the ion beam 10. The shield member 901 may be made from suitable metal (e.g., aluminum), silicone, elastomer, and/or other suitable material. In other embodiments, the shield member 901 may be placed directly on the magnet arrays 701 in order to provide a better coverage of the magnet arrays 701. FIG. 10 illustrates an exemplary sectional view of an electrostatic plasma confinement device 1000 in accordance with an embodiment of the present disclosure. A magnetic field produced by pole piece of the bending magnet 202 creates a strong confinement in one direction as electron diffusion across the magnetic field is very limited. However, elelctrons move freely along field lines to the inner walls of beamguide 214 where the electrons are lost to the walls. The confinement device 1000 may be located on the inner walls of beamguide 214 to confine the flow of electrons. A grounded high transparency fine mesh grid 1001 made from metal may be located near the inner wall of the beamguide 214. A negatively biased plate 1002 made of a metal material may be placed behind the grounded high transparency fine mesh grid 1001. The negatively biased plate 1002 may repel beam plasma electrons that may have propagated beyond the grounded high transparency fine mesh grid 1001. The negatively biased plate 1002 may repel the beam plasma electrons back into the ion beam 10. The grounded high transparency fine mesh grid 1001 may be a delicate element and thus may further include ion scrapers 1003 to protect from damage that may be caused by impinging ions. The grounded high transparency mesh grid 1001 may be smaller than a Debye length (˜1 mm or smaller). The size of the grounded high transparency mesh grid 1001 may be critical because if the size of the grounded high transparency mesh grid 1001 is greater than the Debye length, the grounded high transparency mesh grid 1001 may be shielded by the plasma and fail to provide a ground reference for the ion beam 10. The present disclosure is not to be limited in scope by the specific embodiments described herein. Indeed, other various embodiments of and modifications to the present disclosure, in addition to those described herein, will be apparent to those of ordinary skill in the art from the foregoing description and accompanying drawings. Thus, such other embodiments and modifications are intended to fall within the scope of the present disclosure. Further, although the present disclosure has been described herein in the context of a particular implementation in a particular environment for a particular purpose, those of ordinary skill in the art will recognize that its usefulness is not limited thereto and that the present disclosure may be beneficially implemented in any number of environments for any number of purposes. Accordingly, the claims set forth below should be construed in view of the full breadth and spirit of the present disclosure as described herein.
047284854	claims	1. In a method for regulating the pressure of the primary circuit of a pressurized water nuclear reactor during shut-down phases, for use in an installation comprising, as a branch circuit to the primary circuit, a volume control circuit with a discharge valve of the primary circuit, a volume control tank, and a pump for reinjection to the primary circuit through a charging valve, the primary circuit being connected to a steamliquid pressurizer provided with a circuit for sprinkling the vapor phase controlled by a valve supplied from the charging pipe of the volume control circuit, according to which method, during cooling of the circuit by discharge of secondary steam, continuous sprinkling of cold water on the steam contained in the pressurizer is effected, the discharge flow rate is regulated as a function of a pressure measurement taken immediately upstream of the discharge valve and after an upstream expansion stage, and the charging flow rate is regulated as a function of the liquid level in the pressurizer, the improvement comprising the steps, when the liquid level reaches the upper part of said pressurizer where level detection becomes inoperative, of (a) maintaining a constant flow rate at the charging pipe; (b) regulating the discharge flow rate by rapid acting regulation from direct measurement of the pressure of said primary circuit; (c) performing a first one of a series of successive sprinklings at said pressurizer to reduce the discharge flow rate to a reference value; (d) detecting after said first sprinkling the reduction in discharge flow rate which is immediately caused by the pressure reduction due to the condensation in the pressurizer; (e) continuing alternatingly successive sprinklings and detections until further sprinkling has no further effect on the discharge flow rate; and (f) continuing cooling of the pressurizer with pressure regulation by said discharge valve. 2. The improvement according to claim 1, comprising the step of rendering said method automatic by servo-coupling the flow rate of the sprinkling control valve from the measurement variations in the discharge flow rate, by using a regulation with a slower reaction speed than that of the regulation of the discharge valve from the direct measurement of the pressure in the primary circuit.
	description	This application claims priority under 35 U.S.C. §119(e) from Provisional Application Ser. No. 61/756,136, entitled “Retracting Finger Spent Fuel Assembly Handling Tool,” filed Jan. 24, 2013. 1. Field The invention relates generally to equipment used to transport nuclear fuel assemblies within a nuclear power generating facility, and more particularly to such a piece of equipment that will not get tangled in the components of a top nozzle of a fuel assembly as the equipment is being aligned to grip the top nozzle. 2. Related Art In a nuclear reactor power plant, one design of a fuel assembly is comprised of a plurality of fuel elements or rods oriented in a square-shaped array. For a typical pressurized water reactor, there are on the order of approximately 200 to 300 of these elongated fuel rods in each fuel assembly. At either end of the fuel assembly is a top and bottom nozzle which direct the flow of coolant, typically water, through the fuel assembly. Interspersed among the fuel rods are hollow tubes, or thimbles, into which control rods are inserted. The control rods contain neutron absorbent material and are moved into and out of the plurality of guide thimbles to help control the nuclear reaction. These fuel assemblies also contain a centrally located instrumentation tube which allows the insertion of in-core instrumentation during reactor operation. The thimbles and instrumentation tube project between the top and bottom nozzles. Between the top and bottom nozzles, a plurality of spacer grids are positioned at intervals to provide lateral support for the fuel rods. The top nozzle is positioned at the upper end of the fuel assembly and connects to one end of the guide thimbles to allow the load of the entire fuel assembly to be carried from the bottom nozzle, which is connected to the other end of the guide thimbles, with the fuel assembly weight transferred up the guide thimbles to the top nozzle. The fuel assembly top nozzle and bottom nozzle are configured to aid in channeling coolant through the assembly during operation. In this configuration, the weight of the fuel rods is born by the guide thimbles and not by the fuel rods when the fuel assembly is lifted by the top nozzle. In an equilibrium core a typical fuel assembly will see three operating cycles before it is removed from the reactor and transported under water through a refueling canal to a spent fuel pool in a separate spent fuel building outside the reactor containment. When handling the fuel assembly, a fixture such as a refueling mast or other overhead crane is positioned over the reactor after the reactor head and upper internals are removed and connected to the top nozzle of the fuel assembly. The fuel assembly is then lifted from the core by the refueling machine which transports the fuel assembly under water through a flooded area in the containment above the reactor vessel, to a fuel assembly transport cart. The transport cart translates the fuel assembly to a horizontal position so it can pass through a refueling canal which connects to the spent fuel pool. A separate fuel handling machine in the spent fuel building uprights the fuel assembly and transports it to an appropriate location within the racks within the spent fuel pool. Existing designs of spent fuel assembly handling tools built for certain styles of fuel assemblies, such as that described above, include gripper fingers at a fixed elevation below a tool head of the handling tool. These gripper fingers pivot between a latched and unlatched position by raising and lowering an actuator. This design requires an operator to lower the tool onto a fuel assembly until the tool is resting on the top nozzle of the fuel assembly. The existing tools incorporate two alignment “S-pins” that must be inserted in two alignment “S-holes” on the top nozzle by a skilled technician. If the alignment of these pins to the holes is incorrect, the tool can be lowered in an orientation in which the gripper fingers contact or interfere with the top nozzle hold down springs. Such interference can cause the gripper fingers to become locked under the hold down springs requiring non-normal recovery efforts. In the Spring of 2012, a refueling machine gripper was lowered onto a fuel assembly and became stuck due to gripper finger to top nozzle interaction, which caused a seven-day delay in the refueling outage. During the Fall of 2012, a spent fuel tool lowered onto a fuel assembly became stuck due to finger to top nozzle interaction resulting in an 18-hour delay. The refueling operation usually determines the critical path for an outage during which replacement power has to be purchased at a relatively high cost. Anything that delays the refueling process is to be avoided wherever possible. Accordingly, it is an object of this invention to provide a fuel handling tool design that will not adversely get caught up in the components of the top nozzle. Additionally, it is a further object of this invention to provide such a tool design that is simple to operate. These and other objects are achieved by a fuel assembly handling tool having a bail configured to be connected to an overhead crane or other hoist. A bail plate is connected to and is freely supported by the bail. A tool body is freely supported at an upper end from the bail plate and extends between the bail plate and a lower end with the length between the upper and lower ends being gauged to access a top nozzle of a fuel assembly. A tool head is connected to the lower end of the tool body and is sized to house a gripper assembly in a withdrawn position so that the gripper assembly is out of contact with the top nozzle of the fuel assembly when the tool head contacts or otherwise rests on the top nozzle. The gripper assembly is operable to extend below its withdrawn position to an extended position to grip a portion of the top nozzle of the fuel assembly to support the fuel assembly as the crane or other hoist lifts the bail. Preferably, an actuator arm is accessible from the bail plate and is operable to extend or withdraw the gripper assembly to the extended or withdrawn position. In one embodiment, the gripper assembly is moved to the withdrawn position or the extended position by respectively raising or lowering the actuator arm in a linear motion. In one such embodiment, the gripper assembly fully grips the fuel assembly as the actuator arm is lowered. Desirably, the gripper assembly positively locks in the fully withdrawn and fully extended positions. In another embodiment, the fuel assembly handling tool includes guide pins extending down from the tool head for aligning the tool head with the fuel assembly top nozzle. Referring to the drawings, in particular FIG. 1, there is shown an elevational view of a nuclear fuel assembly, of the type employed in pressurized water reactors, represented in vertically shortened form and generally designated by reference character 10. The fuel assembly 10 has a structural skeleton which, at its lower end, includes a bottom nozzle 12. During the operating life of the fuel assembly 10, the bottom nozzle 12 supports the fuel assembly 10 on a lower core support plate 14 in the core region of the nuclear reactor (not shown). In addition to the bottom nozzle 12, the structural skeleton of the fuel assembly 10 also includes a top nozzle 16 at its upper end and a number of guide tubes or thimbles 18 which extend longitudinally between the bottom and top nozzles 12 and 16 and at opposite ends are rigidly attached thereto. The fuel assembly 10 further includes a plurality of transverse grids 20 axially spaced along and mounted to the guide thimbles 18 and an organized array of elongated fuel rods 22 transversely spaced and supported by the grids 20. Also, the assembly 10 has an instrumentation tube 24 located in the center thereof and extending between and mounted to the bottom and top nozzles 12 and 16. With such an arrangement of parts, fuel assembly 10 forms an integral unit capable of being conveniently handled without damaging the assembly of parts. As mentioned above, the fuel rods 22 in the array thereof in the fuel assembly 10 are held in spaced relationship with one another by the grids 20 spaced along the fuel assembly length. Each fuel rod 22 includes nuclear fuel pellets 26 and is closed at its opposite ends by upper and lower end plugs 28 and 30. The pellets 26 are maintained in a stack by a plenum spring 32 disposed between the upper end plug 28 and the top of the pellet stack. The pellets 26, composed of fissile material, are responsible for creating the reactive power of the reactor. A liquid moderator/coolant such as water or water containing boron is pumped upwardly through apertures 42 in the lower core support plate 14 to the fuel assembly 10. The bottom nozzle 12 of the fuel assembly 10 passes the coolant upward along the fuel rods 22 of the assembly in order to extract heat generated therein for the production of useful work. The coolant exits the core through apertures in an upper core plate (not shown) that sits over the fuel assembly. Hold down springs 40 that extend up from the top nozzle 16 seat against the underside of the upper core plate and serve to hold down the fuel assembly, counteracting the upward force exerted by the flowing coolant. To control the fission process, a number of control rods 34 are reciprocally movable in the guide thimbles 18 located at predetermined positions in the fuel assembly 10. Specifically, a rod cluster control mechanism 36 positioned above the top nozzle 16 supports the control rods 34. The control mechanism has an internally threaded cylindrical member 37 which is coupled to a drive rod (not shown) and a plurality of radially extending flukes or arms 38. Each arm 38 is interconnected to a control rod 34 such that the control rod mechanism 36 is operable to move the control rods vertically in the guide thimbles 18 to thereby control the fission process in the fuel assembly 10, all in a well known manner. To refuel such a reactor, the refueling area above the reactor is flooded, the reactor head and upper internals are removed, including the upper core plate, to expose the fuel assemblies. A refueling machine is then lowered and actuated to couple with the top nozzle 16 of the fuel assembly, gripping on an upper lip 44. FIG. 2 shows one embodiment of a refueling machine which incorporates the principles of this invention for coupling to and lifting a fuel assembly without the risk of entangling the gripping fingers of the lifting device in the hold down springs 40 or other component parts of the top nozzle. The fuel assembly handling machine 46, shown in FIG. 2, basically includes a bail 48 that is connected to a bail plate 60 that supports the remainder of the tool. The bail 48 is designed to be connected to a hook on an overhead crane that will raise and lower the tool 46. In addition to the bail 48 and bail plate 60, the tool includes a handle 52 that raises and lowers to operate the gripping feature of the tool, a mechanism 62 employed to separately lock the handle in an “engaged” and “disengaged” position, a long and slender tubular tool body 54 and a tool head 56. Alignment pins 58 extend from the lower part of the tool head and are designed to be inserted into corresponding openings in the upper surface of the top nozzle 16 of the fuel assembly 10. The tool head 56 in this embodiment has four gripper fingers 64, three of which can be observed in FIGS. 3 and 4 with all four gripper fingers shown in the perspective view shown in FIG. 6. Though it should be appreciated that any number of gripping fingers may be used. In accordance with this invention, the gripper fingers 64 are retracted in a storage position within the tool head 56 when in a “disengaged” or “unlatched” position as shown in FIG. 5. The gripper fingers 64 are lowered and rotated into place when actuated to the “engaged” or “latched” position. Actuation of the tool is achieved by a single linear motion of the tool handle 52. As can be seen from the cross sectional view shown in FIG. 3, the tool bail 48, tubular housing body 54 and tool head frame 56 are all fixed and welded or bolted together. The handle 52 is pinned to an actuating rod which is slidably movable within the tool body 54 and extends substantially the whole length of the tool. The actuating rod 50 is attached at its lower end to an actuator 66, which can better be seen in FIG. 4. The actuator 66 is connected to four “carriers” 68. The fingers 64 are respectively pinned at 70 to the corresponding carrier 68, but are able to rotate. Each finger 64 has a lobe 72 that rides in a cam slot 74. As the handle 52 is raised, the actuator rod 50 to which the handle is coupled, raises the actuator 66, the four carriers 68 and the four fingers 64. As the fingers 64 are raised by raising the handle 52, the lobes 72 on each finger ride in the cam slots which rotate the fingers in a specific way determined by the contour of the cam slot. In the image shown in FIG. 4, the tool is shown in an “engaged” position. As the handle 52 is raised, the fingers 64 would raise slightly while rotating approximately 10° in the first half-inch or so of travel. Then, as the handle continues to rise, the gripping fingers rise vertically because the cam slots are vertical from that point up. As previously stated, FIG. 5 shows the tool head 56 in a completely “disengaged” condition with the fingers 64 completely retracted within the tool head housing 76. FIG. 6 shows the fingers 64 in a fully extended and “engaged” condition. Though, the invention was shown in an embodiment in which four fingers are employed and the disengaged and engaged positions are achieved by way of a single linear motion of the tool handle, to either fully retract or fully extend the gripping fingers relative to the tool head housing, other arrangements are contemplated in which those steps can be separately carried out with any number of grippers. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	summary
	abstract	Electron-beam-induced chemical reactions with precursor gases are controlled by adsorbate depletion control. Adsorbate depletion can be controlled by controlling the beam current, preferably by rapidly blanking the beam, and by cooling the substrate. The beam preferably has a low energy to reduce the interaction volume. By controlling the depletion and the interaction volume, a user has the ability to produce precise shapes.
	description	The present invention pertains to multi focal spot collimators. More particularly, the present invention pertains to multi focal spot collimators for x-rays. X-ray imaging systems have become invaluable in the medical field for a variety of surgical and diagnostic purposes. The implementation of many cardiac, urological, orthopedic, peripheral vascular, and a variety of non-invasive surgical procedures rely on the ability of the surgeon or medical authority to clearly track an implement they have inserted into a patient, such as a catheter, or otherwise monitor a region of interest within the patient through fluoroscopy. An example of a known fluoroscopy system is U.S. Pat. No. 2,730,566 issued to Bartow, et. al. entitled “Method and Apparatus for X-Ray Fluoroscopy. Computer Tomography (CT), in which a moving source-detector pair takes numerous two-dimensional images while rotating around a patient for reconstruction, is one of the preeminent methods of generating three-dimensional internal images used for cancer, other disease, and injury diagnoses. Single tomographic x-ray images are valuable for analysis as well. The process of generating an x-ray image of a region of interest entails the positioning of a patient between an x-ray source and an x-ray detector, emission of x-rays from the x-ray source, the travel of these x-rays through a targeted volume of the patient, and the absorption of these x-rays by the x-ray detector. Since areas of a patient which are x-ray dense—notably, bones or vessels and tissues which have been highlighted by insertion of a contrast element—will absorb or scatter incident x-rays, the amount of x-ray photons reaching a given point on the x-ray detector corresponds to the x-ray density of the patient along a line between the x-ray source and that point on the detector. Therefore, intensity information from the detector can be used to reconstruct an image of the area of the patient through which the x-rays travelled. Increasing the x-ray flux can improve image quality by increasing the amount of x-rays photons that pass through the patient and reach the detector, hence increasing the amount of intensity data available for image reconstruction. However, in addition to image quality considerations, decisions surrounding the x-ray flux are concerned with avoiding unnecessary exposure of the patient and attending medical personnel to x-ray radiation. While exposure of tissue to an extremely high amount of radiation at a given time would be necessary to see immediate negative health reactions such as radiation burns, a few relatively heavy doses to a patient or perpetual smaller doses to medical personnel may significantly increase probability of cancer later in life. To maintain an x-ray flux sufficient for the generation of high-quality images while reducing x-ray exposure to system surroundings, an x-ray dense unit with a single aperture is generally positioned against the face of the x-ray source so that x-rays travelling along paths which, if uninterrupted, would not strike the detector face will be absorbed within its volume. The process of selectively attenuating x-rays is referred to as collimation, and the attenuating unit as a collimator. Detector photon counts from absorption of scattered x-rays, which lower the image quality by contributing incorrect intensity information, are referred to as scatter noise. Systems have been developed with an “inverse geometry” such that the face of the x-ray source is relatively large and the face of the detector relatively small compared to conventional systems. Inverse geometry systems suffer significantly less from scatter noise as a smaller detector face decreases the probability of scattered ray absorption. A notable type of inverse geometry systems is the scanning x-ray beam system such as the one disclosed in U.S. Pat. No. 5,729,584 entitled “Scanning Beam X-Ray Imaging System.” In scanning beam systems, x-ray beams are sequentially emitted from different points on the source, called focal spots, at very high speed rather than from the entire source face simultaneously. Since a number of images (corresponding to the number of emissive points on the source face) are used to reconstruct a single frame, the amount of patient volume exposed to x-rays at a given time, namely a narrow cone connecting a single aperture and the detector face, can be small compared to non-scanning systems where the entire target volume is continuously exposed. Scatter noise may be even lower in scanning beam systems as at a given time, scatter can only occur within this narrow illuminated cone rather than anywhere in the target volume. Information regarding the angular dependence of scanning beam images can also be used to add a three-dimensional, or tomographic, quality to the frames. Non-conventional collimation devices are necessary for inverse geometry, scanning beam, and other multi focal spot x-ray imaging systems for a variety of reasons. A multi focal spot collimator must direct x-rays from a source of large surface area to a small detector rather than from a small source to a large detector. This generally requires a plurality of closely-spaced apertures, each angled and shaped to emit x-rays that will intersect the detector face when illuminated by the source and attenuate x-rays that would spill around the detector face. Furthermore, in scanning beam systems, image reconstruction techniques rely on the assumption that x-rays are being emitted through only the intended aperture or intended apertures when a focal spot illuminates the collimator. Additionally, while many single focal spot sources contain an x-ray reflective element so that the emissive portion of the source is positioned farther back in the body of the source, inverse geometry systems may require transmissive sources in which the target screen is the most outward element of the source. Material being constantly struck with high energy electrons and emitting Bremsstrahlung x-ray radiation will overheat without some sort of cooling system. Fast-moving, coolant fluid which absorbs and carries away excess heat is the key element in many cooling systems. Thus, in a system with a transmissive source, the collimator can be in contact with a coolant fluid system. As a transmissive source may control the position of an electron beam with an applied magnetic field, any external electromagnetic fields may alter the beam path and disrupt the proper functioning of the x-ray source. While the balance between x-ray image quality and dose control, improved by collimated multi focal spot systems, is particularly relevant in medical applications as discussed above, it can also be relevant in baggage screening, security applications, and other x-ray imaging applications. In one embodiment of the present invention, a multi focal spot x-ray collimator based on two subassemblies—a subassembly that reduces the amount of x-ray leakage between apertures and a subassembly that reduces the amount of x-ray radiation that doesn't strike the detector face after emission through the intended aperture(s)—is provided. These subassemblies both have apertures through which x-rays may pass. The subassembly that reduces x-ray leakage can be made up of a number of material sheets where each sheet has a thickness of at least 0.5 mm, can be made of a material with an atomic number of at least thirty-nine, can be made of a material with a value of Young's modulus of at least 200 GPa, can be made of tungsten, or can made to have thickness of at least 1 mm. The subassembly that reduces x-ray radiation spill around the detector face can be made of a number of material sheets where each sheet has a thickness of at least 0.5 mm, can be made of a material with an atomic number between eleven and thirty-eight, can be made of a material with relative magnetic permeability of at least 5,000, can be made of mu-metal, can be made of brass, can be made of steel, or can be made to have a thickness of at least 5 mm. In another embodiment, a further subassembly is positioned in the collimator so that it is the subassembly nearest the x-ray source. This subassembly has numerous apertures, has a thickness of at least 0.5 mm, and is made from a material having an atomic number of at least 39. In another embodiment, a further subassembly is positioned in the collimator so that it is the subassembly farthest from the x-ray source. This subassembly has numerous apertures, has a thickness of at least 1 mm, is made from a material having an atomic number of at least 39. This subassembly can be positioned so that it is separated by an air gap from an adjacent subassembly or can have apertures shaped such that an aperture entrance is smaller than an aperture exit. These and other objects and advantages of the various embodiments of the present invention will be recognized by those of ordinary skill in the art after reading the following detailed description of the embodiments that are illustrated in the various drawing figures. Reference will now be made in detail to embodiments of the present invention, examples of which are illustrated in the accompanying drawings. While the invention will be described in conjunction with these embodiments, it will be understood that they are not intended to limit the invention to these embodiments. On the contrary, the invention is intended to cover alternatives, modifications and equivalents, which may be included within the spirit and scope of the invention as defined by the appended claims. Furthermore, in the following detailed description of embodiments of the present invention, numerous specific details are set forth in order to provide a thorough understanding of the present invention. However, it will be recognized by one of ordinary skill in the art that the present invention may be practiced without these specific details. In other instances, well-known methods, procedures, components, and circuits have not been described in detail as not to unnecessarily obscure aspects of the embodiments of the present invention. FIG. 1 is a diagram illustrating the elements of a multi focal spot x-ray beam system utilizing a collimator of one embodiment of the present invention. A focal spot is an area on a face of an x-ray source from which x-rays may be emitted. Hence, a multi focal spot system may entail an x-ray source configured to emit x-rays through one or more number of points, in contrast to a single focal spot system where the x-ray source may only emit x-rays from a single contiguous area. A multi focal spot x-ray source may be an emissive target screen such as a tungsten sheet on which a high energy electron beam is directed to excite the various points. As shown in FIG. 1, collimator 1 may be attached, or placed very near, the end of x-ray source 2 through which x-rays are emitted. Collimator 1 may have a pattern of holes, or apertures, such that when a given focal spot is illuminated by source 2, corresponding individual aperture 5 projects a beam of x-rays 6 toward detector 3. The details of one multi focal spot x-ray system are described in U.S. Pat. No. 5,835,561 issued to Moorman et al. entitled “Scanning beam x-ray imaging system,” herein fully incorporated by reference. The image quality of x-ray images can increase with the number of x-rays incident on the detector face. This may be particularly true in “inverse geometry” systems, such as a scanning beam system, where the detector is significantly smaller than conventional systems and therefore intercepts very few quality-degrading scattered x-ray beams. However, simply increasing the number of x-rays emitted by the source may not be beneficial since beams which are not fully absorbed within the detector not only increase the dose to the patient without image quality benefits but also may be absorbed by attending personnel. A large amount of x-ray exposure, either in a few large doses or many smaller doses over time, has been shown to have potentially negative health effects such as an increased risk for the development of cancer. In order to maintain high image quality while minimizing potentially harmful x-ray exposure to the patient and medical personnel in the vicinity of an x-ray imaging system, it may be desireable that the cross section of beam 6 in the plane of the detector face entail as much area inside and as little area outside of the detector face as possible. X-rays that either escape or pass through the collimator but do not intersect the detector face are referred to as spill. FIG. 2 is a diagram in which the circular points 21 represent points of intersection between x-rays in beam 6 and the detector face 23 of detector 3, and the triangular points 22 represent points of intersection between x-rays in beam 6 with area outside of the detector face 23, i.e. spill. In multi focal spot collimators, an additional problem can arise as leakage. Leakage is the passage of x-rays through some volume of collimator outside of an intended aperture. In a collimator designed for a single focal spot source, leakage is essentially a form of spill and can be easily reduced if not eliminated by increasing the dimensions of the collimator to the point where an x-ray travelling outside of the aperture has little to no chance of penetration. However, reaching similarly sufficient dimensions in multi focal spot collimators becomes unwieldy, especially in cases where the pitch, the distance between adjacent focal spots, is very small. FIG. 3 is a diagram of the paths of errant x-rays through a single focal spot collimator, and FIG. 4 is a diagram of the paths of errant x-rays through a multi focal spot collimator of equal length along the source-detector axis. In FIG. 3, x-rays from a single focal spot source that do not pass through the entrance to the collimator aperture or are angled very steeply relative to a forward direction of travel must follow paths through a significant depth of collimator material to escape and therefore have a high probability of being scattered or absorbed within the collimator. In FIG. 4, x-rays from a single focal spot which fall outside of a corresponding aperture entrance or are steeply angled may escape by following paths requiring travel through only short depths of collimator material, over which there is a low probability of scatter or absorption. An advantage of embodiments of the present invention is the flexibility to address spill control and leakage control separately through independent subassemblies. Separating these functions allows the designer to more easily select or optimize material, aperture shape, and fabrication method for each function. X-ray interaction with materials is in large part determined by the atomic number of the materials. Atomic number, the characteristic number of protons in the nuclei of elemental atoms (and also the number of surrounding electrons if the atoms are stable and charge-neutral), determines the density of charged particles in a material. The probability that an x-ray will interact with a charged particle and lose some of its energy increases with the density of charged particles so materials with a high atomic number are more likely to attenuate x-ray radiation. These materials tend to be more costly and weigh significantly more than materials with a lower atomic number so the ability to choose a material with an atomic number appropriate to a specific attenuation strength may have weight and cost benefits. High Z materials are materials with high atomic numbers e.g. an atomic number of at least thirty-nine, and lower Z materials are materials with low atomic numbers e.g. an atomic number greater than ten and less than thirty-nine. In an embodiment of the present invention, a subassembly with the function of leakage control may be constructed from a high Z material such that it will attenuate errant x-rays within a distance similar to the pitch e.g. a material with an atomic number of at least 39 or alternatively 40, 41, 42, 46, 47, 48, 49, 50, 51, 52, 55, 56, 73, 74, 77, 78, 79, 80, 82 or 83 or any range of atomic numbers between 39 and 83. For example, lead is one high Z material that would suffice for leakage control. The subassembly may be composed of a number of 0.5 mm thick plates or alternatively 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14 or 15 mm or any thickness between 0.5 and 15 mm or any range of thickness between 0.5 and 15 mm. Two to thirty plates may be layered to comprise the subassembly or a single plate can be used or any range of number of plates between one and thirty plates. The shape and size of apertures through the leakage control subassembly may be a system-specific design consideration. The apertures may be holes of standard shapes such as circles or squares or have less regular edge geometries. In one embodiment of the present invention, the apertures are round or can be a constant width or radius through the thickness of the leakage subassembly in order to consistently reduce the passage of x-rays between adjacent apertures. The method of creating apertures through the leakage control subassembly may be chemical etching, an electrical discharge machining method, or standard drilling or milling. In an embodiment of the present invention in which the leakage control subassembly is comprised of lead sheets, apertures can be created using chemical etching. Spill may be reduced by incorporation of a functional subassembly with the specific purpose of spill control. This spill control subassembly may be constructed out of a lower Z material since it can attenuate x-rays over the length of the collimator, which may be long compared to the pitch e.g. a material with an atomic number greater than ten and less than thirty-nine or alternatively 11, 12, 13, 14, 15, 16, 17, 19, 20, 22, 24, 25, 26, 27, 28, 29, 30, 31, 32, 33, 34, 35 or 38 or any range of atomic numbers between 11 and 38. Steel and brass are two examples of lower Z materials that would be sufficient for spill control. The spill control subassembly may also be composed of 0.5 mm plates or alternatively 1, 2, 5, 7, 10, 15, 20, 25, 30, 35, 40, 45, 50 or 55 mm or any thickness between 0.5 and 55 mm or any range of thickness between 0.5 and 55 mm. The number of plates may range between ten and 110 or a single plate can be used or any range of number of plates between 10 and 110. The shape and size of apertures through the spill control subassembly may be a system-specific design consideration. The apertures may be holes of standard shapes such as circles or squares or have less regular edge geometries. In one embodiment of the present invention, the width or radii of apertures linearly increase through the thickness of the subassembly from a smallest width or radius at the aperture entrance to a largest width or radius at the aperture exit. The method of creating apertures through the spill control subassembly may be chemical etching, an electrical discharge machining method, or standard drilling or milling. When the spill control subassembly is made of brass or steel, apertures may be created using chemical etching. FIG. 5 illustrates an embodiment of the present invention, a collimator comprising just two functional subassemblies, a spill control subassembly 41 and a leakage control subassembly 42. It can be seen that spill control subassembly 41 is constructed from ten plates of a lower Z material such as brass, and the leakage control subassembly 42 is constructed from five plates of high Z material such as lead. The lower Z material can have an atomic number greater than ten and less than thirty-nine or alternatively 11, 12, 13, 14, 15, 16, 17, 19, 20, 22, 24, 25, 26, 27, 28, 29, 30, 31, 32, 33, 34, 35 or 38 or any range of atomic numbers between 11 and 38. The high Z material can have an atomic number of at least 39 or alternatively 40, 41, 42, 46, 47, 48, 49, 50, 51, 52, 55, 56, 73, 74, 77, 78, 79, 80, 82 or 83 or any range of atomic numbers between 39 and 83. Additional problems intrinsic to multi focal point collimation may be addressed by constructing the two plates of the FIG. 5 embodiment out of specific materials and/or adding further subassemblies. Transmissive x-ray sources may be comprised of a beam of high energy electrons directed at an emissive target screen. If the path of the electron beam is controlled by an applied magnetic field, it may be necessary to magnetically shield the x-ray source to prevent external magnetic forces from redirecting the beam. Magnetic permeability is a measure of the tendency of a material to become magnetized and can be quantified in units such as henries per meter. Relative magnetic permeability simply refers to a magnetic permeability value which has been divided by the magnetic permeability of free space and is thus unit-less. If a magnetically permeable material is placed in an external magnetic field, it becomes magnetized and draws the force of that magnetic field to itself. Therefore, a volume of magnetically permeable material can terminate a magnetic field before it reaches some unwanted location. This is one method of magnetic shielding. While intrinsically permeable materials may be used for magnetic shielding, the magnetic properties of some other materials may be altered by heat and other treatment methods and can also become suitable for magnetic shielding purposes. A material is considered magnetically permeable rather than transparent if its relative permeability is greater than one, but as materials can be found with very high permeability values, a material with a relative permeability greater than 10,000 may be chosen for magnetic shielding applications. It is also desirable that the material be magnetically “soft,” i.e. quick to release magnetization once a field is removed, so that the shield responds quickly to changes in magnetic environment. Magnetic shielding may be incorporated as a function in an embodiment of the present invention by adding a further subassembly made of magnetically permeable material or other magnetic shielding material or by fabrication of the aforementioned spill reduction plates out of a lower Z material that is magnetically permeable or otherwise suited for magnetic shielding. Mu-metals, a class of nickel-iron alloys with relative magnetic permeability values between 80,000 and 100,000, comprise one class of materials from which either of these subassemblies may be fabricated. Nickel has an atomic number of twenty-eight and iron an atomic number of twenty-nine. Possible additions to mu-metal alloys are molybdenum and copper, which have atomic numbers of twenty-six and forty-two respectively. Other materials can be used with relative magnetic permeability of at least 100 or values between 100 and 1,000,000 or any range of relative magnetic permeability between 100 and 1,000,000. If a separate magnetic shielding subassembly is incorporated into the collimator, the shape and size of apertures through it may be a system-specific design consideration. The apertures may be holes of standard or non-standard shapes with radii or width as large or larger than the desired x-ray beam radius in the plane of the subassembly and small enough that the subassembly mimics the shielding properties of a continuous sheet. The method of creating these apertures may be chemical etching, an electrical discharge machining method, or standard drilling or milling. If the function of magnetic shielding is incorporated into the spill control subassembly in the collimator, the shape and size of apertures may be determined by the previously discussed beam-shaping considerations and machined using chemical etching, an electrical discharge machining method, or standard drilling or milling. In an embodiment of the present invention in which a subassembly with the function of spill control and magnetic shielding is made from mu-metal, the apertures through the subassembly may be created via chemical etching. X-ray imaging systems such as “Scanning beam x-ray imaging system” and others which utilize transmissive x-ray sources such as the one described in U.S. Pat. No. 5,682,412 entitled “X-ray Source,” and herein incorporated by reference, can require stabilization against the pressure applied by a fluid-based coolant system because the collimator will be in contact not only with the emissive target screen but also a coolant fluid system. The collimator must be able to withstand the pressure from adjacent fast-flowing coolant or be otherwise stabilized. Without some sort of stabilization, elements in contact with the flowing coolant can bow. The tendency of a material to bow decreases as its stiffness increases. The stiffness of a material relates to the amount of strain, the amount of deformation relative to its original dimensions, exhibited by the material when an external stress is applied and is characterized by a quantity called Young's modulus. Stabilization may be incorporated by the addition of a further subassembly made of a sufficiently stiff material or by constructing the leakage control subassembly from a sufficiently stiff, high Z material e.g. a material with an atomic number of at least 39 or alternatively 40, 41, 42, 46, 47, 48, 49, 50, 51, 52, 55, 56, 73, 74, 77, 78, 79, 80, 82 or 83 or any range of atomic numbers between 39 and 83. The value of Young's modulus required to sufficiently stabilize a system may depend on the thickness of the subassembly as well as the properties of the coolant fluid system, and may be at least 200 GPa. Alternatively, a material with Young's modulus of 150, 150-185, 159, 181, 193, 200, 190-210, 207, 248, 276, 287, 329, 345, 400-410, 435, 450, 450-650, 517, 550, 1000, 1050-1200, 1220 GPa or values between 150 and 1220 GPa or any range between 150 and 1220 GPa can be used. Carbon fiber, diamond, silicon carbide, steel, tungsten, tungsten carbide, iron, silicon, beryllium, molybdenum, sapphire, osmium, graphene, chromium, iridium, or tantalum can be used. A subassembly for stabilization (and leakage control) may be constructed as a solid layer of thickness greater than 2 mm and less than 1.2 cm or any range of thickness between 2 mm and 1.2 cm. If a separate stabilization subassembly is incorporated into the collimator, the subassembly may be made from stainless steel. The shape and size of apertures through a separate stabilization subassembly may be a system-specific design consideration. The apertures may be holes of standard shapes with radii or width as large or larger than the desired x-ray beam radius and small enough that the subassembly maintains a degree of stiffness sufficient to prevent bowing under pressure from a cooling fluid. The method of creating these apertures may be chemical etching, an electrical discharge machining method, or standard drilling or milling. If the function of stabilization is incorporated into the leakage control subassembly in the collimator, the subassembly may be made from tungsten. Tungsten has an approximate Young's modulus between 400 GPa and 410 GPa and an atomic number of 74. The shape and size of apertures through a stabilizing leakage control subassembly may be determined by the previously discussed x-ray leakage considerations and machined using chemical etching, an electrical discharge machining method, or standard drilling or milling. In an embodiment of the present invention in which a subassembly with the function of leakage control and stabilization is made from tungsten, the apertures through the subassembly are created using an electrical discharge machining drill. In another embodiment of the present invention, a subassembly is added to the face of the collimator nearest the source with the function of providing preliminary x-ray focusing such as the attenuation of x-rays emerging from the source completely unaligned with any particular aperture. The subassembly may be a layer of high Z material e.g. a material with an atomic number of at least 39 or alternatively 40, 41, 42, 46, 47, 48, 49, 50, 51, 52, 55, 56, 73, 74, 77, 78, 79, 80, 82 or 83 or any range of atomic numbers between 39 and 83. Its thickness can be greater than 0.5 mm or alternatively 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14 or 15 mm or any thickness between 0.5 and 15 mm or any range of thickness between 0.5 and 15 mm. For ease of reference, this subassembly will be referred to as an entrance plate in further descriptions. The shape and size of apertures through the entrance plate may be a system-specific design consideration. The apertures may be holes of standard shapes such as circles or squares or have less regular edge geometries. The radii or width of the apertures may be larger than the radii or width of apertures in subsequent collimator subassemblies. The method of creating apertures through the entrance plate may be chemical etching, an electrical discharge machining method, or standard drilling or milling. In embodiments of the present invention in which the entrance plate is made of lead, apertures may be created using chemical etching. In another embodiment of the present invention, a subassembly is added to the face of the collimator farthest from the source with the function of providing a shield against x-rays which, after passing through the rest of the collimator, maintain a path of travel that would not strike the detector face if uninterrupted. This subassembly may be comprised of a layer of high Z material e.g. a material with an atomic number of at least 39 or alternatively 40, 41, 42, 46, 47, 48, 49, 50, 51, 52, 55, 56, 73, 74, 77, 78, 79, 80, 82 or 83 or any range of atomic numbers between 39 and 83. Alternatively, this subassembly may be composed of a lower Z materials, e.g. a material with an atomic number greater than ten and less than thirty-nine or alternatively 11, 12, 13, 14, 15, 16, 17, 19, 20, 22, 24, 25, 26, 27, 28, 29, 30, 31, 32, 33, 34, 35 or 38 or any range of atomic numbers between 11 and 38. Its thickness can be greater than 1 mm or alternatively 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14 or 15 mm or any thickness between 1 and 15 mm or any range of thickness between 1 and 15 mm. For ease of reference, this final layer of spill reduction will be referred to as an exit plate in further descriptions. The shape and size of apertures through the exit plate may be a system-specific design consideration. The apertures may be holes of standard shapes such as circles or squares or have less regular edge geometries. In one embodiment of the present invention, the radii or width of apertures linearly increase through the thickness of the subassembly from a smallest radius or width at the aperture entrance to a largest radius or width at the aperture exit. The radius or width at the aperture entrance may be as large or larger than the aperture exit of the spill control subassembly or other subassembly positioned adjacent to the exit plate. The method of creating apertures through the exit plate may be chemical etching, an electrical discharge machining method, or standard drilling or milling. In embodiments of the present invention in which the exit plate is comprised of lead, apertures may be created using chemical etching. An embodiment of the present invention may be suitable for use in a system with a rectangular x-ray detector, where one dimension of the detector face is longer than other dimension of the face and longer that the dimension of square detector faces used in conventional scanning beam systems. In this embodiment, the apertures through the exit plate may be rectangular, where the long dimensions of the apertures corresponds to the long dimension of the detector. The length of the long dimension of the apertures required for rectangular beam collimation may increase with increases in detector length or with decreases in the distance from the source face to the detector face. For some geometries, the required aperture width may be as wide or wider than the pitch so that apertures within a long-dimension row “overlap,” forming a slot rather than a series of holes. Therefore, in a further embodiment of the present invention suitable for use with a rectangular detector, apertures through the exit plate may be comprised of slots. In this embodiment, the exit plate may control spill only along the short dimension of the detector as significant material along the long dimension has been removed. It may therefore be desirable to increase the amount of spill control along the long dimension in planes closer to the source by adding additional spill control subassemblies or using more highly attenuating materials for near-source spill control subassemblies. FIG. 6 illustrates a side-view vertical cross-section of an approximate configuration of one embodiment of the present invention which combines four of the functional subassemblies described above. Beginning from the side of the collimator nearest the x-ray source, the configuration is comprised of entrance plate 51 comprised of two 0.5 mm lead sheets with aperture pattern of squares fabricated by chemical etching; stabilization and leakage control plate 52 comprised of a 6.5 mm layer of tungsten with aperture pattern of squares fabricated with an electrical discharge machining drill; magnetically shielding spill control plates 53 comprised of twenty-one intermixed mu-metal and lead sheets with aperture pattern of squares fabricated using chemical etching; an air gap 54 of 1.5 cm in length; and an exit plate 55 comprised of twenty 0.5 mm brass sheets with aperture pattern of squares fabricated using chemical etching. The air gap 54 is another feature which may be incorporated. The placement of air gaps between adjacent subassemblies can increase material efficiency and reduce collimator weight while maintaining or increasing collimation performance. FIG. 6 also depicts an x-ray 57 angled relative to an axis 56 through the center of an aperture such that if its path were any more obtuse it would intersect the magnetically shielding spill control plates 53. Few to no x-rays would be absorbed by material inserted in the space of the air gap which isn't already absorbed by the exit plate. However, if the air gap were removed by exit plate 55 being placed in direct contact with magnetically shielding spill control plates 53, x-ray 57 would not be attenuated before leaving the collimator and may become spill. Placement of air gap 54 incurs little to no additional fabrication cost and adds no material weight but enhances spill reduction. Air gap dimension can be 0.5 mm or alternatively 1, 2, 5, 7, 10, 15, 20, 25, 30, 35, 40, 45, 50, 55, 60, 65 or 70 mm or any value between 0.5 and 70 mm or any range of values between 0.5 and 70 mm. Alternatively, air gap dimension can be 1, 2, 5, 7, 10, 15, 20, 25, 30, 35, 40, 45, 50, 55, 60, 65 or 70 percent of the thickness of the collimator or any percentage between 1 and 70 percent or any range of percentages between 1 and 70 percent. Air gaps may be inserted between any two functional subassemblies, between two sheets within subassemblies composed of a plurality of sheets, or within an otherwise solid layer subassembly, and may incur benefits such as those described above. FIG. 7 is a diagram illustrating an embodiment of the present invention comprising entrance plate 51 and exit plate 55 positioned on either end of a section of intermixed mu-metal sheets and lead sheets 61. Subassemblies comprised of a plurality of sheets may be interleaved with one another. In FIG. 7, this technique has been applied to a mu-metal magnetically shielding spill control subassembly and a lead leakage control subassembly such that these two subassemblies together form section 61. An aperture design consideration which may pertain to embodiments of the present invention will now be briefly discussed. Reference has been made to the radii or width of apertures being made “as large or larger than the desired x-ray beam radius in the plane of the subassembly.” FIG. 8 is a diagram illustrating the effect that focal spot blurring may have on the size of the desired x-ray beam radius in the plane of a subassembly. “Focal spot blurring” refers to the fact that focal spots in a scanning beam source may have some finite radius rather than existing as a single point on the transmissive target screen. Focal spot blurring may be necessary to avoid destroying the target screen by concentrating too much energy, and hence too much heat, in too small of an area. In the upper image of FIG. 8, beam width 73 in plane 79 is determined by x-ray 74a and x-ray 74b, which lie along the outer edge of a beam emanating from point focal spot 71 and covering the face of detector 76. However, if an x-ray beam emanating from blurred focal spot 72 is shaped to beam width 73 in plane 79, it will cover an area including the face of detector 76 and some area around it. It can be seen that x-rays 75a and 75b, which lie along the outer edge of such a beam, will become spill. Therefore, in the lower image of FIG. 8, corrected beam width 77 is drawn in plane 79. Corrected beam width 77 is determined x-rays 78a and 78b, which lie along the outer edge of a beam emanating from blurred focal spot 72 and covering the face of detector 76. It can be seen that corrected beam width 77 is smaller than beam width 73. For embodiments of the present invention, the determination of the desired x-ray beam radius in the plane of the subassembly may take into account the effects of focal spot blurring. To obtain a desired beam radius for the subassembly plane, one may calculate a width using a point focal spot model, e.g. calculate beam width 73, and then decrease this width by ten percent. The radius may also be approximated by decreasing the width from a point focal spot model by some other percent in light of prior source behavior or known focal spot size. The percentage decrease can be 5, 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18, 19, or 20 percent or any range of percentages between 5 and 20 percent. Apertures may then be sized accordingly. The foregoing descriptions of specific embodiments of the present invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise forms disclosed, and many modifications and variations are possible in light of the above teaching. The embodiments 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 and various embodiments 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 and their equivalents.
	description	Field of the Invention This invention relates to bone-seeking radioactive metal-chelant compositions that are suitable for administration to a Patient having: bone pain; one or more calcific tumors; or in need of a bone marrow suppression procedure. Description of Related Art Radiopharmaceuticals based on metal-chelant complexes have been used to diagnose and treat bone cancer. For example, Quadramet® (trademark of Lantheus Medical Imaging, Inc.) is a commercially available chelate formed between Sm-153 and ethylene-diaminetetramethylenephosphonic acid (EDTMP) that is currently indicated for the pain associated with bone metastases (U.S. Pat. No. 4,898,724). Typical dosages are 1 mCi of Sm-153 per kg body weight of the patient. Thus for a 70 kg patient the dosage would be 70 mCi. U.S. Pat. No. 5,059,412 teaches the use of Sm-153, Gd-159, Ho-166, Lu-177 and Yb-175 chelates with chelants derived from the 1,4,7,10-tetraazacyclododecane moiety including 1,4,7,10-tetraazacyclododecanetetramethylenephosphonic acid (DOTMP), while U.S. Pat. No. 5,064,633 teaches the above metals plus Y-90. Compositions of Sm, Gd, Ho, Lu and Y with DOTMP comprising predominately non-radioactive metal with the corresponding radioactive metal (e.g. Sm-152 with Sm-153 at μCi levels) were prepared and biodistribution data in rats was obtained. A therapeutically effective biodistribution (fate of the activity after administration) for a therapeutic bone agent includes high bone uptake, low soft tissue uptake, rapid clearance of the activity not associated with bone, and high lesion-to-normal bone ratio. Compositions that do not have these characteristics are detrimental to the patient. For example, high soft tissue uptake would result in the patient receiving a high radiation dose to the liver, bone marrow or other soft tissue leading to undesirable side effects. Radionuclides such as Sm-153 are prepared in a nuclear reactor by bombarding purified targets of the element containing one less neutron and in the process generate radionuclidic impurities. For example, to produce Sm-153 the target that is irradiated is Sm-152. When Sm-153 decays, Eu-153 is formed and an unwanted impurity, radioactive Eu-154 is formed from neutron capture by Eu-153. The impurities can be detrimental to institutions from both a patient and a waste disposal standpoint. For example, too much Eu-154 administered to a patient would result in the isotope giving an undesirable dose to a patient for a long period of time because of its half-life of 8.8 years. In addition, the dose that is excreted in the urine by the patient containing Eu-154 is a concern and institutions may be forced to collect the radioactive urine. Disposal of the product vials containing residual activity can be a problem. These vials and syringes are typically allowed to decay for 10 half-lives prior to disposal. This is a reasonable amount of time for Sm-153 (about 20 days) but not for Eu-154 (about 88 years). Processes must be implemented in order to deal with waste disposal of vials and syringes that are used. This makes the use of these types of radiopharmaceuticals more complex and institutions may chose not to use the drugs. In addition, these long-lived impurities cause issues with the radioactive licensing process for the institution. Typically institutions are only allowed small amounts of long-lived radionuclides (having half-lives greater than 120 days) before they are required to have financial assurance. Financial assurance can be very expensive especially for institutions that only handle short-lived isotopes. The specifications for Quadramet® call for the product to contain less than 0.093 microcuries (μCi) of Eu-154 per millicurie (mCi) of Sm-153 at Expiration Date (http://health.phys.iit.edu/extended_archive/0001/msg00922.html, http://acnp-cal .org/SM1531NS.html) or 4 days from the manufacture date (http://www.ibamolecular.eu/products/quadrainet). This restriction limits the expiration time of the drug. Since Sm-153 decays faster than Eu-154, the longer the Sm-153 solution decays, the higher the amount of Eu-154 in the sample relative to Sm-153. Thus expiration of not only formulated Quadramet® (e.g. Ca-EDTMP + Sm-153) but also the Sm-153 used to produce Quadramet® is limited by the amount of Eu-154 in the sample. In nuclear reactors such as the one at the University of Missouri in Columbia, Mo., the Sm-152 samples are irradiated for one week in the “flux trap” in order to produce the high specific activity Sm-153 required for the production of Quadramet®. The flux trap is only accessed once a week and therefore high specific activity Sm-153 can only be produced on a weekly basis. Because of the growing amount of Eu-154 compared to Sm-153, the isotope can only be used for a short period of time. Thus the drug is not available to treat patients on some days of the week. The flux trap portion of the reactor is also the most expensive to access (requiring reactor shut-down), thus increasing the production cost of the isotope. Clearly, there is a need for a product with a longer shelf life and a better impurity profile. The present invention provides a method for the treatment of a Patient comprising administration to the Patient having bone pain, one or more calcific tumors, or in need of a bone marrow suppressing procedure, a pharmaceutically-acceptable formulation of a chelate composition comprising a Clinically Relevant Dosage of the composition that is therapeutically effective, said composition possessing an extended Expiration Date of greater than or equal to about 5 days and said chelate comprises Sm-153 and DOTMP or a physiologically-acceptable salt thereof wherein the Sm-153 dosage is at least 35 mCi. The formulation of this invention comprises a chelate composition either as a pre-mixed drug ready for use or a kit having two separate components, the chelant and the isotope, which components are mixed to form the chelate composition at the appropriate time prior to use in the method. Also provided is the chelate composition comprising a Clinically Relevant Dosage of the composition that is therapeutically effective and pharmaceutically-acceptable, said composition possessing an extended Expiration Date of greater than or equal to about 5 days and said chelate comprises Sm-153 and DOTMP or a physiologically-acceptable salt thereof wherein the Sm-153 dosage is at least 35 mCi. It is understood that the terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting. As used in this specification, the singular forms “a”, “an”, and “the” include plural referents unless the content clearly indicates otherwise. The following terms in the Glossary as used in this application are to be defined as stated below and for these terms, the singular includes the plural. Various headings are present to aid the reader, but are not the exclusive location of all aspects of that referenced subject matter and are not to be construed as limiting the location of such discussion. Also, certain US patents and PCT published applications have been incorporated by reference. However, the text of such patents is only incorporated by reference to the extent that no conflict exists between such text and other statements set forth herein. In the event of such conflict, then any such conflicting text in such incorporated by reference US patent or PCT application is specifically not so incorporated in this patent. Glossary % means weight percent, unless stated otherwise Clinically Relevant Dosage means enough activity to cause either pain palliation or reduction of tumor burden. This dosage is about 0.5 mCi per kg body weight or about 35 mCi for a 70 kg patient; more preferred 1.0 mCi per kg body weight or about 70 mCi for a 70 kg patient. Higher amounts of radioactivity may be administered to the patients or for treating tumor regression or bone marrow ablation in patients. DOTMP means 1,4,7,10-tetraazacyclododecanetetramethylenephosphonic acid EDTMP means ethylenediaminetetramethylenephosphonic acid Expiration Date means the number of days after production when a Sm-153 bone agent formulation contains equal to or greater than 0.093 microcuries of Eu-154 per mCi of Sm-153. FDA means US Food and Drug Administration including its regulations Patient means an animal or human in need of treatment Ci means curies μCi means microcuries mCi means millicuriesDiscussion The specific activity of an isotope is sometimes a source of confusion because it is expressed in many ways (see Practical Aspects of labeling DTPA and DOTA peptides with Y-90, In-111, Lu-177 and Ga-68 for Peptide-Receptor Scintigraphy and peptide-Receptor Radionuclide Therapy in preclinical and Clinical Applications (http://pharmacyce.unm.edu/program_information/freelessonfiles/Vol16Lesson5.pdf). For this invention, the specific activity of an isotope is defined as the radioactivity of the isotope in question divided by the mass of all of the isotopes (stable and radioactive) of that element. For example for reactor produced Sm-153 where the starting material is Sm-152 that is converted to Sm-153, the specific activity of Sm-153 is the amount of radioactivity of Sm-153 in the sample divided by the total mass of Sm in the sample (e.g. activity Sm-153/sum of masses of Sm-152 and Sm-153). The units of the number are typically in Curies per gram (Ci/g) or milliCuries per milligram (mCi/mg). In some cases the percent of the isotope that is radioactive is reported. For example in reactor produced Sm-153, only about 2% of the Sm is Sm-153 and about 98% is non-radioactive Sm-152. Traditionally, nuclear medicine scientists strive to increase the specific activity of the isotopes of interest. For example, two government grants for providing high specific activity isotopes have been recently granted (High Specific Activity Sm-153 by Post Irradiation Isotope Separation, DOE SBIR grant Solicitation Number DE-FOA-0000676, Production of Commercial High Specific Activity Sn-117m Radiochemical and Chelates, DOE grant Solicitation Number DE-FOA-000782). The use of high specific activity isotopes allows for less mass of the element needed to achieve the same amount of radioactivity. This leads to lower amounts of chelating agents and/or proteins needed in the radioactive drug. In addition, in many cases such as with labeled antibodies and proteins, the receptors on cells (such as cancer cells) that the drugs target are limited. If the specific activity of the isotope is low (e.g. 2% of the atoms are radioactive), then the amount of active drug that reaches the target is relatively small. However, if the specific activity is high (e.g. 100% of the atoms are radioactive), then the amount of effective drug that reaches the target is much higher, which explains why so much effort is put forth in radioisotope production to achieve higher and higher specific activity. Contrary to this conventional wisdom where higher specific activity isotopes are sought-after as desirable, this invention utilizes Sm-153 produced in a lower flux portion of the nuclear reactor for a shorter period of time, resulting in a lower specific activity isotope with a significant cost reduction and lower impurity profile. When combined with DOTMP a product can be produced which comprises a Clinically Relevant Dosage of Sm-153-DOTMP with a reduced radionuclidic impurity profile, a longer shelf life, a lower cost to manufacture, and can be made available to patients on a more frequent basis. The formulations of the present invention may be in a kit form such that the two components (chelant and isotope) are mixed at the appropriate time prior to use or provided pre-mixed as the ready to use drug. Whether pre-mixed as the drug or as a kit where the drug is made on site, the formulations require a pharmaceutically-acceptable carrier. Such carriers comprise any suitable pharmaceutically-acceptable carrier such as one or more of a suitable solvent, preservatives, diluents, excipients and buffers. Useful solvents include, for example, water, aqueous alcohols and glycols. The formulation is administered to the Patient by injection intravenously or intramuscularly. The invention will be further clarified by a consideration of the following examples, which are intended to be purely exemplary of the invention. Materials and Equipment The radioactive isotopes were purchased from The University of Missouri Research Reactor. Chelants were purchased from commercial sources or were prepared as described in U.S. Pat. No. 5,059,412. General Procedure In the following examples, the lettered examples are comparative, and the numbered examples are this invention. Sm-153 is prepared by irradiating Sm-152 in a nuclear reactor for 48 hours with a thermal neutron flux of 8.00×1013 neutrons/cm2-sec. The specific activity of the Sm-153 is 1,430 mCi/mg at end of irradiation and contains 0.0005 μCi of Eu-154 per mCi of Sm-153. After 5 days of radioactive decay, the specific activity of the Sm-153 is 237 mCi/mg and the the activity of the Eu-154 impurity is 0.00296 μCi of Eu-154 per mCi of Sm-153. This is below the FDA allowable amount of Eu-154 (0.093 μCi of Eu-154 per mCi of Sm-153). However, the Quadramet® formulation requires a minimum EDTMP to Sm mole ratio of 273:1 in order to properly control the biodistribution of Sm-153 (calculated from the data in Quadramet® package insert). Because of this requirement the maximum mass of Sm used in a Quadramet preparation is about 0.134 mg. Therefore at a specific activity of 237 mCi/mg only about 32 mCi can be prepared. This is not a sufficient dosage to treat a 70 kg patient at 1 mCi/kg or even a 70 kg patient at 0.5 mCi/kg. Sm-153 is prepared by irradiating Sm-152 in a nuclear reactor for 48 hours with a thermal neutron flux of 8.00×1013 neutrons/cm2-sec. The specific activity of the Sm-153 is 1,430 mCi/mg at end of irradiation and contains 0.0005 μCi of Eu-154 per mCi of Sm-153. After 5 days the specific activity is 0.237 Ci/mg and the activity of the impurity Eu-154 is 0.00296 μCi Eu-154 per mCi of Sm-153. This is 3.2% of the FDA allowable maximum amount of Eu-154. Since Sm-153-DOTMP can be prepared using a 1:1 mole ratio of DOTMP to Sm, a preparation using 10 mg of DOTMP and 657 mCi (2.77 mg Sm) of the 5 day old Sm-153 is made. This composition produces sufficient quantities of Sm-153 to treat 9 patients weighing an average of 70 kg at 1 mCi/kg and has an Expiration Date of greater than 5 days. A formulation, prepared as in Example 1, is allowed to decay 10 days. The amount of Sm-153 in the formulation is now 110 mCi which is sufficient to treat one patient with a weight of 70 kg at 1 mCi of Sm-153 per kg body weight. The formulation contains 0.0178 μCi of Eu-154 per mCi of Sm-153 which is 19% of the allowable amount of Eu-154. Therefore, the composition has a greater than 10 day Expiration Date. A formulation, prepared as in Example 1, is allowed to decay 13 days. The amount of Sm-153 in the formulation is now 37 mCi which is sufficient to treat one patient with a weight of 70 kg at 0.5 mCi of Sm-153 per kg body weight. The formulation contains 0.05228 μCi of Eu-154 per mCi of Sm-153 which is 56% of the allowable amount of Eu-154. Therefore, the formulation has a greater than 13 day Expiration Date. Although the invention has been described with reference to its preferred embodiments, those of ordinary skill in the art may, upon reading and understanding this disclosure, appreciate changes and modifications which may be made which do not depart from the scope and spirit of the invention as described above or claimed hereafter. Accordingly, this description is to be construed as illustrative only and is for the purpose of teaching those skilled in the art the general manner of carrying out the invention.
	description	This application is based on and claims priority under 35 U.S.C. § 119(e) to U.S. Provisional Patent Application No. 62/688,255, filed Jun. 21, 2018, the entire contents of which are incorporated herein by reference. The present disclosure relates generally to fission reactors and structures related to the active reactor space in fission reactors. In particular, the disclosed fission reactor and reactor space includes fissionable nuclear fuel loaded into the spaces between channels for coolant flow, and is scalable in size while each location with fissionable nuclear fuel remains identical in cross-sectional area and/or volume, regardless of reactor size. Support and ancillary equipment, such as control rods, control rod drivers, and moderators and also scalable in size. The present disclosure also relates to methods to manufacture such reactors and structures, particularly by additive manufacturing techniques producing an integral and unitary structure for the fuel loaded reactor space, and provides predictive quality assurance for the manufacture of such reactors and structures. In the discussion that follows, reference is made to certain structures and/or methods. However, the following references should not be construed as an admission that these structures and/or methods constitute prior art. Applicant expressly reserves the right to demonstrate that such structures and/or methods do not qualify as prior art against the present invention. Traditional fission reactors utilize fissionable nuclear fuel, such as uranium-based fuel, placed inside fuel elements, which can be round tubes, plates, or hexagon-shaped. These fuel elements are collected and arranged into fuel assemblies, which are the base element of the core of a nuclear reactor. Conventional fuel assemblies 10 (see FIG. 1) are complex arrangements of, for example, fuel elements 12 (which contain the fuel 14 and burnable poisons), mechanical support for the fuel assembly structure, spacer grids 16 (which ensure a spacing of components and guiding of the fuel elements), and non-fuel tubes for, e.g., control rods 18 or in-core instrumentation 20 and the like. Depending on the design, a reactor vessel may have dozens of fuel assemblies 10 (also known as fuel bundles), each of which may contain 200 or more fuel elements 12. Within the core, primary coolant (such as water) flows through and/or around the fuel assemblies 10 and provides both a moderator for the fission reaction (in the case of water-type cooled reactors) and a heat extraction medium for heat generated by fission reaction in the fuel elements. The heated primary coolant circulates within a primary cycle (meaning those systems subject to, in contact with or otherwise exposed to the primary coolant) and traditionally transfers thermal energy to a secondary system, where thermally excited fluid is generated and flows to turbines which, in turn, can be used to spin an electric generator. Complexity of structure extends to other systems in a nuclear reactor, including the various components of the primary cycle, such as, depending on design, tubing, pumps, instrumentation, heat exchangers, and steam generators. Accordingly, construction of fuel elements, fuel assemblies, reactor cores, and reactor systems are all subject to rigorous design and manufacturing standards as well as extensive pre-, during, and post-manufacturing controls, such as those related to sourcing, handling, installing, inspecting and testing. Thus, it would be advantageous to have a design of these complex structures, in particular the fuel element and fuel assembly, that improves any of the design and manufacture of such complex structures and quality assurance. In general, the disclosure is directed to a fission reactor that places fissionable nuclear fuel in locations in the reactor core between and around non-fuel tubes for primary coolant, moderator, control rods, scram rods and/or ancillary equipment. This placement of fissionable nuclear fuel and non-fuel tubes is opposite to (or inverted from) the conventional arrangement of fissionable fuel located in tubes and primary coolant flowing between and around the fuel tubes. Embodiments disclosed herein include a fission reactor comprising a shell encompassing a reactor space having a longitudinal axis and an axial cylinder including an inner diameter surface defining a central longitudinal channel having an axis that is co-located with the longitudinal axis of the reactor space. A plurality of axially extending rings is located within the reactor space and concentrically positioned relative to the axial cylinder. The plurality of axially extending rings is radially separated forming, for any two adjacent axially extending rings, both a radially inward adjacent ring and a radially outward adjacent ring. An outer diameter surface of the radially inward adjacent ring and an inner diameter surface of the radially outward adjacent ring define an annular cylindrical space. The fission reactor includes a first plurality of primary axial tubes is located circumferential within each annular cylindrical space. Each primary axial tube includes an inner diameter surface forming a primary channel and an outer diameter surface. A plurality of webbings connects at least a portion of, alternatively all of, the plurality of primary axial tubes to adjacent structure, such as the outer diameter surface of each of the plurality of primary axial tubes connected to the radially inward adjacent ring by a first webbing and connected to the radially outward adjacent ring by a second webbing. The fission reactor includes a plurality of secondary channels within each annular cylindrical space, wherein circumferentially adjacent primary axial tubes are separated by one of the plurality of secondary channels. A fissionable nuclear fuel composition is located in at least some of the plurality of secondary channels. Embodiments disclosed herein also include a method of manufacturing a fission reactor. Embodiments of the method apply predictive and causal analytics to prepare a model of the fission reactor, fabricate the fission reactor on a layer-by-layer basis using additive manufacturing techniques, during fabricating, in-situ monitor fabrication of the fission reactor with machine vision and accelerated data-processing, analyzes data from the in-situ monitoring, and adjusts the fabricating of the fission reactor based on the real-time analyzed data. In some instances, manufacturing equipment, particularly additive manufacturing equipment, have a limited manufacturing volume that impacts the maximum size of any single monolithically manufactured piece (although repositioning technology may accommodate increases in the size of such monolithic manufactured pieces). According, the fabrication methods disclosed herein for, e.g., the fission reactor (or other structures), can be adapted to manufacture structures on a monolithic basis or on a segmented basis for subsequent assembly. Embodiments of the method can also prepare a digital version of the fabricated fission reactor; and correlate a characteristic of the fabricated fission reactor based on an analysis of the digital version of the fabricated fission reactor. Further, embodiments disclosed herein can be used to qualify designs and validate acceptable fabrication of fission reactors as well as individual components of fission reactors. For example, methods of manufacturing fission reactors disclosed herein can also be used to determine and confirm the performance and integrity of the as-built structures. As such, the methods can serve as a new means to qualify a reactor with, or provide information for acceptance criteria by, third parties, for example, government regulatory agencies, government agencies and departments, commercial entities such as power companies, and the like. Although the disclosed reactor and core have complex mechanical geometries, integral and iterative manufacturing, such as 3D printing, of elemental metal or metal alloys or ceramics (including using such materials in, e.g., particle, wire or powder form), enables the inverted reactor to be more easily manufactured. Other advantages include an improved power to weight ratio, a reduction of internal stresses, and expandability by, for example, adding additional dimensional units in the form of rings or ring spacing. FIG. 2A shows a perspective, axial cross-sectional view of an example fission reactor. The fission reactor 100 comprises a shell 102 that contains a fissionable nuclear fuel composition (an example of which is shown as fissionable nuclear fuel composition 104 in FIG. 2B), control rods and ancillary equipment 106 which movably penetrate the shell 102 and a reactor space 108, a reflector 110 around an outer diameter surface of the shell 102, tubes for primary coolant flow 112 to and from the shell 102, and a containment housing 114. FIG. 2B shows a magnified, perspective, radial cross-sectional and axial cut-away view of some of the features of FIG. 2A. For purposes of illustration and clarity, other features of fission reactors and of fission power plants are not shown in FIGS. 2A-B, such as other features of the primary system and secondary system, but are known to those of ordinary skill. FIG. 3A shows a perspective, radial cross-sectional view of a portion of a fission reactor 100. The illustrated shell 102 has a longitudinal axis 120 extending from a first end to a second end of the reactor space 108. Shell 102 encompasses a reactor space 108 that, in this embodiment, has internal features similar to a honeycomb structure, both radially and axially. For example, within shell 102 there is an axial cylinder 130 including an inner diameter surface 132 defining a central longitudinal channel having an axis 134 that is co-located with the longitudinal axis 120 of the reactor space 108. Also located within the reactor space 108 are a plurality of axially extending rings 140 that are concentrically positioned relative to the axial cylinder 130. With reference to FIG. 3B, which shows a magnified, perspective, radial cross-sectional view of a portion of FIG. 3A, at least some of the plurality of axially extending rings 140 are radially separated and, when any two axially extending rings 140 are considered, form a radially inward adjacent ring 140a and a radially outward adjacent ring 140b. An outer diameter surface 142 of the radially inward adjacent ring 140a and an inner diameter surface 144 of the radially outward adjacent ring 140b define a cylindrical space 150. Located circumferential within annular cylindrical space 150, there are a plurality of primary axial tubes 160. Each primary axial tube 160 includes an inner diameter surface 162 forming a primary channel 164 (primarily used for flow) and an outer diameter surface 166. A plurality of webbings 170 connect the outer diameter surface 166 of each of the plurality of primary axial tubes 160 to, in a first instance, the radially inward adjacent ring 140a and, in a second instance, the radially outward adjacent ring 140b. In some embodiments, the axial tubes 160 are connected by webbings 170 to at least one of, alternatively to both, the radially inward adjacent ring 140a and the radially outward adjacent ring 140b; in other embodiments, only some of the axial tubes 160 are connected by webbings 170 to at least one of, alternatively to both, the radially inward adjacent ring 140a and the radially outward adjacent ring 140b. The number, location and frequency of use of webbings 170 can vary based on the dimensional integrity to be provided to the overall design by making the connections using the webbings 170. The inner diameter surface of the primary axial tubes (i.e., primary flow channel) can be uniform as a function of axial position or can vary. For example, in some embodiments, the inner diameter surface of the primary axial tubes forming the primary channel can vary as a function of axial position relative to the longitudinal axis of the primary axial tube, for example, to influence the flow properties of primary coolant. Also for example, in other embodiments, the primary channel is chambered to form different areas or zones along the axial length. These zones can be used to house instruments and/or other equipment or materials to monitor or influence reactor performance. In some embodiments, one or more of the central longitudinal channel of the axial cylinder 130 and the primary channels 164 are accessible from an outer surface of the fission reactor. When accessible, the central longitudinal channel and/or the primary channel(s) can be used to prepare irradiated samples, such as irradiated medical equipment, medical isotopes, scientific isotopes, and so forth. Also located within the reactor space 108 are a plurality of secondary channels 180. With reference to FIG. 3B, the plurality of secondary channels 180 are located within annular cylindrical space 150 and separate circumferentially adjacent primary axial tubes 160a, 160b. For example, inner surfaces of the secondary channel 180 include portions of the outer diameter surface 166 of the circumferentially adjacent primary axial tubes 160a, 160b, surfaces of a first webbing 170 and a second webbing 170 associated with each of the circumferentially adjacent primary axial tubes 160a, 160b, and portions of the outer diameter surface 142 of the radially inward adjacent ring 140a and portions of the inner diameter surface 144 of the radially outward adjacent ring 140b. Typically, circumferentially adjacent primary axial tubes 160a, 160b are non-contactingly distributed within the cylindrical space 150, forming a secondary channel 180. Also located within the reactor space 108 is a fissionable nuclear fuel composition 190. For example and as seen schematically in FIG. 4, the fissionable nuclear fuel composition 190 can be located in at least some of the plurality of secondary channels 180. The fissionable nuclear fuel composition 190 is in thermal transfer contact with at least some, if not all, of the inner surfaces of the secondary channel 180. During operation, a primary coolant is flowable through the primary channel 164 of each of the circumferentially adjacent primary axial tubes 160 that are separated by one of the plurality of secondary channels 180 which contain the fissionable nuclear fuel composition 190 to effect the thermal transfer. In the illustrated embodiment, a cross-section of the secondary channel perpendicular to the longitudinal axis has a shape of a cross-section of a hyperboloid of one sheet, however, other cross-sectional shapes can be used. A suitable fissionable nuclear fuel composition includes uranium oxide and is less than 20% enriched, uranium with 10 wt. % molybdenum (U-10Mo), uranium nitride (UN), and other stable fissionable fuel compounds, including metal-based fissionable fuels and ceramic-based fissionable fuels. As is known in the art, during fission reaction of fissionable nuclear fuel, the breakdown of uranium produces many alternative elements in different phases (gas, liquid, or solids). Due to the design of the secondary channels 180 that contain the fissionable nuclear fuel composition 190 disclosed herein, the increase in internal pressure in the secondary channel 180 due to this transmutation of elements places the secondary channels 180, i.e., the fuel chambers, in compressive forces and improves resistance to failure. This phenomenon is also seen when thermal expansion occurs. In contrast, traditional nuclear reactor fuels with uranium located inside tubes usually made typically of zirconium, the transmutation of elements increases internal tube pressures placing hoop stress (a form of tensile hoop stress) on the tubing that can lead to structural failure, such as cracking. Also, materials subject to tensile stresses are susceptible to different types of corrosion mechanisms, such as stress corrosion cracking, as compared to materials subject to compressive stresses. Further, hydride-forming metals (such as zirconium) are subject to hydrogen embrittlement and can become brittle and fracture, which is exasperated when the relevant part is under tensile stress as compared to compressive stress. It should be noted that in the illustrated embodiment in FIG. 3A, an innermost plurality of primary axial tubes is not separated from the axial cylinder 130 by an axially extending ring 140. Thus, the reactor 100 includes a plurality of primary axial tubes 160 located circumferential between an inner diameter surface of the most radially inward, axially extending ring 140 and an outer diameter surface of the axial cylinder 130. Similar to that described in connection with FIG. 3B, the outer diameter surface of each of these plurality of primary axial tubes is connected to the outer diameter surface of the axial cylinder 130 by a first webbing 170 and is connected to the most radially inward, axially extending ring 140 by a second webbing 170. It should also be noted that in the illustrated embodiment in FIG. 3A, an outermost plurality of primary axial tubes is not separated from the shell 102 by an axially extending ring 140. Thus, the reactor 100 includes a plurality of primary axial tubes 160 located circumferential between an inner diameter surface of the shell 102 and an outer diameter surface of the most radially outward, axially extending ring 140. Similar to that described in connection with FIG. 3B, the outer diameter surface of each of these plurality of primary axial tubes is connected to the outer diameter surface of the most radially outward, axially extending ring 140 by a first webbing and is connected to the inner diameter surface of the shell 102 by a second webbing. Various supporting and ancillary equipment can be located in one or more primary channels 164. For example, at least one of a moderator, a control rod, and a scientific instrument, such as a temperature sensor or radiation detector, can be located in one or more primary channels. FIG. 5A is a schematic illustration of a plurality of primary channels 164 in which supporting and ancillary equipment in the form of a control rod 200, such as iridium control rod, and a moderator 210, such as a zirconium hydride neutron moderator, are located. The control rod 200 can also incorporate a neutron poison which absorbs neutrons and can be used to regulate the criticality of nuclear reactors. Additionally, the poison material can absorb enough neutrons to shut down the fission reactor 100 (e.g., when the control rods 200 are completely inserted into the reactor space 108) or can be axially positioned to maintain criticality of the fission reactor 100 (e.g., when the control rods 200 are withdrawn from the reactor core 109 a distance to allow a continuous fission chain reaction). In some embodiments, the moderator 210 is cooled by flowing He and stabilized with a tri-fin design. Any suitable number of control rods 200 and moderators 210 can be used and suitably distributed throughout the reactor space 108 in order to obtain one or more of a desired flux profile, power distribution, and operating profile. In exemplary embodiments, control rods 200 are threaded, which contributes to save axial space, maximizes control rod diameter, and allows for direct roller nut contact for reliable SCRAM operation. All or a subset of control rods 200 can be individually controlled by independent motors to provide discrete reactivity control and/or for power shaping. In some embodiments, an insert of neutron moderating material in the form of a rod with one or more axial protrusions can be located in the primary channel 164. FIGS. 5A and 5B illustrate an example of such a neutron moderating material in the form of a rod 210 in primary channel 164. The rod 210 includes one or more fins 212 or other protrusion(s) that contribute to maintaining a consistent gap 214 between the inner diameter surface of the primary channel 164 and the outer surface (or at least a majority of the outer surface) of the moderator rod 210. The fins/protrusions 212 can extend axially along the length of the rod 210. This design is particular relevant to gas-cooled reactors in which the gap 214 allows for sufficient flow for the gas to, for example, both generate thrust for a space reactor or drive a closed-loop power generating system, and cool the moderating material. The moderator material also acts to thermalize the neutrons, creating a more neurotically efficient core. Individual moderator rods can advantageously be inserted into any number of desire locations in the core, can be independently replaced or serviced as needed, and allows for a larger diameter coolant hole during manufacture of the core. The moderator rod 210 may also take the form of an annulus to allow for additional cooling or accommodate the insertion of a control rod 200 or other material, such as is shown in the alternative embodiment of a moderator rod 210 illustrated in FIG. 5C. A cladding material may also be used for hydrided materials to limit or prevent migration of hydrogen, the key component for neutron moderation, from the metal. Cladding can also be used as a barrier between the moderator material and coolant gas. As noted in discussing FIG. 2A, the fission reactor 100 includes control rods which movably penetrate the shell 102 and the reactor space 108. The positioning and operation of the control rods, such as control rod 200, is controlled by a control rod system 220 (See FIG. 5D). An embodiment of a control rod system 220 includes three major items: a control rod drive motor 230 used to move the control rod 200 into and out of the reactor space 108; a threaded drive shaft 240 connected to the control rod 200 that drives the control rod 200 into and out of the reactor space 108; and the control rod 200, which is usually a cylindrical neutron absorbing poison that travels into and out of the primary channel 164. Driving the control rod 200 into and out of the reactor space 108 is usually performed by rotating a threaded nut located internally to the control rod drive motor and coupled to the threaded drive shaft such that rotation of the internally threaded nut causes translational movement, i.e., in the longitudinal direction, of the control rod 200. In some applications, such as space reactors, the size and weight for the fission reactor and its components is limited to the weight/cost penalty incurred when such systems are launched into space. Thus, other embodiments of a control rod system seek to simplify their design since maintenance or replacement of reactor components cannot be performed once launched or once they have operated. Thus, it is beneficial to reduce the size, weight and complexity of the items in the control rod system. Although not necessarily size and weight limited, terrestrial reactors can benefit from similar improvements in design because of the reduction in maintenance and reduced part replacement. To address such design concerns, embodiments of the control rod system can combine the threaded drive shaft and the control rod poison by manufacturing the threaded drive shaft itself out of neutron absorbing material. When the threaded drive shaft is manufactured using neutron absorbing material, separate control rod poison can be reduced or eliminated from the fission reactor. Control rod 200 in FIG. 5A shows an exemplary embodiment of such a threaded control rod that is manufactured from, or otherwise incorporates into its structure, neutron absorbing material. Suitable materials that can used in manufacturing the control rod (or otherwise incorporated in its structure) include: iridium, hafnium, stainless steel, tungsten, boron carbide in an aluminum oxide matrix (Al2O3—B4C), molybdenum, and tantalum. While any one or more of various high temperature metallic neutron absorbing materials can be used, it is currently contemplated that iridium would be used as the neutron absorbing material. FIG. 6 illustrates in perspective cross-sectional view an example number and distribution of control rods 200 and moderators 210 in an exemplary embodiment of a fission reactor. As discussed above, transmutation of elements increases internal pressures associated with the space occupied by the fissionable nuclear fuel. To reduce such internal pressures, embodiments of the fission reactor can design flexibility into components of the fission reactor to reduce the stresses that develop. For example, instead of a continuous volume of a fuel, embodiments of the disclosed fission reactor can incorporate a space, gap, hole or other opening between sections of the fissionable nuclear fuel composition within the secondary channels 180 or within the fissionable nuclear fuel composition 190 itself. One example of such a space or gap is illustrated in FIG. 7, in which one or more gaps 250 is incorporated into the fissionable nuclear fuel composition 190. Example locations for the gap(s) 250 include between the fissionable nuclear fuel composition 190 and the webbings 170 (see area 252) and in the body of the fissionable nuclear fuel composition 190 (see area 254). Modeling of stress in designs incorporating gaps showed a lower stress in the active reactor space 108 in areas having the gaps relative to areas which had no gap. Additionally, the overall hoop stress in the shell 102 was decreased. Further, the fissionable nuclear fuel composition 190 showed better interfacing with the surfaces and structures forming the secondary channel 180, resulting in better thermal transfer performance. For example, incorporating gaps 250 into the design has been shown to improve the contact with the outer diameter surface 166 of each of the plurality of primary axial tubes 160 (as compared to a design without such a gap 250), which contributes to improved thermal transfer between the fissionable nuclear fuel composition 190 and the primary coolant flowing through primary channel 164 formed by the inner diameter surface 162 of the primary axial tube 160. In some embodiments, the fission rector 100 is a core of a gas-cooled nuclear reactor, in which thermal transfer occurs via gas flowing through holes in a reactor space 108, such as the primary channels 164 shown in FIGS. 3A-B, 4, and 5A, which are sized to allow efficient heat transfer from the solid reactor core. In embodiments of gas-cooled nuclear reactor, primary coolant removes heat from the reactor core, which in turn, heats the gas. The heated gas can then be used for thrust as in nuclear thermal rockets or used to drive a closed-loop power generating system. In order to generate the heat in the fissionable nuclear fuel that is transferred to the primary coolant, nuclear reactors rely on neutron moderating materials to thermalize, or slow, neutrons released in the fission process. Moderation of neutrons is required to sustain the nuclear chain reaction in the core and thus the production of heat. Water-cooled reactors rely on the water to both cool and moderate the neutron population; however, gas-cooled reactors require an additional material for moderation. The use of the additional moderating material to thermalize neutrons can allow reduction in the amount of fuel, and thus weight of the fission reactor, because thermalized neutrons more efficiently split fissile atoms. In some embodiments, features of the fission reactor including at least the shell, the axial cylinder, the plurality of axially extending rings, the plurality of primary axial tubes, and the plurality of webbings are an integral, unitary structure. In other words, these features of the fission reactor are formed integrally by, for example, an additive manufacturing process. An example of a suitable additive manufacturing process utilizes 3-D printing of a metal alloy, such as a molybdenum-containing metal alloy, Zircalloy-4 or Hastelloy X, to form the noted structural features. In other embodiments, the fissionable nuclear fuel composition can be included within the integral, unitary structure when suitable multi-material, additive manufacturing processes with multiple metals within the feedstock are employed. Other alloys that can be used when suitable multi-material, additive manufacturing processes with multiple metals within the feedstock are employed include: steel alloys, zirconium alloys, and Molybdenum-Tungsten alloys (for the shell); beryllium alloys (for the reflector); and stainless steel (for the containment housing). Powder feedstock can also be utilized. The reactors shown and described herein have a six-fold rotational symmetry relative to the longitudinal axis of the reactor space. For example and with reference to FIG. 8, one can see that similar features within the reactor space 108 are arranged in a six-fold rotational symmetry relative to the longitudinal axis 120 of the reactor space. Examples of this six-fold rotational symmetry are shown in FIG. 8 superimposed on the radial cross-sectional view of an exemplary embodiment of a fission reactor. For example, a first six-fold rotational symmetry 300 is illustrated between control rods 200; a second six-fold rotational symmetry 310 is illustrated between moderators 210, and a third six-fold rotational symmetry 320 is illustrated between the plurality of primary axial tubes 160, in the corresponding cylindrical space 150 for such features. It should be noted that the reactor space 108 (and, by extension, the reactor 100) is scalable by the addition or subtraction of one or more axially extending rings 140 and associated features disclosed herein such as primary axial tubes 160, as long as the underlying six-fold rotational symmetry relative to the longitudinal axis of the reactor space is maintained. For example, the radial configuration should geometrically progress as a factor of 6, e.g., 1, 6, 12, 18, 24, 30, 36 rods, etc. . . . . This allows each secondary channel, which contains the fissionable nuclear fuel, to have the same volume regardless of position in the reactor space 108 and promotes uniform and optimal heat transfer between the fissionable nuclear fuel, the material of the reactor space, and the primary coolant. Thus, for example, the fissionable nuclear fuel composition located in at least some of the plurality of secondary channels form a set of fissionable nuclear fuel elements that are volumetrically identical throughout the fission reactor. Also for example, a ratio of an area of a radial cross-section of the primary channels to an area of a radial cross-section of the secondary channels is constant throughout the fission reactor (as considered between one or more primary channels and one or more secondary channels). Additive manufacturing techniques, such as 3D printing techniques utilizing multiple feedstocks, can be used to produce the integral and unitary structure for the fission reactors and fuel loaded reactor spaces disclosed herein. For example, additive manufacturing technology creates complex geometries and, when coupled with in-situ sensors, machine vision imagery, and artificial intelligence, allows for tuning of the manufacturing quality as the components are built on a layer-by-layer additive basis (often, these layers are on the scale of 50 microns) and provides predictive quality assurance for the manufacture of such reactors and structures. Additive manufacturing techniques for the manufacture of the integral and unitary structure for the fission reactors and fuel loaded reactor spaces disclosed herein include the steps of: (a) predictive and causal analytics, (b) in-situ monitoring combined with machine vision and accelerated processing during the layer-by-layer fabrication of the structure, (c) automated analysis combined with a machine learning component, and (d) virtual inspection of a digital representation of the as-built structure (also referred to herein as a “digital twin”). FIG. 9 summarizes the additive manufacturing method 400 for manufacturing the integral and unitary structure for the fission reactors and fuel loaded reactor spaces disclosed herein. The method 400 includes predictive and causal analytics 410 in which existing and experimental data is used to determine initial Critical-to-Quality (CTQ) factors and provide training of an initial machine learning algorithm. The initial input data for the machine learning algorithm can be one or more of developed organically, provided by a third party, or based on historical data sets (such as open source and/or based on operations and experiments that have captured prior experience with additive manufacturing techniques logged as potential features and observations relevant to the current additive manufacturing process). In each case, the initial machine learning algorithm is an algorithmic representation of each step in the manufacturing process as well as an algorithmic representation of the ideal, final structure. Additional complexity can be added to the initial machine learning algorithm by, for example, including additional variables for inputs, outputs, manufacturing conditions such as environmental conditions, quality of supplies, etc. . . . . While algorithms applied to the noted initial input data alone help describe final critical-to-quality (CTQ) factors on reactor products, they are not expected to be sufficient for qualification of the as manufactured product. Data science methodology applicable to the step of predictive and causal analytics 410 including the following: (1) define defects; (2) translate to measured outputs (layer fusion, shape, position, etc.); (3) clean dataset with “tidy data” principles (variables in columns, observations in rows, linked tables, test reproducibility); (4) split data into training, test, and validation sets; (5) characterize data set, exploratory analysis, interrogate against physical theory; (6) extract candidate features; (7) state hypotheses of relationships to test from existing data; (8) build multivariate regression algorithms using resampling techniques for randomization; (9) assess in-and-out of sample errors; (10) evaluate hypotheses and establish foundational parameters for physical tests; (11) create known additive manufacturing geometries, validate predictive models, generate causal relationships and output parameters; and (12) re-evaluate hypotheses and update foundational defect definitions for machine learning basis. Successful initial machine learning algorithms determine foundational hypotheses about possible critical factors to the additive build with existing data prior to physical tests, and form the basis of machine learning. The final predictive model is then used to inform the in-situ measurement plan for physical production and initial machine learning conditions. The step of predictive and causal analytics 410 typically occurs prior to the layer-by-layer deposition of material to fabricate the structure of the manufactured object. The method 400 includes in-situ monitoring combined with machine vision and accelerated processing during the layer-by-layer fabrication of the structure 420. In this step, suitable in-situ monitoring captures data related to the layer-by-layer fabrication of the structure and accelerated processing digitizes the data for input into machine learning for analysis. In-situ monitoring can be by any suitable means. For example, industrial machine vision cameras can provide visual information including position information, thermocouples can provide temperature information of both supplied and as-deposited material, current and voltage sensors can provide information on deposition conditions, speed and rate of deposition can be monitored, environment conditions can be monitored, x-ray techniques can monitor material characteristics as well as provide materials characterization, infrared thermography for temperature distribution, and weld pool characteristics including structure and stress state are just examples of in-situ monitoring that can be conducted and the results then utilized in the additive manufacturing method. Other parameters that can be included in in-situ monitoring include cool-down profile, detection of voids, porosity measurement, defect detection such as for cracks, lamination, and dimensional irregularity. It should be noted that additive manufacturing methods pose sensor challenges as, for example, cameras and other sensors must be placed to collect data between the film layer and deposition head to detect structural placement and alignment. Parallel processing, such as GPU acceleration, may be advantageous—if not necessary—to handle the large amounts of data from in-situ monitoring and process dozens of desired features to return real-time corrections. Outputs of accelerated processing are fed back into machine control in a loop for self-correction or identification for off-line analysis with an as-built model. The repetition of the in-situ monitoring and automated analysis step on a layer-by-layer basis as the integral and unitary structure for the fission reactors and fuel loaded reactor spaces disclosed herein are manufactured allows for continuous feedback to the manufacturing process. That feedback is then a basis for (i) layer-by-layer adjustments in the additive manufacturing process, (ii) archiving of monitored and analyzed information in the digital twin, allowing for subsequent analysis and evaluation, and (iii) updating and adjustment of the manufacturing protocols and layer-by-layer instructions for use in a future additive manufacture of the integral and unitary structure for the fission reactors and fuel loaded reactor spaces disclosed herein. During layer-by-layer fabrication the structure of the manufactured object, the method 400 includes automated analysis combined with a machine learning component 430. Machine learning creates intelligence from input of machine vision and in-situ monitoring, applies that input to previously existing data, and updates the processing via machine training as well as self-adjusts and runs predictive qualification analysis during the additive manufacturing process. Machine learning can include anomaly detection algorithms to monitor process function. For example, anomaly detection algorithms can check for variations in deposition speed, unexpected latency or volume consumption, temperature, alignment or chemistry. An automatic analysis of this process data can choose one or more features X1 that indicate an anomaly, fit parameters U1 to characterize the distributions of each chosen feature, and compute the probability that an observed X fits within acceptable Gaussian error for each feature U. Machine learning can also include classifiers of imagery for in-process anomaly detection. For example, pixelated imagery can be used as input data, in which each sample is a small pixel area of an image. The number of areas that fit in an image represents the number of dimensions that can be used to differentiate and classify anomalies or shapes. The vectorized image of hundreds of dimensions (parts of images) permit the machine to learn what a proper shape or anomaly looks like via an optimization function on each feature. Once the classifier is trained to detect an anomaly, it can be further trained to identify prior detected deposition conditions that lead to an anomaly (within a certain statistical confidence level) and then apply that information to in-situ conditions to anomalies before they actually happen and proactively intervene in the additive manufacturing process to avoid the anomaly. This iterative corrective ability is important to make in-process adjustments before depositing numerous layers with defects. Neural networks can be utilized for non-linear hypotheses related to machine learning. For example, neural networks use an array of features that take input conditions and fit models to correctly predict output conditions using a hidden layer to develop weighting parameters. Forward propagation algorithms provide predictive capability, and backward propagation is used to uncover the weighting scheme learned by the system. The hidden layer makes it possible to arrive at workable solutions when the number of features is large (like the imagery data) or interactions are complex, or both. The method 400 also includes virtual inspection of a digital representation of the as-built structure (also referred to herein as a “digital twin”) 440. Typically, the step of virtual inspection of the digital twin 440 occurs after completion of the layer-by-layer deposition of material to fabricate the structure of the manufactured object. The digital twin can be analyzed using various computer assisted structural analysis and modeling techniques, such as finite element analysis, to investigate structural analysis, heat transfer, fluid flow, and mass transport properties. Additionally, internal and hard to access features in the as-built structure can be readily accessed, viewed and analyzed in the digital twin. This provides a full 360 degree inspection as well as an “inside-out” verification capability. Because the digital twin replicates the actual, as-built structure, the results of such analysis on the digital twin is highly correlated to the actual, as-built structure. As such, one can statistically assess confidence in as-built product (for a desired parameter such as strength) based on the testing results on the digital twin. In contrast to the myriad pre-, during, and post-manufacturing quality assurance methods of conventional manufacturing, which are often destructive-based techniques and/or inspection-limited, establishing the quality assessment during the manufacture of the integral and unitary structure for the fission reactors and fuel loaded reactor spaces disclosed herein not only offsets the difficulty of post-build validation of a complex product, but also renders it unnecessary with the more direct assurance provided by an inside-out assessment inspecting with high resolution layer-by-layer (at a minimum) resolution, regressed against a continuous input of process monitoring data along each additively manufactured layer. Further, a virtual inspection can be accomplished with a model constructed out of as-built data processed analytically as it is collected during manufacture (e.g., the “digital twin” to the physical as-built product). Combined with the predictive power of a machine that learns of anomalies of consequence to quality (using stored and in-process data) and the ability to monitor, interpret and report the status (compared to the continuously updated baseline) of the final as-built product, the additive manufacturing methods disclosed herein avoid defects prior to occurrence as well as statistically assesses confidence in as-built product viability on a global basis across the whole as-built product based on the as-manufactured condition as monitored and recorded during the manufacturing process. A fission reactor having dimensions of 16-inch diameter and 24-inch height was modeled. The reactor had 6 cylindrical spaces and its axial height was divided into 20 equally spaced axial levels resulting in 2520 individual primary channels and 2520 individual secondary channels for potential fuel loading. FIG. 3A illustrates the arrangement of features on an axial level from a top perspective view. Example dimensions for structural considerations included: (a) outside perimeter of shell 102 of 0.75 inches, (b) thickness of walls of primary channels of 3 mm, and (c) thickness of axially extending rings and the webbings of 2 mm, but smaller dimensions can be used, which will provide the potential for additional fuel loading, or larger dimensions can be used, which provides strength. Following are the volumes for the shell metal, primary channels, and secondary channels (Uranium metal): Shell metal volume2258.9 in3 (46.8% of core volume)Primary channel volume1202.2 in3 (24.9% of core volume)Secondary channel volume (fuel1364.3 in3 (28.3% of core volume)volume) The Uranium capacity for Example 1 was calculated. In the above 16″×24″ configuration with a combined secondary channel volume (fuel volume) of 1364.3 in3, a maximum U235 weight of 187 pounds would be possible if all secondary channels were filled with U235 at 20% enrichment. However, a plenum in each chamber allowing for an off-gas volume of 10% was incorporated. This resulted in a maximum U235 weight of 149.6 pounds at 20% enrichment. While this is much more than needed for criticality, this excess capacity enables tuning enrichments radially and axially for optimal fuel cycle efficiency and cycle length. The size of the primary channels was arbitrarily chosen to be 18 mm in diameter with a 3 mm wall to provide a balance between strength and flow area. As a result, 127 holes were created with a combined flow area of over 50 int. The number of holes and size of the flow area allows for a vast number of possible flow channels, moderator rods, control rod locations, scram rod locations, and instrumentation needs. Desired power levels, fluid choices, moderator materials, control rod material and enrichments will drive specific purpose for each hole. As seen in the drawings, every secondary channel with fissionable nuclear fuel, e.g., uranium, is connected to two halves of two different primary channels. Therefore, as long as every other primary channel is dedicated for fluid flow, each secondary channel with fuel will transfer heat into an adjacent primary channel for heat transfer purpose. Furthermore, 60 primary channel locations can be dedicated for non-flow needs (moderator, control, scram, and instrumentation). Considerations of plug design can also enable heat transfer in non-flow locations by not using the entire 18 mm primary channel size. For example, control rod designs with a finned shape could provide reactivity control and adequate flow simultaneously in a primary channel. The example reactor has radial and axial enrichment advantages. Because each secondary channel with fissionable nuclear fuel is independent to every other secondary channel with fissionable nuclear fuel, custom enrichments can be chosen in both the radial and axial direction. Experience has shown this can improve fuel cycle efficiency by as much as 20% and balance shell temperatures. It is also conceivable to provide an infinite number of Uranium enrichments via additive manufacturing. For example by using a combination of only depleted Uranium wire and 20% enriched Uranium wire, any enrichment between these two extremes are possible (in contrast, in traditional practices, nuclear manufacturing companies typically limit themselves to less than 10 different enrichments due to, e.g., the complexity of non-additive manufacturing). The example reactor is expandable and scalable. Although modeled with dimensions for a 16″ by 24″ reactor, any reactor size greater than 12″ by 18″ is possible. In addition, if the chamber radial width is kept to the details within this design and following the six-fold radial symmetry concept discussed above, any number of additional axially extending rings 140 can be added. The result will be a continuing highly symmetric configuration with all secondary channels with fissionable nuclear fuel identical in size, regardless of reactor size. A computational platform (referred to herein as the Universal Inverted Reactor Computational Platform or “UIRCP”) consisting of ANSYS engineering simulation and 3D design software, SolidWorks solid modeling computer-aided design and computer-aided engineering computer program, and Monte Carlo N-Particle Transport Code (“MCNP”) nuclear processes simulation program were utilized and applied to solve for the ideal thermal configuration of a universal inverted reactor design. Because fuel and cladding abut each other in the universal inverted reactor design disclosed herein (see, e.g., FIG. 4 and related description herein), there is a high potential for increased thermal stresses due to varying thermal expansion rates for the two materials that share the noted interface. The UIRCP was applied to address this issue and the fuel enrichment and reactor geometry was iterated to homogenize the radial thermal gradient. It should be noted that the axial thermal gradient and overall peak temperature are unavoidable due to the nature of a linear heat exchange and was not part of the UIRCP process. MCNP was applied in the UIRCP process to the modeled universal inverted reactor design to calculate the MeV/gram of each contiguous fuel element and check criticality. First, an input deck must be made based on the user's geometric and material inputs. Other user inputs can include material for coolant, fuel, clad, and reflector. This is done by reading the user inputs, outputting the geometry in MCNP format (binary geometry), labelling each cell with the desired material, and setting up the neutronic physics. The input deck is then run and the user given the option to review the geometry at this point by utilizing the visualization software functionality of the MCNP. The MCNP output is searched to find the MeV/gram associated with each fuel element, converted to W/m3, and saved in a separate file. All these steps are controlled by a governing batch file that calls the necessary commands and sub-programs. FIG. 10A shows a screenshot 500 of the initial user interface and FIG. 10B shows a screenshot 510 of the MCNPX geometry review. SolidWorks was applied in the UIRCP process to update the computer aided design (CAD) solid model of the reference reactor geometry based on user inputs. The user inputs were base geometry selection. The variables that were updated by the user were geometry variables: ring spacing, number of rings, clad thickness, passageway ID and OD, number of axial segments, and overall height. FIGS. 11A and 11B show diagrams 600, 610 with geometric structure and dimensions for geometric-related variables used in this example. Note the similarity between geometric structure in FIGS. 11A-B and the design of the reactor space 108 in FIG. FIGS. 3A-B. Using the universal inverted reactor design as the base design (such as the design shown and described with respect to FIGS. 11A-B), solid models of the fuel, cladding, and passageways exist a priori. The UIRCP process calls a batch file that opens SolidWorks, runs sub-programs that update geometry global variables, runs a VBA (Visual Basic for Applications) program to suppress unwanted geometry, rebuilds the design with the new geometric parameters, and saves a one-sixth core parasolid. The parasolid 700 (see FIG. 12) is an example of a core parasolid resulting from the process. Note the similarity between this core parasolid 700 in FIG. 12 and the design of the reactor space 108 in FIG. FIGS. 3A-B, including the shell, axial cylinder, plurality of axially extending rings, plurality of primary axial tubes, primary channels, plurality of webbings and plurality of secondary channels. The UIRCP process and the SolidWorks modeling can vary and update number of rings, passageway size, fuel size, overall reactor size, and interstitial clad to optimize the universal inverted reactor design. ANSYS was applied in the UIRCP process to solve a thermal-hydraulic problem in which the fuel produces heat that warms the coolant flowing through the passageways. A computational fluid dynamics (CFD) tool, such as ANSYS FLUENT, and a structural analysis tool, such as the finite element analysis (FEA) based ANSYS Mechanical, can be used. Using ANSYS FLUENT and ANSYS Mechanical, a j_script journal was called to insert the core parasolid, such as core parasolid 700, resulting from the UIRCP process applying SolidWorks. ANSYS mechanical was then opened to mesh the parasolid, differentiating between solid and fluid. A script was generated to control FLUENT; calling that script started by opening FLUENT. The script included referencing user inputs for the materials of the fuel, cladding, and coolant, which were then updated based on those inputs. Any mesh interfaces between the fuel and cladding were split (which prevents errant results that would otherwise result because the solid mesh would consider the interface as a uniform piece). The coolant inlets and outlets were set to the user input velocity and ambient temperature. Fuel elements were given the appropriate internal heat generation based on the MCNP output. FLUENT then ran a thermal-hydraulic simulation and produced a temperature contour map. FIG. 13 is an example of a temperature contour map 800 resulting from running FLUENT as outlined by the steps above. The UIRCP process interfaced the ANSYS, SolidWorks, MCNP programs into a singular software automation in which the individual software iterates towards a final result based on a user's identified optimization technique. The interface operations were performed by saving the necessary information from one software's output, which waits until it is called while another software is being automated. For example, the SolidWorks and ANSYS programs communicate through solid modeling. After SolidWorks performs the geometry update and saves a parasolid, ANSYS calls that parasolid as a base geometry to perform the thermal-hydraulic analysis. In another example, the MCNP and ANSYS programs communicate via the relationship between fuel enrichments and radial thermal gradients. During the first iteration, MCNP runs the initial neutronic simulation using the initial fuel enrichment level for all fuel elements, this equates to an internal heat generation (W/m3) per fuel element. The internal heat generation values are saved, waiting for ANSYS to call them. When ANSYS runs the thermal-hydraulic simulation, it saves the radial thermal profile. The thermal profile informs MCNP on what the next iteration enrichments will be in an attempt to trend towards a zero slope for radial thermal gradient (if that is the identified optimization technique). This process in repeated until an acceptable level of thermal gradient is reached. This acts as the main iterative loop for the UIRCP process. The neutronics of an example embodiment of the fission reactor 100 were investigated. The investigated fission reactor 100 utilized low enriched uranium (LEU) with 19.75 wt. % U-235. The fission reactor had 10 fuel rings located within the reactor space inside a shell. The core diameter was 434.7 mm and the core height was 800 mm. A 15 cm thick beryllium reflector surrounded the core. The neutronics were modeled using Monte Carlo N-Particle Transport Code 6 (“MCNP6”) nuclear processes simulation program. It was determined that steady state operation (k-eff=1.0) would require a sequence of control rod maneuvers throughout the operational life of the core. FIGS. 14A and 14B show core power-peaking profiles from the MCNP6 nuclear processes simulation program when control rods are removed from the core. The profile 900 in FIG. 14A shows power (normalized to average) as a function of core height (meters, top to bottom) and has a peaking factor (all control rods fully withdrawn) of 1.49, an axial peaking location at 0.39 m, and a k-effective of 1.06375±0.00036 (control rods fully withdrawn). The profile 910 in FIG. 14B shows power (normalized to average) as a function of radial distance (meters) and has a peaking factor (all control rods fully withdrawn) of 1.12, an axial peaking location at 0.0186 m, and a k-effective of 1.06375±0.00036 (control rods fully withdrawn). FIG. 14C shows a profile 920 of neutron flux (normalized to total flux) as a function of neutron energy (MeV) and has a k-effective of 1.06375±0.00036 (all control rods fully withdrawn). For shutdown purposes, k-effective for the modeled reactor with control rods fully inserted was 0.94211±0.00034. The UIRCP and the process described above gives engineers flexibility to change reactor type, materials, and base geometry. The final result is a reactor design that can be manufactured using additive manufacturing techniques and gives engineers a tool for a range of power applications. Additionally, specific features of the disclosed fission reactor can be optimized through dedicated routines in the UIRCP process. For example, in addition to enrichment optimization discussed above, the UIRCP process can be used to optimize: (a) passageway size, based on one or more of coolant substance heat transfer efficiency, radial thermal gradient, and axial thermal gradient; (b) ring width, for example, based on radial thermal gradient; and (c) cladding thickness, for example, based on radial thermal stresses. Additionally, the UIRCP and the process described herein can be effectively applied to new reactor designs. Engineers are tasked to vet criticality, thermal-hydraulics, and material specifications. This initial project design and evaluation can take months (up to a year) and cost millions to obtain an initial answer on the viability of a new reactor design. However, the UIRCP and the process described herein provides thermal-hydraulic, neutronic, and geometric knowledge of a new reactor in a couple days. As a result, one can have a preliminary determination of the usability of a reactor design, and optimize that design, in reduced time and with reduced costs relative to current practices. The fission reactor 100 in FIG. 2A has baseline characteristics including power output of 1 MWth (+250 kWe), ZrH moderated, Helium-cooled, Brayton thermodynamic cycle, and monolithic with rotationally symmetric. However, the fission reactor 100 can be larger or smaller, i.e., is scalable, and can have alternative characteristics as disclosed and described herein. Fission reactors 100 disclosed herein can be used in suitable applications including, but not limited to, terrestrial power sources, remote power or off-grid applications, space power, space propulsion, isotope production, directed energy applications, commercial power applications, and desalination. Although generally described herein in connection with a pressurized water reactor (PWR reactors) and with water as a primary coolant, the structures and methods disclosed herein can also be applicable to other reactor systems including boiling water reactors (BWR reactors), deuterium oxide (heavy water) moderator reactors such as CANDU reactors, light water reactors (LWR reactors), pebble bed reactors (PBR reactors), nuclear thermal propulsion reactors (NTP reactors), both commercial and research reactors, and utilize other primary coolants, such as helium, hydrogen, methane, molten salts, and liquid metals. Although described herein using additive manufacturing techniques, subtractive manufacturing techniques as well as a combination of additive and subtractive manufacturing techniques can be employed to manufacture the fission reactor and related structures. As such, the in-situ techniques and predictive quality assurance methods can be adapted for use in such subtractive manufacturing/combination manufacturing environments. An example of subtractive manufacturing techniques include machining, such as milling and boring, a body to a rough, semi-finished shape followed by finish machining, such as electrical discharge machining (EDM). Other subtractive manufacturing methods can be used, such as electron beam machining (EBM). Although described in connection with manufacturing the universal inverted reactor shown and described herein, the additive manufacturing methods and predictive quality assurance methods disclosed herein can be applied to the manufacture of other technologies, including in the petro-chemical industries (for example, for chemical reaction vessels), in the aerospace industry (for example, for parts of turbines including turbine blades and housings, and for parts for missiles and rockets including combustion chambers, nozzles, valves, and coolant piping). While reference has been made to specific embodiments, it is apparent that other embodiments and variations can be devised by others skilled in the art without departing from their spirit and scope. The appended claims are intended to be construed to include all such embodiments and equivalent variations.
	abstract	A computing system determines a full motion range of a target, wherein the full motion range of the target defines an internal target volume (ITV). The computing system identifies a partial motion range of the target, wherein the partial motion range is a subset of the full motion range of the target. The computing system generates a partial-ITV based on the identified partial motion range, wherein the partial-ITV is a volume swept by the target as the target moves through the partial motion range, the partial-ITV being smaller than the ITV. The computing system generates a treatment plan to deliver treatment to the partial-ITV.
040627233	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1 there is shown generally a liquid metal cooled, fast spectrum reactor 10 comprising a reactor vessel 12 and a reactor vessel head 14 vertically disposed and supported within a concrete containment structure 16. The reactor vessel 12 houses a nuclear core 18 comprised of a plurality of fuel assemblies 20 arranged and supported in a fixed array by support structure (not shown), as is well known in the art. The fuel assemblies 20 are comprised of fuel pins within a shroud or housing that contains nuclear material for sustaining a nuclear chain reaction. The housings 22 of fuel assemblies 20, as is conventional, are each of a hexagonal cross-section. Interspersed within an array of hexagonally shaped fuel assemblies 20 and arranged in a regular pattern are a plurality of hexagonally shaped control assemblies 24. The control assemblies 24 are comprised of an elongated, hexagonally shaped housing 26 which forms an aperture at its upper end in the array, and control elements (not shown), such as rods of a neutron absorbing material, adapted for longitudinal movement therewithin. Drive means 28 and 30, supported on the reactor vessel head 14 and extending downward therethrough (to be described in greater detail below) are coupled to the control elements for effecting longitudinal movement thereof to control and regulate the nuclear chain reaction in the core 18. By way of the nuclear chain reaction, substantial amounts of heat are generated within the core 18 and conveyed to a primary coolant, such as for example liquid sodium, which is introduced into and removed from the reactor core 18 by coolant inlet and outlet conduits 32 and 34 respectively. The heated coolant may then be cooled by a heat exchange system (not shown) to generate steam which is passed to a turbine which drives an electric generator for the ultimate transformation of thermal energy into electrical energy. Generally, the reactor vessel head 14 serves to seal the reactor vessel 12 and provide biological shielding, thermal insulation and alignment between the nuclear core 18 and interfacing system such as control, instrumentation, and core access systems. It is a general requirement that all penetration and/or leakage paths around and through the reactor vessel head 14 be sealed to prevent ingress of gases into the reactor vessel 12 which might react with the liquid metal coolant and egress therefrom of cover gas possibly contaminated by released fission products and nuclear radiation. The biological shielding and thermal insulation may comprise a series of shielding blocks 36 and reflector insultating plates 38 housed within a welded steel enclosure structure 40. For practical considerations related to safety and ease of manufacturing and construction of liquid metal cooled fast spectrum reactors 10, control of the reactor operations take place through the top of the reactor vessel 12. Additionally, it is desirable that instrumentation of fuel assemblies 20 for monitoring the behavior of the liquid coolant in a nuclear core 18 also take place through the reactor vessel head 14. Still further, as can be appreciated, symmetrical patterns of vessel head penetrations are desirable. A symmetrical control pattern permits more efficient and finer control of the reactor 10 while a symmetrical instrumentation pattern minimizes the complexity of mechanisms for locating the instruments in the fuel assemblies 20. Complicating achievement of the above noted desirable features is the fact that it is equally desirable to provide a core access system in which only a small portion or area of the core 18 is exposed for refueling at any one time. While some systems have been developed which accommodate a symmetrical control and instrumentation pattern and which provide for through the head, line of sight refueling (see for example U.S. application Ser. No. 537,283 filed Dec. 30, 1974 for "Control Rod and/or Instrument Tree Assembly" by Noyes et al and U.S. application Ser. No. 537,284 filed Dec. 30, 1974 for "Core Access System for Nuclear Reactors" by Dupen). Such systems are not always applicable nor can they be utilized in the manner described for all reactor designs. For example, the above mentioned pending U.S. Applications disclose a control and/or instrumentation scheme and core access system respectively for gas spectrum reactors which have cores comprised of relatively large hexagonally shaped assemblies. While such a system is theoretically applicable for smaller subassembly sizes, it is not easily adaptable because of physical restraints in regards to the size of components for use with relatively small size assemblies comprising the nuclear core. For example, with small sized assemblies, say on the order of 4 inches across the flat, the number of penetrations through the reactor vessel head per the instrumentation and control schemes disclosed in the above cited applications is greatly increased for a given area. This then presents problems with regard to the spacing between the control drive mechanisms supporting nozzles on the reactor vessel head which in turn presents problems with regard to the placing of rotational supporting bearings for rotatable access plugs between adjacent nozzles. Accordingly, different systems might be appropriate for a given nuclear reactor depending on the size of assemblies making up the core. It is to an alternative and improved instrumentation and control scheme and core access system that the present application and its companion application, entitled "Plug Mounted Core Servicing Apparatus" and filed on the same date as the present application are directed. It is to be noted, of course, that while the present invention as well as the inventions described in the other above mentioned co-pending applications have been designed for a particular application with a given size of fuel assembly, that the inventions may have application with regard to other size fuel assemblies. The core servicing apparatus of the present invention provides a symmetrical quite regular arrangement of core servicing elements with respect to the nuclear core so as to achieve the desirable advantages which result from such an arrangement. By servicing elements or means, it is meant that the means provides a service to the reactor. Customarily, this includes the services of monitoring, inspecting or controlling the nuclear core, although other services may also be performed. To carry out such conventional services, a servicing element may include for example a thermo-couple, an eddy current flow meter probe, a neutron detector, or a neutron absorption element. The symmetrical pattern afforded by the present invention, accordingly results in a regular pattern or arrangement of such elements and thereby provides for complete and adequate services for the core. As noted hereinabove, with respect to smaller size assemblies comprising the nuclear core, the problem of limiting the number of penetrations in the reactor vessel head becomes more acute. As noted in both the above mentioned co-pending application, Ser. Nos. 537,283 and 537,284 it is desirable that servicing means be placed either in or with respect to every assembly of the nuclear core. With respect to each penetration through the reactor vessel head then it is desirable that the core servicing apparatus be placed in and extending through such penetrations to provide services to a large number of the assemblies and thus a large area of the nuclear core. To accomplish this, and still maintain a reduced penetration size, it is necessary or desirable that means be provided for laterally displacing the servicing means, which extend through the limited size penetration, outward with respect to the boundaries of that penetration. That is, it is desirable to pass the core servicing apparatus in a compacted position through a relatively small size penetration and then laterally expand or extend the apparatus from the compacted state. In this way a larger area of the core may be serviced through a limited sized penetration. Such an arrangement for providing lateral displacement of servicing means is complicated further when the size of the subassemblies is reduced, as a certain size must be maintained for the penetration in order to provide room for components which can't be reduced in size and which are necessary for other interfacing systems. Accordingly, the core servicing apparatus of the present invention includes a removable plug from which is supported and suspended the servicing element for the nuclear core. The removable plug in turn is supported within an opening provided in the reactor vessel head. Briefly stated, and to be described in greater detail hereinbelow, the core servicing apparatus comprises a plurality of support columns suspended from the plug, rigid support means laterally extending and supported by each of the support columns, and a plurality of servicing means which are in turn supported from the laterally extending rigid support means. Means are then provided for rotating the rigid support means and servicing means from a condensed position which lies totally within the coextensive boundaries of the plug opening, to an expanded position in which some of the rigid support means and servicing means lie outside of the coextensive boundaries of the plug. In this way a core area can be serviced by a removable plug in which the cross-sectional dimension of the plug is less than the cross-sectional area of the core which is serviced by the servicing means suspended from the plug. With such an arrangement then, in which servicing apparatus is supported from a plurality of removable plugs which are in turn supported by the reactor vessel head, direct line of sight through the head refueling can be accomplished as another aspect of the present invention. Such refueling involves replacing the removable plugs in the reactor vessel head with extension skirts and refueling plugs mounted on top of the reactor vessel head within the penetration vacated by the core servicing plug. Briefly stated, and to be described in greater detail hereinbelow, the skirt extension has laterally extending portions on which are supported rotational supporting bearings for the refueling plug. The reactor vessel head includes a large rotatable cover which overlies the entire core and which includes penetrations therethrough for the core servicing and refueling plugs. With such an arrangement in which the core servicing plugs are replaced with refueling plugs which are rotatable, the entire core can be serviced for refueling with a limited number of penetrations for the refueling plugs by proper rotation of the refueling plugs and of the vessel head large rotating cover. Such rotation can position the access penetration in the reflecting plugs over each and every fuel assembly forming the core of the reactor. PLUG MOUNTED CORE SERVICING APPARATUS More specifically now, referring to FIG. 1 the reactor vessel head 14 includes a relatively large rotatable plug or cover 42 which is supported in an opening 44 in the reactor vessel head. This large rotatable cover 42 is virtually identical to that disclosed in co-pending application Ser. No. 537,284 with respect to its composition. Of course, the size or lateral dimension of the plug and the penetrations therethrough for the plug mounted core servicing apparatus are different. During reactor operation the large rotatable cover is stationarily supported on appropriae flanges 46 provided in the annular support ring 48 which defines the boundary of the large reactor vessel opening. As more fully described in co-pending application Ser. No. 537,284, the plug 42 is mounted for rotational movement by means of appropriate bearings 50 spaced about the periphery of the large cover and supported on the annular support ring 48. The capability for rotational movement is achieved by actuating hydraulic screw jackets 52 to raise the large cover off the support flanges and to place the load on the bearings. The large rotatable cover 42 includes a plurality of penetrations 54 therethrough, in which are supported the core servicing plugs 55. In the embodiments shown and described herein, eight of such penetrations are provided. As best shown in FIG. 5, the boundaries of these openings or penetrations through the large rotatable cover are defined by upwardly extending skirts 56 welded to the upper surface of the reactor vessel head about the penetrations. As with the large annular support ring for the large rotatable cover, the penetrations through the large cover include inwardly extending flanges 58 which serve to support the removable plugs of the core servicing apparatus during normal reactor operation. Also included along the upper surface of the large rotatable cover are a plurality of upwardly extending nozzles 60 welded to and spaced about the ligament between the penetrations 54 of the large rotatable cover. These nozzles are hollow and extend through the entire thickness of the large rotatable cover 42 and serve as a housing to support single core servicing apparatus to complete the core servicing or instrumentation and control of the nuclear core as will be described in more detail hereinbelow. As can be seen from FIGS. 1 and 5, the upwardly extending nozzles on the large rotatable cover are located between adjacent upwardly extending skirts 56 and extend a short distance above the upper elevation of these skirts. Each of the core servicing plugs 55 are of the same composition as the rest of the reactor vessel head, i.e., they include graphite blocks 62 and reflector insulating plates 64 within a weld steel enclosure 66. The lower end 68 of the plug 55 is of a reduced diameter to provide a flange 70 to rest on the flange 58 within the penetration in the large rotatable cover. Appropriate sealing means such as O-ring type seals 72 are provided along the cylindrical surface of the plugs 55 so as to effectively mate and seal the plugs when mounted in the penetrations 54 of the large rotatable cover. As noted above, these plugs are stationarily supported and do not rotate, thus no rotational bearings are provided. In the preferred embodiment, each of the core servicing plugs 55 includes seven upwardly extending nozzles 74, 74', similar to those nozzles provided on the large rotatable cover 42. Within the nozzles there is provided a penetration 76 through the plug within which are supported a downwardly extending support column 78, 78' of the core servicing apparatus 80, 80' as will be described in detail below. There is one central nozzle somewhat larger in diameter than the remaining six nozzles which surround the central nozzle and are located near the periphery of the plugs 55. As with the arrangement disclosed in co-pending application Ser. No. 537,284, the size or lateral dimension of the core servicing apparatus support plug 55 is related to the size of specified interfitting groups of assemblies 82 formed in the nuclear core. Again as in the aforementioned application, these interfitting groups of assemblies 82 are designated potential control clusters and in the preferred embodiment comprise a central hexagonally shaped assembly and six surrounding hexagonally shaped assemblies contiguous with the center assembly (see FIGS. 4 and 6). The nozzles 74, 74' on each of the plugs 55 are located such that the center lines of two adjacent nozzles 74, 74' coincide with the center of two adjacent and interfitting potential control clusters 82. The specific size of each of the nozzles and thus the internal diameter of the nozzles and of the penetrations through the plug can and do vary depending on whether the nozzle is located at the periphery of the plug or located in the center of the plug as best shown in FIG. 5. Peripheral nozzles will hereinafter be designated by the reference number 74 and central nozzles by the number 74'. Also throughout the description, where like reference numerals are used, the number bearing a "prime" (') designation will be associated with a central core servicing assembly. This size difference is a result of the fact that the central and peripherally located core servicing assemblies service different numbers of fuel assemblies in the core. Accordingly, the diameter or lateral dimension of the small plugs 55 is somewhat greater than twice the distance between the centers of two adjacent and interfitting potential control clusters 82. In any event, however, it is less than four times the distance as the boundary of the plug must fall between two adjacent nozzles, one on the plug and one on the large rotatable cover. As noted above, there are two general types of core servicing assemblies, central instrumentation servicing assemblies 80' and peripheral instrumentation servicing assemblies 80. Each of the peripheral assemblies 80 service 10 of the fuel assemblies 20 found in the core and the central assemblies 80' service 19 of such fuel assemblies. With either type of servicing assembly 80, 80', there is provided a guide tube and instrument tree support column 84, 84', laterally extending support structure 86, 86', and servicing means 88 such as instrumentation for monitoring the flow and temperature of the liquid metal coolant or means for effecting longitudinal movement of control or neutron absorbing control elements. The servicing means 88, i.e., either the instrumentation or control element drive extensions are supported from the support column and have means extending upwardly within the guide tube and instrument tree support column to exit above the reactor vessel head. Each of the servicing means is rigidly laterally fixed with respect to the support column 84, 84' and in some of the assemblies, namely the peripheral servicing assemblies 80, means are provided for rotating servicing means about the center line or access of the peripheral servicing assembly. As seen generally in FIG. 7, the laterally extending structure 86, 86' is fixed to a central support column 78, 78' which extends upwardly into and through the outer guide tube and instrument tree column 84, 84' which terminates some distance above the nuclear core. In the case of instrumentation, the instrument probes 88, which typically comprise either a thermo-couple or a flow meter, are suspended from the ends of the support structure 86, 86' in a fixed array with respect to the nuclear core 18 and with respect to the support column 84, 84'. Electrical leads 116 housed in a flexible conduit 92 connected to the instrument probes pass upwardly from the support structure 86, 86' and radially inward around the central support shaft 78, 78'. The leads then extend upward around the shaft into the guide tube and instrument tree 84, 84' and exit above the nozzles 74, 74' supported on the plug 55 through an electrical connector assembly 94. In case of control drive extensions, in the preferred embodiment the extensions extend downward within the central support shaft 78, 78' and enter into appropriate apertures in a control assembly which is in alignment with the axis of the servicing assembly. Of course, this means that the large rotatable cover and the servicing plugs must be maintained in a fixed position with respect to the core. This can be easily accomplished by the use of guide pins (not shown). More specifically now, turning first to a peripheral servicing assembly 80 as shown in greater detail in FIGS. 9-12, the peripheral assembly includes a vertically extending support column comprised of an outer cylindrical tubular member 84 and a concentrically positioned inner cylindrical tubular member 78 supported for longitudinal movement relative to the plug 55 within one of the peripheral nozzles 74 supported on the servicing plugs. Basically, the general arrangement and composition of the components of the support column 84 at the elevation of the reactor vessel head 14 is the same as the arrangement and composition of the components for the control rod and/or instrumentation assemblies of previously cited co-pending application Ser. No. 537,283, which application is hereby incorporated by reference. Accordingly, the detailed description of the arrangement will not be set forth herein. The outer guide tube 84 is supported within the upwardly extending nozzle 74 and extends downward through the thickness of the head to a position above the nuclear core 18. The outer guide tube 84 includes a laterally extending support flange 96 which, in the assembly's lowermost position, rests on the ledge 98 of the nozzle 74. Within the outer guide tube a central tube 100, concentric with the guide tube 84 extends the full length of the outer tube and downwardly below the termination of the outer tubes and is in alignment with and engages an assembly 20 of the nuclear core. For control type assemblies which include control element drive extensions, the inner tube 100 houses and serves as a guide for the control element drive extension 102 which also extends downwardly from above the reactor vessel head 14 where it is connected to a drive mechanism, either a safety drive mechanism 28 or a shim type drive mechanism 30, and extends downwardly into a control assembly 24 where the actuator is then coupled to the control element (not shown). Spaced equally about the inner tubular member are a plurality of instrumentation pull tubes 104 which are supported in guide tubes 106 at the upper end of the assemblies. The guide tubes 106 extend downwardly partially into the reactor vessel head 14 where they are in registry with appropriate bores 108 and shielding cylinders 110 which are provided along the thickness of the reactor vessel head. The shielding cylinders 110 are attached to the central tubular member 100 and thus the guide tube and instrument tree and accordingly move therewith during vertical movement of the assembly. As pointed out above, the shielding cylinders 110 have holes or bores 108 therethrough in alignment or registry with a mating guide tube 106 through which the pull tubes 104 extend downward to the termination of the outer tubular member 84 (see Number 112 FIG. 9). Within each of the pull tubes the electrical leads 92 connected to the instrumentation probes 88 extend upwardly and are packed with an appropriate shielding material such as powdered steel shot. At the upper ends of the guide tube, the pull tubes 104 are maintained in fixed and sealing relationship by means of an appropriate yoke 114 (see FIG. 10). The leads 116 from the pull tube extend upwardly therefrom within the outer guide tube 106 and nozzle 74 and exit from the assembly through an electrical connector assembly 94 mounted to the top of the nozzle 74. At the lower end of the outer tubular member 84 the electrical leads 116 are housed within a flexible conduit 92 which in turn is attached to the bottoms of the pull tubes 104 and extends downwardly along the outside of the inner tubular member or support column 78. Referring to FIGS. 9, 11 and 12, the conduits 92 are initially spaced equally about the inner tubular member 78 as they exit from the outer tubular member 84 and eventually near the lower end of the inner tubular member 78 are grouped around one-half or side thereof. Appropriate guides and clamps 118 are provided along the outer surface of the inner tubular member 78 to maintain the conduits 92 in a fixed position. The lower end 120 of the inner tubular member 78, which is of an enlarged size compared to the upper portion for the purpose of housing the control rod latch mechanisms, includes a spreader assembly 122 which is adapted to fit over the top end of the control rod receiving assembly 24. The spreader assembly 122 includes appropriate openings 124 for directing the liquid metal coolant, which flows upwardly around and through the assemblies, outwardly from the interior of the spreader 122. Extending laterally from the upper end of the spreader assembly 122 is the lateral support structure 86 comprised of a plurality of segments 128 located at various distances and at various angles from the enlarged lower section 120 of the inner tubular member. This lateal support structure includes a plurality of sockets 130 therein at the end of each arm 128 through which the flexible conduits 92 containing the instrument probe leads pass and are fixed so as to be supported in a generally vertical direction above the fuel assemblies 20 surrounding the control assemblies 24 which are to be serviced. As best seen in FIG. 9, when the spreader assembly 122 is engaged with a control assembly 24, the lateral support structure 86 is maintained in a fixed position several inches above the upper end of the fuel assemblies 120. this support structure 86 serves to tie the instrument probes 88 together in a fixed array which corresponds to the arrangement of fuel assemblies within the core; that is, the sockets 130 and accordingly the instrument probes are maintained in a fixed pattern such that when properly oriented they will lie directly above the centers of the fuel assemblies 20 surrounds the central assembly 24. As is seen in FIGS. 4 and 12, the fuel assemblies served by the instrument probes from a peripheral servicing assembly 80 comprise five (four if the central assembly is a control assembly) fuel assemblies of a potential control cluster plus five fuel assemblies from three adjacent interfitting potential control clusters. The assemblies serviced by a typical peripheral servicing assembly are identified by a star in FIG. 4 with the central potential control cluster and the three adjacent potential control clusters outlined in lines of heavy thickness. The significance of this pattern will be apparent as the description continues. As can be seen from FIG. 9, the instrument probes 88 are situated within guide rings 132 attached to the upwardly extending fins 134 of the fuel assemblies 20. The only difference between an instrument servicing assembly and a control servicing assembly lies in the fact that instrument servicing assemblies are provided with an instrumentation lead along the central axis in place of a control element drive extension. In such a configuration, the instrument lead is supported within the central tubular member 78 in an appropriate manner so as to extend vertically down therein and into the upper end of the fuel assembly 20 over which the spreader assembly 122 is situated. As pointed out above, in such a situation the peripheral assembly will serve ten fuel assemblies instead of the nine fuel assemblies plus one control assembly served by a control peripheral assembly. As has been noted hereinabove, the instrumentation probes or servicing means of a peripheral assembly 80 can be moved between a compacted state and an expanded or normal state. In the preferred embodiment, this displacement is accomplished by combined vertical and rotational movement of the instrumentation probes 88. Vertical movement of the peripheral assembly is accomplished in a manner similar to that utilized for the control rod and instrument assemblies of co-pending incorporated application Ser. No. 537,283. This involves the use of a control rod drive extension rod and an auxiliary drive unit which are temporarily removably secured to the upwardly extending nozzles 74 on the reactor vessel head 14 and which serve to replace control rod drive mechanisms and electrical connector assemblies. A further description of the process for raising the instrumentation and control servicing assemblies will be set forth hereinbelow. During vertical upward movement of the guide tube and instrument tree assembly (hereinafter GTIT) within the upwardly extending nozzles 74, an effective seal is maintained between these elements to prevent the egress of fission gases and contaminated cover gases from the reactor and ingress of oxygen or air into the reactor through the use of bellows type seals. In the preferred embodiment, these bellows type seals 136 are sealingly secured to the upwardly extending nozzles 74 and to the upper end of the GTIT outer tube 84. Depending on the extent of vertical travel which is desired, the bellows type seal may be attached to the outwardly extending flange from the GTIT and to the nozzle at its other end, so as to permit vertical movement while maintaining the seal. In the preferred embodiment, rotational movement of the instrument probes of peripheral assemblies is accomplished by means of a pin and slot arrangement at the upper elevation of the reactor vessel head. A plurality of pins 138 are provided along the inside wall of the upwardly extending instrumentation nozzle 74. As best seen in FIG. 17, these pins extend laterally inward and are adapted to engage and fit into appropriate slots 140 machined along the outer surface of the outer cylindrical tube 84 of the GTIT. These slots, which in the preferred embodiment comprise two slots located on diametrically opposite sides, extend along the length of the outer tube 84 and have a helical portion near the upper end thereof. The helical portion extends approximately half way around the tube in a gradual slope so that as the guide tube 84 and instrument tree is lowered within the nozzle 74 the pins 138 engaging the slots 140 serve to rotate the guide tube and instrument tree assembly 80 approximately 180.degree. . The slots 140 are arranged so that the full rotational movement of 180.degree. is achieved for the peripheral assemblies 80 near the lower end of vertical travel of the guide tube and instrument tree assembly. However, this rotational movement is completed with the instrument probes 88 are located a short distance above the nuclear core, so that the rotational movement will not cause interferences between downwardly depending instrument probes and the upwardly extending fins 134 on the assemblies forming the nuclear core. Straight portions of the slot 140 are provided at both the upper and lower ends of the helical portion so that only longitudinal, vertical motion of the GTIT assembly and accordingly the instrument probes occurs when the pins engage these portions of the slots. Referring now to FIGS. 13-16, the general arrangement for a central servicing assembly 80' is similar to that of a peripheral servicing assembly 80 with appropriate changes so as to accommodate a greater number of instrumentation probes and deletion of the provision for rotational movement. As the servicing means of a central assembly always remains within the coextensive boundaries of the instrumentation plug 55, it is not necessary to provide for rotation or other displacement of the servicing means between compacted and extended positions to permit removal of the plug from which it is suspended and to permit servicing of a portion of the nuclear core. In the preferred embodiment, the central servicing assemblies 80' are designed to service 19 assemblies, comprising a 7 assembly potential control cluster 82 and 12 adjacent assemblies of surrounding and interfitting potential control clusters. This arrangement can best be seen in FIG. 16. As with the peripheral assembly, the central assembly 80' overlies and is in axial alignment with the center assembly of a potential control cluster. This central assembly may be either a control assembly 24 or a conventional fuel assembly requiring instrumentation. Depending upon which it is, a control element drive extension 102 or an instrumentation lead is provided within the central inner tubular member 78'. Again, the central inner tubular member 78' extends downward into the core 18 and engages through a spreader 122', the upper end of a central assembly. Also provided at the lower end of the central assembly and attached to the inner tubular member is a support structure 86' comprising an array of support arms 128' extending laterally outward therefrom and having appropriate sockets 130' into which instrumentation conduits 92 are received and supported. A socket overlies the center of each of the 18 assembly positions which surround and are closest to the central assembly. These 18 surrounding assemblies comprise the first two rows radially outward from the central assembly. The flexible conduits 92 within which the instrumentation leads 116 are housed are attached to the upper ends of the instrumentation probes 88 and extend upwardly and inwardly to be positioned about the outside surface of the inner central tubular member 78'. From there the instrumentation leads 116 extend upwardly into pull tubes 104' which are located in the annular space between the outer tubular member 84' and the inner central tubular member 78'. As seen in FIG. 14, the pull tubes 104 are arranged in two rows at different radial distances from the centerline of the assembly. The pull tubes 104 extend upwardly and terminate a short distance above the reactor vessel head where they are housed within and sealed within appropriately provided guide tubes, also arranged in two rows about the centerline of the assembly. Again, an appropriate bellows seal is provided which is attached to the upwardly extending nozzle 74' and also to the outer guide tube 84' of the central instrument tree in order to provide sealing during vertical movement of the assembly. This arrangement is similar to that shown in FIG. 9 for a peripheral assembly and thus is not shown in a separate drawing. With the arrangement described above, the core of a liquid metal cooled fast breeder reactor, or at least a central portion thereof, can be fully instrumented, and a regular pattern of control obtained. As best seen in FIGS. 2, 4 and 6, the nuclear core is comprised of 469 assemblies, each assembly being 4 inches across the flats in the preferred embodiment. Eight core servicing plugs 55 are provided in a specified pattern shown in FIGS. 2 and 4. Each of these plugs is capable of servicing 79 assemblies if the preferred arrangements described hereinabove are utilized. As no open portions are provided in the nuclear core, it is necessary to provide additional instrumentation or control mechanisms at various positions between adjacent plugs. These are indicated in the figures by small individual nozzles 142 and guide tubes 144 respectively, which are supported over an assembly located between adjacent plugs 55. In accordance with the preferred embodiment, the assemblies over which these small nozzles and/or guide tubes are positioned are the central assemblies of potential control clusters and, accordingly, will be provided with either control rod drive mechanisms or instrumentation probes. In either case, the guide tube 144 which is supported within the nozzle 142 on the reactor vessel head extends all the way down into the core and engages the appropriate assembly with which it is aligned. Referring to FIG. 7, a control rod guide grid 146 is provided above the nuclear core having appropriate rings joined together to guide the guide tubes depending from the reactor vessel head for the smaller single servicing means and which provide lateral support therefore. The grids are open beneath the plugs so as not to provide any interference with regard to the assemblies thereof. A particular instrumentation and control pattern is disclosed in FIG. 2. As can be seen from FIG. 2, since each nozzle provided on the reactor vessel head overlies the central assembly of a potential control cluster, there are provided 7 safety control assemblies 148, 24 shim control assemblies 150, and 43 instrumented central assemblies 152. The control rods of a safety control assembly 7 are normally maintained during reactor operation in a raised position relative to the core to provide a safety margin for shutdown in the event of a reactor accident while the shim control rods provide the regulation of power output during normal reactor operation. The specific drive mechanisms and control rod extensions for such arrangements are adequately described in co-pending application Ser. No. 537,283, and as they are virtually identical in the preferred embodiment of the present invention, a description of these is not necessary. The remaining assemblies within the central portion of the core are provided with instrumentation probes such as described hereinabove. The pattern of these instrumentation probes with respect to the core is shown for the preferred embodiment in FIG. 4. As can be seen from FIG. 4, the central axis of the reactor vessel head conicides and is aligned with the center of the nuclear core (see reference number 153), and the central core servicing plug 55 (position shown in outline) is offset from the center. This arrangement of plugs having a concentric reactor vessel head and eccentric plug arrangement has been chosen in order to provide the requisite refueling coverage when refueling plugs and extension skirts replace the core servicing plugs 55 and rotation is permitted as will be described hereinbelow. While not all assemblies comprising the nuclear core are instrumented, the majority of the assemblies are. The three outermost rows of assemblies forming the core comprise the shielding and reflector assemblies and as such do not contain any nuclear fuel. The next three innermost rows comprise blanket region assemblies which contain nuclear fuel but which do not generate power. As can be seen from FIG. 4, all the assemblies in Rows 1-10 (counting from the center outward), the power generating portion of the nuclear core, are serviced by the plugs and single penetration servicing elements. Of the blanket assemblies, only 23 of the assemblies are serviced. While the concentric large plug and eccentric array of small plugs arrangement does not provide all the servicing of all the assemblies forming the active core, it does provide servicing for the major portion and presents no problems as a result of non-servicing (in particular not instrumenting) of some of the blanket region assemblies. An alternative arrangement which will provide complete servicing of all assemblies forming the active core is disclosed in FIG. 18. In this arrangement an eccentric large plug 42 is shown having a concentric array of small plugs 55. As can be seen from this figure, 7 plugs provide the requisite core servicing capability and are arranged concentrically with respect to the core; the reactor vessel head, however, is positioned eccentrically with respect to the core. In this arrangement the entire active core is provided with servicing elements and some other reflector and shielding assemblies also have instrumentation. While such an arrangement does provide for a better instrumentation coverage and thus servicing of the nuclear core, it is not preferred due to the increased cost in providing an eccentric reactor vessel head which will result in a larger vessel head and reactor vessel. In order to minimize and reduce the possibility of flow induced vibrations being set up in the guide tubes 78, 78' which support the instrument trees, a locking arrangement is employed to pin the 7 guide tubes in each plug together in a rigid structure. As best seen in FIGS. 6 and 7 and 16, the locking mechanism comprises a spider 154 which is welded to the central instrument tree guide tube 78' some distance above the lower end thereof. This spider arrangement is supplied with 12 spokes 156 which extend laterally outward therefrom and which are each furnished with an alignment pin 158 which extends downwardly at the extremity of the spokes. Appropriate collars 160 are fixed to each of the guide tubes 78 of the peripheral assemblies at an elevation above the lower end thereof. These collars 160 have mating holes or sockets 162 vertically arranged and oriented generally on the diametrically opposite side of the guide tube from the instrumentation probes 88 and support arms 86. The alignment pins 158 on the spider spokes 156 are adapted to engage and fit into the sockets 162 on the collars of the peripheral assemblies when the central assembly instrument tree is lowered into place in the core. Such an arrangement of locking insures that the peripheral assemblies 80 will be in correct alignment with respect to each other and with respect to the core assemblies and elements. The installation procedure for the plug mounted core instrumentation or servicing will now be described. Initially, 7 instrument tree assemblies 80, 80' mounted on each of the plugs 55 are maintained in a raised and compacted state and the plug 55 is lowered into position to rest on the appropriate flanges 70 within the opening 54 in the large rotatable cover 42. The arrangement and location of the instrument trees with the control rod drive extension 102 and instrument probes 88 is as shown in FIG. 8. The mechanism for maintaining the instrument trees in the raised position will be described hereafter. As can be seen from FIG. 8, the central instrument tree 80' is positioned the greatest distance above the nuclear core 18 and in the preferred embodiment this corresponds to a distance of approximately 53 inches. Directly beneath the central instrument tree 80' there are located a first set of three of the peripheral trees 80 in which the laterally extending support structure 86 and instrument probes have been rotated inwardly to lie completely beneath the plug and within the coextensive boundaries thereof as is evident upon viewing the lower end 68 of the plug in this figure. These three peripheral trees correspond to 3 alternately spaced trees and are located approximately 38 inches above the nuclear core. The lowermost set of instrument trees 80 correspond to the remaining three trees and are located approximately 25 inches above the nuclear core and are also in a compacted condensed state in which the lateral support structure 86 and probes are located totally within the coextensive boundaries of the plug. It is necessary to maintain these three elevations of instrument trees in order to fit all of the probes and support arms within the coextensive boundaries of the plug so as to permit insertion of the plugs through the penetration opening 54 in the reactor vessel head 14. Next the lowermost set of alternately spaced instrument trees are lowered with respect to the vessel head by appropriate means and, as this occurs, the guide pins 138 located within the nozzles 74 and engaging the camming slots 140 in the guide tubes 78 serve to rotate the laterally extending support structure 86 and instrument probes 88 approximately 180.degree. as the GTIT moves downward into the nuclear core wherein the instrument probes 88 are positioned within the guide rings 132 joining the upwardly extending fins 134 of several of the fuel assemblies forming the nuclear core. If these peripheral trees comprise control assemblies the control rod drive extension 102 is lowered within the control assembly is alignment with the center guide tube of the peripheral tree and engages a control rod such as shown in FIG. 7. The complete rotation of 180.degree. of course is achieved while the GTIT is totally above the nuclear core so as to prevent any interference between the instrument probes and the upper ends of the fuel assemblies. Next, the remaining three alternately spaced peripheral trees 80 are lowered, the guide pins, again, working in the cam slots to rotate the support structure 180.degree. to bring the probes 88 into alignment so that they may be lowered into engagement with the nuclear core. Finally, the central instrument tree 80' is lowered and the alignment pins 158 on the locking spider 154 slide into engagement with the appropriate holes 162 on the collars 160 of the peripheral trees 80 to lock the seven guide tubes together in a rigid structure. This is the position shown in FIG. 7. For refueling purposes, as will be described hereinbelow, it is necessary to completely remove the plug mounted instrumentation and core servicing assemblies from the reactor vessel head. In order to accomplish this removal, without interference, the trees must be stacked at three different levels such as shown in FIG. 8, this being the same arrangement shown for insertion of the plugs. Initially, the central tree 80' is raised approximately 53 inches to provide room underneath the plug for the peripheral trees 80, then three alternately spaced peripheral trees are raised approximately 38 inches and rotated inwardly by means of the pins engaging the camming slots and finally the three remaining peripheral trees are raised approximately 25 inches and also rotated inwardly to a compact state. Upon completion of these steps, all six of the peripheral instrument trees are within the coextensive plug boundaries and the plug may be removed from the reactor vessel head through the plug opening 54. Accomplishment of the raising and lowering of the instrument trees to permit insertion and removal of the plugs is accomplished in a manner similar to that described for the control rod and instrument tree assemblies of copending application Ser. No. 537,283. With the control rod drive mechanism 28 and electrical connector assemblies 94 removed from the nozzles 74, 74', 142 on the reactor vessel head 14, and on the plugs 55, an instrument tree drive extension lock 164 is mounted to the nozzle 74. As seen in schematic FIGS. 21a and 21b, the instrument tree drive extension lock 164 includes an extension at its lower end and the lock assembly at its upper end. An instrument tree drive extension lock is needed for each of the instrument tree assemblies and accordingly for each plug nozzle. Two lock assemblies 166, 168 are associated with each of the drive extension locks. One lock 166 is adapted to be moved into locking engagement with a lock seat 169 on the guide tube end cap when the guide tube end cap, which holds the instrument pull tubes 104 and maintains them in their lowered position, and accordingly the guide tube and instrument tree, have been lifted vertically a sufficient distance for the lower end of the guide tube instrument tube to clear the core and be maintained at its appropriate elevation. The other lock 168 is positioned for locking engagment with another lock seat (not shown) in the control rod drive extension when the guide tube and instrument tree tube is raised as noted hereinabove and the drive rod extension is withdrawn upwardly to its fullest extend within the guide tube. This final configuration is depicted schematically in FIG. 21b. In order to raise the guide tube instrument tree to any control rod drive extension, an auxiliary instrument tree drive unit 170 is removably secured, such as by bolting, to the upper end of the drive extension rod. The auxiliary drive unit 170 contains one drive which substitutes for the conventional drive means to operate the drive rod extension and a second drive which vertically moves the guide tube and instrument tree tube. Because the auxiliary drive unit 170 is installed following a scram or a shutdown of the reactor and the drive rod extension is in its fully inserted position the drive for the rod extension must run down the fullest extent for engagement with the upper end of the rod extension and the connection is made by rotating the axis through an angle sufficient to couple the drive means with a breech block lock on the drive extension. The second drive of the auxiliary drive unit 170 is run down and rotated into engagement with the seat and the guide tube end lock cap and, upon actuation of the drive, serves to raise the guide tube instrument tree tube. The extension portion of the CDEL is necessary in order to raise the guide tube and instrument tree assembly the requisite distance in order to provide space beneath the plug for stacking of all the instrument trees. As the upwardly extending nozzles on the reactor vessel head are approximately 28 inches in height, different length drive extensions are provided, the length depending upon the desired final height of the guide tube and instrument tree assemblies to be raised. With the instrument tree assemblies maintained in their upper position by means of the drive extension locks, the auxiliary drives may be removed and the plug removed from the penetration in the reactor vessel head in a manner described hereinbelow in discussing the refueling scheme. It should be apparent there are other types of drive means and locking means which may be employed in order to raise and maintain the instrument trees in their raised and condensed position. REFUELING PLUG SYSTEM Rotating plug systems for providing core access in liquid metal cooled fast breeder reactors provide a convenient arrangement for exposing only a small area of the core at one time and for permitting line of sight through the head refueling, as is well known in the art. When nozzles are provided on the reactor vessel head for guiding and directing control elements and instrumentation probes into the nuclear core, penetrations for the rotating plug system are generally required to be located between the upwardly extending nozzles and between the penetrations through the head for such core servicing apparatus. When the nozzles for the core servicing apparatus on the reactor vessel head are arranged in a regular pattern, such as disclosed hereinabove with reference to the instrumentation system, a problem is created in having sufficient distance between adjacent nozzles within and without the rotating plugs for placement of rotational supporting bearings therebetween for supporting the rotating plugs for rotational movement in the penetrations in the reactor vessel head. This problem is further compounded where smaller assemblies make up the nuclear core since such smaller assemblies necessitate the nozzles being located closer together for a given pattern of control and instrumentation. Unless nozzle size can be made very small, it is not possible, or it is extremely difficult to place rotational supporting bearings between adjacent nozzles to provide the rotational support for the rotating plugs. The present invention overcomes such problems by providing a skirt extension 172 and a refueling plug 174 which are substituted for instrumentation supporting plugs in the reactor vessel head. The skirt extension and refueling plug are interchanged with the instrumentation and control supporting plugs in the reactor vessel head. The skirt extension and refueling plug are interchanged with the instrumentation and control supporting plugs when core access is desired such as when the reactor is shut down for refueling. The refueling plug 174 is provided with an opening 176 therethrough which is sized to permit passage of fuel assemblies therethrough or for any other purposes in which core access is necessary. To prevent the escape of fission gases and contaminated cover gas from the reactor and to prevent the ingress of air into the reactor, sealing means are provided as well as removable closure means for sealing the penetration openings and other leakage paths which may exist by virtue of the substitution of refueling plugs for instrumentation and core servicing plugs. Referring now specifically to FIGS. 19 and 20, the skirt extension or bearing support skirt 172 comprises a generally cylindrical member 178. The lower end of the bearing support skirt is provided with flanged portions 180 extending laterally outward therefrom which are adapted to mate with similar flange portions 182 on the upper end of the upwardly extending skirt 56 surrounding the penetration opening 54 in the reactor vessel head 14. The flange portions 180, 182 are arranged such that they interfit between adjacent upwardly extending individual nozzles 143 arranged about the periphery of the penetration opening as best seen in FIG. 19. Sealing means, such as a double O-ring type seal 184, are provided where the bearing supporting skirt 172 and the upwardly extending skirt 56 meet to provide a gas tight seal. The upper end of the bearing supporting skirt 172 is provided with an outwardly extending flange 186 about its circumference on which is supported the rotational bearings 188 of the refueling plug 174. The bearings are of the crossed-roller type in which the axis of rotation of the rollers 190 alternates about the circumference of the bearing 188. These rollers are captured between inner and outer races and this arrangement is in turn supported in position on the flange 186 of the upper end of the bearing supporting skirt 178 by means such as bolts 192 spaced about the circumference of the annular bearing and flange. Appropriate O-ring type seals 194 are provided along the interior surface of the bearing supporting skirt near its upper end for sealing the leakage path between the skirt 178 and the refueling plug 174. As can be seen from FIG. 20, the outwardly extending flange 186 at the upper end of the bearing supporting skirt 178 and thus the bearing 188 are positioned above the upper elevation of the upwardly extending individual nozzle 142 adjacent to the skirt 56 surrounding the penetration in the reactor vessel head. Also in the embodiment shown, this flange is above a control element drive extension box 196 mounted on the nozzle 142 which maintains the instrument or control rod drive extension in an elevated position to permit rotation of the large rotatable cover 42 without interference by the core servicing elements therein with the nuclear core. In the preferred embodiment the refueling plug 174 is comprised of a handling plug 198 and a floor valve 200 each of which have axial penetrations or bores 202, 204 respectively therethrough sized to permit the insertion and removal of fuel assemblies and control assemblies therethrough. More particularly, these bores are sized to permit insertion of a fuel handling machine which is capable of extending down into the nuclear core and into which fuel assemblies will be raised and maintained in a bath of sodium or other liquid metal coolants as it is removed from the reactor vessel. Such a fuel handling machine is similar to that disclosed in copending application Ser. No. 430,292 entitled, "Nuclear Fuel Handling Apparatus." The handling plug 198 comprises a large substantially solid circular plug made of a suitable material, such as steel, which is inserted into the penetration opening provided in the large rotatable cover and which replaces the instrumentation servicing plug. The handling plug 198 is provided with a flanged surface 206 near its upper end which is sized to permit the lower end of the plug to be inserted in the opening defined by the skirt 56 and which engages a mating flanged surface 208 in the opening in the vessel head to limit the penetration of the plug into the opening. Appropriate sealing means such as O-ring type seals 210 are provided near the upper end to provide a gas tight seal. When first installed in the reactor, a shield plug 212 is inserted within the circular opening 202 in the handling plug 198. The shield plug 212 effectively seals the penetration to prevent leakage of gases into or out of the reactor vessel 12. The floor valve 200 which comprises the other half of the refueling plug 174 includes a generally cylindrical lower portion 214 which is adapted to fit within the opening defined by bearing supporting skirt 56. Alignment pins 216 are provided on either the floor valve 200 or the handling plug 198 to ensure proper alignment of the two bores 202, 204 through the floor valve and the handling plug. A large ring gear 218 provided around the outer periphery of the floor valve 200 at about mid-elevation includes outwardly extending gear teeth adapted to be engaged by an appropriate drive mechanism for rotating the refueling plug 174 relative to the bearing supporting skirt 172. The drive mechanism, not shown, is similar to that described with reference to the small rotating plugs of U.S. application Ser. No. 537,284. Beneath the ring gear 218, there is provided along the outer surface of the floor valve an outwardly extending flange 220 which is adapted to engage and mate with the inner race of the bearing 188. In this way the floor valve 200 is supported on the bearing for rotational movement about its axis and about the axis of the penetration opening. Sealing means such as O-ring type seals 222 provided along the inner surface of the skirt extension 172 provide a gas tight seal between the extention and the floor valve. The valve portion 224 of the floor valve is positioned above the bearings and ring gear in the upper portion of the floor valve. During refueling, the upper surface 226 of the floor valve 200 interfaces with the refueling machine to form a hermetically sealed passage to transfer fuel assemblies from the reactor vessel. When the valve 224 is closed it provides a barrier of lead shielding for personnel protection. Basically, the floor valve is a heavily shielded gate valve. The valve disc 228 is tapered on its lower face to mate with a similarly tapered valve seat 230 provided in the body 231 of the upper portion of the floor valve. This configuration protects the O-ring lower sealing element 232 from damage when the disc is moved to close or open the passage. The valve body 231 is otherwise a hermetically sealed unit with connections provided for purging the body of the valve with clean argon or other similar gas. Movement of the valve disc 228 is achieved by a ball screw drive arrangement 234 mounted to the side of the floor valve and above the ring gear 218. The disc 228 is mounted on rails and rides on a series of ball bushings 236 and the ball screw drive arrangement 234 provides a motive force. The drive motor 238 is mounted outside of the valve body 231 and drives the ball screw shaft through a coupling 240. The shaft and motor are axially stationary and engage a ball nut 242 attached to the valve disc 228. Upon actuation of the motor the disc is caused to be inserted or retracted with the ball screw shaft 238 extending through a longitudinal cavity within the disc 228 when the valve is fully open. This arrangement is similar to that disclosed for the floor valve of co-pending application Ser. No. 430,292 with the exception that the penetration opening 204 in the floor valve is circular whereas in the co-pending application it was oval or obround. In order to provide the requisite refueling coverage (to be described in more detail hereinbelow) it is necessary that the floor valve 200 which is rotatably supported on the bearings 188, be attached to the handling plug 198 so that the two may rotate together. This is accomplished by a series of screw jacks 244 provided in the upper surface of the floor valve which extend downward therethrough and into appropriate sockets 246 in the handling plugs. Upon actuation or tightening of the screw jacks, the handling plug is raised off the flanged surface 182 provided on the upwardly extending skirt so as to be coupled and mate with the floor valve. In this way the two components of the refueling plug rotate as a unit. As can be appreciated, the bearing supporting skirt extension 172 is necessary with the core arrangement and nozzle arrangement on the reactor vessel head in order to place the bearings 188 in a location to permit rotation of the refueling plugs or whatever plug is provided inside of the skirt. In prior arrangements, enough space was available between adjacent nozzles on the rotating plug and on the reactor vessel head, however, with the extremely small fuel assemblies in the present configuration, this is not possible. Instead it is necessary to remove the drive mechanism and electrical connectors from the upwardly extending nozzles and to utilize the skirt extension so that the bearings will be supported above the nozzles. As can be seen from the drawings, the laterally extending lip 186 which supports the bearings 188 on the skirt extension 172 overlie at least a portion of some of the nozzles 142 which are adjacent to the upwardly extending skirt 56 surrounding the boundary of the penetration. In order to interchange the refueling plug and the plug mounted instrumentation assembly it is necessary to provide a special two-position handling cask 248 similar to that design employed to replace the instrument tree plug and the handling plug of co-pending application Ser. No. 537,283. This is schematically shown in FIGS. 22a through 22c. Initially, the core servicing means 88 are raised out of engagement with the core 18 so as to be free from the core in a manner as described hereinabove; i.e., the plug mounted core servicing apparatus are displaced from their expanded and lowered positions to their condensed and raised positions in which the instrumentation or other servicing means of the various trees are maintained in a stacked relationship completely within and beneath the plug boundaries. For the single core servicing means the control rod drive extension or instrumentation lead is raised a short distance out of engagement with the core and maintained in this raised position through use of appropriate drive extension locks. The auxiliary drives are then removed and the servicing means of both single and plug mounted instrumentation nozzles are maintained in a raised position by means of appropriate drive extensions. Next the two-position cask 248 is used to maintain the reactor cover gas and prevent release of fission products and secondly to shield the contaminated plugs. The handling cask 248 is mounted on the upwardly extending skirt 56 and the lower portion 249 extends upwardly to a distance above the drive extension locks on adjacent nozzles where it then expands to a larger size which is necessary in order to provide two chambers which are large enough to hold the core servicing plug 55 and the handling plug 198. The cask includes a lifting means for liftingly engaging the core servicing plug 55 and for moving it vertically up into one of the chambers of the cask. After the cask is positioned in sealing relationship on the upwardly extending skirt 56 the core servicing plug 55 is lifted into one of the chambers of the cask leaving the penetration 54 in the reactor vessel head 14 open. When the plug is entirely within the cask chamber it is moved laterally out of vertical alignment with the penetration opening and a second chamber of the cask moved into alignment therewith. This second chamber also includes an appropriate lifting and drive means which is engaged to the handling plug portion 198 of the refueling plug 174. The handling plug 198 is then lowered into the penetration opening 54 and allowed to rest on the inwardly protruding flange surfaces 206 of the upwardly extending skirt 56. When the handling plug is placed within the penetration opening, the shield plug 212 with appropriate seals is in place in the access opening 176 or bore therethrough. The handling cask 248 is then removed from the upwardly extending skirt 56; the bearing supporting skirt extension 172 is positioned in place and the floor valve 200 lowered to rest upon the bearings 188. Next, the screw jacks 244 on the upper surface of the floor valve 200 are actuated to engage the handling plug and raise it into engagement with the floor valve 200. While this is being done, the valve disc 228 of the floor valve 200 may be in either the closed or open position since the shield plug 212 is in place in the handling plug. After the floor valve 200 is properly installed, the valve disc 228 is opened and a second handling cask 250 (see FIG. 22c) is placed over the valve opening 204 and actuated to remove the shield plug 212 from the handling plug 198. Following this step the floor valve 200 is closed and the handling cask 250 containing the shield plug 212 is removed to a remote location. The refueling plug system is then in condition for operation. Upon proper rotation of the refueling plugs 174 and the large rotatable cover 42, the access port in the floor valve and handling plug may be placed over any desired core location. The next step in the refueling operation is to position the fuel handling machine, not shown, above the reactor access port by an appropriate means. Once aligned the entire machine is then coupled to the floor valve 200 and the passage between the floor valve and the handling machine purged with argon and checked for leak tightness to ensure that no air is permitted to enter the reactor. Such operations are similar to those described with respect to the fuel handling machine described in co-pending application Ser. No. 430,292. For operation with the floor valve 200 and handling plug 198 having a circular access opening 202, 204, the fuel handling machine of co-pending application Ser. No. 430,292 will have to be modified for proper operation. Such modifications are within the purview of persons skilled in the art and familiar with the description of fuel handling machine of the co-pending application Ser. No. 430,292. Replacement and installation of fuel assemblies may then be accomplished. With each of the plug mounted instrumentation schemes disclosed hereinabove with reference to the core servicing plugs 55 only four refueling plugs 174 are needed in order to adequately provide access directly over each assembly comprising the nuclear core. For the preferred plug mounted instrumentation scheme, shown in FIG. 4, in which the reactor vessel head 14 is concentrically positioned with respect to the nuclear core 18 and in which the array of instrumentation plugs 55 is eccentric with respect to the core in the reactor vessel head, the penetrations into which refueling plugs should be placed in order to provide access to the core over each assembly are depicted in FIG. 23 and are labled plugs A, B, C and D. FIG. 23 is a schematic representation of the reactor vessel head with the core 18 being shown in outline. The circles drawn about the center of the reactor vessel head represent the limit of reach of each of the refueling plugs A, B, C and D and accordingly depict the zones of the core which may be serviced by each of these plugs. That is, they depict the inner and outer boundaries of refueling coverage obtainable with each of the plugs A, B, C and D. As noted above, insertion and withdrawal of fuel assemblies is accomplished with the use of a fuel handling machine which enters through the access ports in the floor valve and the handling plugs to reach into the core and engage or release a fuel assembly. These access ports are indicated on the refueling plugs as circular portions a, b, c, d within the circles A, B, C, D respectively indicating the refueling plugs. The zones of the core served by each of the refueling plugs are indicated as circular or annular spaces and labled zone A, B, C or D respectively. As can be seen in FIG. 23, there is at least a small overlap of the core area which is served by each of the plugs. A similar schematic representation based on the plug arrangement shown in FIG. 18 and again indicating the refueling plugs as A, B, C and D is shown in FIG. 24. As with FIG. 23, each of the zones served by each of the refueling plugs A, B, C and D are labled and indicated as annular or circular rings. This refueling coverage corresponds to an eccentric reactor vessel head and a concentric array of plugs. With either of the refueling schemes depicted in FIGS. 23 and 24, the plug or plug penetrations not including refueling plugs may retain the core servicing plugs in place. The instrument trees attached and supported by these servicing plugs, of course, will necessarily have to be maintained in a raised and compacted position so as not to interfere with the core upon rotation of the large rotatable cover and the refueling plugs, and so as not to interfere with the fuel handling machine upon its insertion into the acccess port in the refueling plugs. SUMMARY Accordingly, there has been disclosed hereinabove a novel core servicing apparatus which is mounted to plugs which in turn are supported within penetrations of the reactor vessel head in order to provide servicing functions to the assemblies comprising the nuclear core of a liquid metal cooled fast breeder reactor. The core servicing apparatus includes a plurality of support columns suspended from a removable plug mounted in the reactor vessel head. Laterally extending rigid support arms are fixed to the support columns and a plurality of core servicing means are supported by and extend downwardly from the lateral support arms. Core servicing means are supported in a fixed array with respect to the support columns. Rotational motion means are provided for rotating and moving vertically the support columns to move the servicing means between condensed and expanded states. When in the condensed state, in the preferred embodiment, the servicing means of the plurality of support columns are maintained in stacked relationship in which the servicing means of one of the columns are supported vertically above and within the co-extensive boundaries of the plug above the servicing means of another of the support columns, the servicing means of all the support columns being maintained within the co-extensive boundaries of the plug. When in the expanded position, the servicing means of the support columns are maintained at the same vertical elevation. Also disclosed herein is a refueling arrangement for a liquid metal cooled fast breeder reactor of a type having a reactor vessel head on which are mounted upwardly extending nozzles in which are supported core servicing apparatus of the nuclear core. Some of the nozzles are mounted on removable stationary plugs. The refueling arrangement comprises a bearing supporting extension skirt and a refueling plug. The extension skirt is mounted upon an upwardly extending skirt surrounding the boundary of the penetration in the reactor vessel head provided for the core servicing plug and serves to support rotational bearings above the elevation of adjacent nozzles on the reactor vessel head. The refueling plug is rotatably supported on the rotational bearings of the bearing support extension skirt and is provided with an access port therethrough for providing refueling access to the core of the nuclear reactor. In the preferred embodiment the refueling plug comprises a handling plug supported within the upwardly extending skirt on the reactor vessel head and a floor valve having its bore in line with the bore of the handling plug and being supported on the rotational supporting bearings. The floor valve is coupled to the handling plug so that the two rotate together and the floor valve is provided with a closure means for sealingly closing the access port therethrough. The embodiments shown and described are merely illustrative of the present invention and changes may be made as well as modifications without departing from the scope of the present invention. What is thought to be protected here and is only that which is set forth in the appended claims.
	abstract	A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cm·K. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.
042959357	abstract	A bimetallic spacer means designed to be cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The subject bimetallic spacer means in accord with one embodiment of the invention includes a member formed, at least principally, of zircaloy to which are attached a plurality of stainless steel strips. The latter stainless steel strips are located on the external surface of the zircaloy member and with the major axis of each of the plurality of stainless steel strips extending substantially perpendicular to the major axis of the zircaloy member. In accord with another embodiment of the invention, the subject bimetallic spacer means includes a member formed at least principally of zircaloy to which a plurality of stainless steel strips are attached so as to be positioned thereon externally thereof and with the major axis of each of the plurality of stainless steel strips extending substantially parallel to the major axis of the zircaloy member. In accord with a further embodiment of the invention, the stainless steel strips are attached to preselected members, each embodying at least a cladding of zircaloy, which are located in the rows of fuel rods that define the perimeter of the fuel matrix of the nuclear fuel assembly. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. Namely, the stainless steel strips expand laterally relative to the fuel assembly and thereby occupy the space adjacent to the external surface of the fuel assembly, into which portions of the fuel assembly would otherwise be receivable should they undergo bowing.
059848537	abstract	A method and apparatus of creating a miniaturized source of radiation and delivering radiation to a location such as therapy location. The radiation source comprises a member made of a material emitting electrons when energy is supplied to the member. There is an electron retarding member disposed opposite the electron emitting member, and the electron retarding member is made of a material emitting ionizing radiation when electrons are retarded therein. The radiation source is further provided on an elongated member in the distal region thereof, and the elongated member is insertable into the body.
	description	The invention relates to a method for processing items, in particular items in a production sequence, in different processing steps. The invention also relates to a device for processing items, by which device the mentioned method in particular can be carried out, and it also relates to the use of said device in a production process. Today, the irradiation of items or even people is used universally, in industry, in trade, in the household, during leisure time or in medicine. In particular, irradiation procedures are used at different points in industrial processes, for example for heating semi-finished products or for activating chemical reactions. To achieve the distribution of radiation required for this purpose, generally a homogeneous distribution, a specific device is required which is configured for a respective specific component, for a specific shape. Human intervention is necessary in order to adapt a device of this type to different components/shapes. When irradiating the respective item or items, a radiation source is usually used which has a defined radiation characteristic. Radiators of this type are usually arranged in fields and are fixed in this arrangement. These fields of radiators are configured by an operator for the respective use. Examples include, for example, infrared radiators for heating semi-finished products, UV radiators for curing paints or resins or infrared radiators for drying surfaces. These devices are usually directed onto the surface to be irradiated, then switched on and switched off again at the end of a defined time or upon reaching a particular measurement result. If relatively complex shapes, such as three-dimensional surfaces with recesses, are to be irradiated uniformly, this has to be considered before the irradiation device is used by adapting the construction or by adjusting an intensity distribution. The reason for this is that the incident radiant power decreases with the square of the distance from the radiation source, i.e. the power has a dependence proportional to 1/r2. In addition, most radiation sources do not have a homogeneous radiation characteristic over the solid angle. Furthermore, different materials absorb electromagnetic radiation at different wavelengths to different extents. Radiators of this type can be connected for example in zones which can be controlled individually. These zones are always at least as big as one radiation source, thus in the above-mentioned case, an infrared lamp. Consequently, irregularities in distribution occur which are potentially undesirable for many processes or which rule out irradiation as a process technology. It is therefore an idea of the present invention to provide a method and a device, which device allows a measurement and subsequent irradiation of items using simple means and in an economical manner. Accordingly, the realization of said idea consists in measuring in the method at least one item, using at least one sensor arrangement having a plurality of sensors, by means of at least a partial quantity of the sensors as a detection means in a detection mode, then determining by a control means the impacting of the item with an irradiation means, while considering a result of the measuring of the item, and finally irradiating the item by at least a further partial quantity of the plurality of sensors as an irradiation means in an irradiation mode; if appropriate, at least part of the method is repeated with renewed measuring, determination and/or irradiation. Equally, the above idea is also achieved by a device for processing items, comprising a detection means having at least one sensor arrangement with a plurality of sensors, at least one item being measured by means of at least a partial quantity of the sensors in a detection mode; a control means determining an impacting of the item by an irradiation means, while considering a result of the measuring of the at least one item, and at least a further partial quantity of the plurality of sensors irradiating the at least one item as the irradiation means in an irradiation mode. Therefore, according to an aspect of the invention, sensors are used to measure the item which is to be subsequently irradiated, and to obtain the dimensions thereof, to computationally calculate the parameters required for an optimum irradiation from the result of this determination and subsequently to irradiate the item in question as desired by the radiation sources which were previously used as sensors. Thus, the invention is essentially based on the fact that radiation sources are used which are capable of acting not only as radiation sources but also as detectors. Another embodiment of the method according to the invention and of the device in which the sensors can be used and can be connected as a detector and as a radiation source, which can be effectively controlled and provides a good luminous efficiency, can consist in configuring each of the sensors of the sensor arrangement by a light-emitting diode (LED). This approach is based on firstly charging the LEDs against their transmitting direction and then measuring the time which they require for a full discharge. This time depends on how much light energy is incident thereon. This type of measurement can already be implemented using minute circuits and it provides an excellent signal-to-noise ratio (S/N). However, an LED can generally only detect radiation which lies within a very narrow range around the wavelength which it itself emits. Further developments which allow a good handling using the sensors of the sensor arrangement and which make the use thereof flexible can consist in the fact that, in the case of developments of the method and device according to an embodiment of the invention, the sensors of the arrangement can be activated or are activated individually, combined severally together and combined into groups. Here, the sensors of the sensor arrangement can form, for example, an array of sensors, for example an array of this type with individual sensors, with partial quantities of sensors or collectively can have a regular arrangement over a surface to be irradiated. Thus, for example the device can comprise a field with LEDs which are interconnected such that they are capable of measuring an incident luminous flux and of actively operating as radiation sources. Fields of this type can form, for example, flat LED modules which are capable of heating relatively large components. For this purpose, the LEDs are mounted for example in a close arrangement on a board. To irradiate a surface, the LEDs can then be activated individually and the intensity thereof can be controlled. In the simplest case, the surface to be irradiated is approximately parallel to the LED board. Due to the arrangement of the LEDs, with the same radiation intensity of all the LEDs, a bell curve is obtained for the radiation impacting the surface, over the width of the board, with a maximum in the centre and with less radiation in the peripheral region. The same applies in the second dimension. Therefore, in order to achieve a homogeneous radiation on the surface, the peripheral regions must radiate more intensively than the centre. A distribution of this type usually has to be calculated in a simulation model and then converted into a program by an operator. At this point, the device starts to itself determine a distribution of this type. For this purpose, for example the LEDs are activated in specific patterns, such as a chessboard arrangement. In this respect, for example one half of the LEDs are operated in the detector mode, while the other half operates as a radiation source. The radiating LEDs emit radiation which is reflected by the surface of the component and is measured by the LEDs which are in the detector mode. For the example of a flat plate, the detector LEDs in the centre would detect significantly more radiation if all the radiating LEDs radiate with the same intensity. The measured values can be saved in a program. The chessboard pattern can then be reversed and the measurement can be completed in this way. Accordingly, further variants of the method according to the invention consist in connecting or activating the sensors of the arrangement differently such that a partial quantity of sensors is operated in a detector mode as the detector means and a partial quantity of the sensors is operated in the irradiation mode as the irradiation means. The mentioned operating modes of the respective partial quantities of sensors can change after a predeterminable or predetermined period of time, and the partial quantities of sensors of the respective mode can form a pattern, for example a regular pattern, for example the above-mentioned chessboard pattern. Here, the mentioned chessboard pattern is not to be understood as a fixed arrangement, but merely as an example. Theoretically, all possible arrangements and combinations can be implemented. These then have to be calculated by a computer model in which all the information regarding the radiation sources, for example the radiation characteristic thereof (the intensity over the solid angle etc.) and other geometric information is stored. As a result, it is possible for significantly more precise distributions to be measured. In a variant of the method at least a partial quantity of sensors is activated by the control means such that a surface of the item to be irradiated is exposed to a radiation intensity which is uniform over the surface. To rule out environmental influences during the measuring procedure of the at least one item or of a plurality of items, a further expedient variant of the method according to the invention can provide that a calibration measurement is optionally carried out before the at least one item is measured, in which calibration measurement the sensors which are used, for example all the sensors, are in the detection mode. As a result, the influence of the ambient lighting is eliminated for example in the subsequent calculation. Furthermore, a variant is also conceivable in which disjoint partial quantities of sensors are operated simultaneously as the detector means and as the irradiation means. To be able to appropriately determine the required type of irradiation of the respective item which does not necessarily have to be a homogeneous irradiation (although it frequently is), another variant of the method according to the invention can consist in determining an irradiation, to be carried out, of the item using sensors in an irradiation mode by an iterative process, in the steps of which an intensity variable of the irradiation treatment is measured once or several times by at least a partial quantity of sensors in the detection mode and is changed by the control means, or the control means analytically calculates and fixes the intensity variable of the irradiation treatment using sensors in an irradiation mode. Thus, two courses of action are possible: a computer program can adapt the intensity of the radiating LEDs in an iterative process until a distribution of the detector LEDs, which is for example as homogeneous as possible (or another desired distribution), is measured. For this purpose, the radiation intensity would be slightly altered, a renewed measurement would be made using a double chessboard pattern and then the measured values would be checked for homogeneity. This process can be continued until a defined criterion is reached or until similar abort criteria, used in iterative optimisation processes, are present. The result is an intensity distribution of the two overlaid patterns which provide an intensity distribution for the actual irradiation. A distribution can also be calculated analytically by means of a computer program. The only difference from the first variant is that here, the iterative procedure is not used. This variant requires a significantly more complex program which reproduces more complex mathematical connections. Thereafter, the irradiation can be carried out with the calculated radiation distribution. The irradiation can be influenced by particular parameters, such as a surface temperature which is to be attained. This would only be considered as a multiplier for the calculated intensity distribution. This would then be, for example, the only information to be provided by the operator. Conventional irradiation means require for a defined irradiation that the item to be irradiated is positioned in a particular location within the field of radiators so that radiation will be carried out in the precise actual location of the item. If this is not observed, some of the radiation is lost and a part of the item is not irradiated as intended. Due to the detection at the start of the process, it is possible to determine where one or more items are actually located, i.e. which region has to be irradiated. If an object is not in a location, thus changed, this can be seen from the measuring signal, and this region can be omitted. In this way, a small item can also be irradiated using a large array of LED radiators as sensors, without the necessity of adapting the circuitry or the like. This can also be transferred to a use in which an item to be irradiated is located on a different surface, for example a semi-finished product in a tool. Here, a large number of different semi-finished product/tool combinations can be optimally irradiated using a single LED irradiation means, without a program having to be created for this purpose. Likewise, new items can be irradiated without being previously set up. In continuous processes, items such as foodstuffs, component parts or semi-finished products are moved past irradiation means and are irradiated. In this respect, a plurality of items usually run next to one another or the semi-finished products have available a particular width. Both can change during the process, between batches or also generally, for example in order to operate production capacities in a continuously optimum manner. As a result, regions in the process are also often irradiated which are empty at this time because no item is present. The device according to the invention and the method can suitably compensate for this in that, during the start of the process, the place is detected where items which need to be irradiated are actually located. This detection can also be carried out during the process and does not require any human intervention. Consequently, it is possible to reduce costs in terms of energy and staff. Inter alia, for such scenarios, in an embodiment of the device according to the invention each of the sensors of the sensor arrangement can be respectively controlled at least in respect of its radiation intensity. It is of particular importance that, by means of the device according to the invention, together with the method, it is possible to homogeneously irradiate three-dimensional structures, more specifically without a high configuration, calculation or operating expense. Due to the measuring procedure, it is possible to characterise the reflection behaviour of the item and, in this way, to calculate an intensity distribution which fulfils the required parameters, such as a homogeneous irradiation. The measured values, recorded by the LEDs in the detector mode are directly related to the distance of the surface from the radiation sources. Furthermore, angles in the surface also have an influence on the measured result. By different arrangements of the sensors/radiation sources, these can also be detected, so that an expedient variant of the method according to the invention can consist in also detecting profiled, non-planar structures of the item in the measuring procedure by different arrangement patterns and/or connection patterns of the sensor arrangement and/or relative positions of the detection means and of the item to be measured, and in subsequently irradiating them, if appropriate, after a corresponding calculation. There are various possible uses for this, because surfaces which are curved, graduated, interrupted or combinations of the mentioned characteristics can be irradiated as required in this manner. For this, developments of the invention can be expedient in which the at least one item and the sensor arrangement of the plurality of sensors are provided such that they are movable relative to one another by at least one adjusting device. Furthermore, for example the at least one item and at least a partial quantity of sensors of the sensor arrangement can form mutually parallel surface portions in the position of use. For this purpose, it could be imagined, for example, that a plurality of operating cycles of the method is carried out on a complex shape of an item, in which operating cycles another surface, which is respectively positioned parallel to a sensor arrangement by whichever's relative movement, is measured, the calculation of an optionally previously performed irradiation is also considered and is then irradiated (again). Using the device according to an embodiment of the invention and the method, it is also possible to consider other aspects of the irradiation procedure in the form of the surface, in other words of the irradiated material. The reflection characteristics of an item not only depend on its shape, but also on the material. In this respect, the nature of the surface (rough/smooth), the colour or the structure can play a part. In the device according to an embodiment of the invention, the sensors of the sensor arrangement can be formed by diodes of different colours. Here, the different colours of the sensors can cover a wavelength range of a few 100 nm, for example extending between the UV range and the NIR range. While LEDs usually only operate as radiation sources/sensors within a very narrow frequency band, a plurality of colours can also be displayed by combinations of different-coloured LEDs. Furthermore, LEDs are also available in the UV and NIR ranges. While a partial quantity of LED sensors of one colour radiates, another partial quantity of sensors records the reflected or transmitted radiation (for example in the case of films). Due to the combination of the different sensor/radiation sources, broad spectra of the UV-NIR range can be covered even with few wavelengths. The information which is obtained corresponds to the results of a large-surface spectroscope, using which it is possible to measure complete components instead of individual small samples. Uses for this are initially material characterisations, contamination detection or the like. The use of a device according to an embodiment of the invention as well as the use of the method is particularly advantageous for processing items in a production sequence. The above embodiments and developments can be combined together in any meaningful manner. Further possible embodiments, developments and implementations of the invention also include combinations, not explicitly mentioned, of features of the invention described previously or in the following in respect of the embodiments. In particular, a person skilled in the art will also add individual aspects as improvements or supplements to the respective basic form of the present invention. In all the figures, identical or functionally identical elements and devices have been provided with the same reference numerals, unless indicated otherwise. FIGS. 1 to 4 show a device, denoted in its entirety by reference numeral 10, for processing items 35, having a detection means 20 with at least one sensor arrangement 15 having a plurality of sensors 16, at least one item 35 being measured by at least a partial quantity of the sensors 16 in the detection mode; a control means (not shown) determining an impacting, to be carried out, of the item 35 by an irradiation means 30, while considering a result of the measuring procedure of the at least one item 35, and at least a further partial quantity of the plurality of sensors 16 irradiating the at least one item 35 as the irradiation means 30 in an irradiation mode. In this respect, FIG. 1 shows an embodiment of the device 10, in which the sensor arrangement 15 consists of an array of sensors 16 formed from LEDs which are interconnected such that they are capable of measuring an incident luminous flux and they are also capable of actively operating as radiation sources. In the latter type of operation, they are capable, for example, of heating relatively large components as items 35. For this purpose, the sensors 16, which are shown, of the sensor arrangement 15 as LEDs are mounted in a close arrangement on a board 17, which is square in this case. This board is shown by way of example in the drawing, the individual LEDs are identified by indices. In order to irradiate a surface, the sensors 16/LEDs can be activated individually and the intensity thereof can be controlled. In the simplest case, the surface to be irradiated is arranged parallel to the LED board 17. FIG. 2 shows the sensor arrangement 15 of the device 10 which, with its sensors 16 configured as LEDs, irradiates an item 35 in the form of a surface 36. Due to the arrangement of the sensors 16, with the same radiation intensity of all the LEDs, a bell curve is produced for the radiation incident on the surface of the item 35, over the width of the drawing shown above, with a maximum in the centre and with less radiation in the peripheral region of the item 35. The same applies in the second dimension which extends into the viewing plane, transversely thereto. In order to achieve a homogeneous irradiation on the surface of the item 35, the peripheral regions of the array of sensors 16 must therefore radiate more intensively than the centre thereof. A distribution of this type is calculated in a simulation model of the control means (not shown) and is subsequently converted into a program. To be able to satisfactorily achieve a conversion of this type, the corresponding distribution itself is determined by the device according to the invention, as can be clearly seen from FIG. 3. In this figure, the sensors 16 are activated in specific patterns, for example as a chessboard arrangement, on the board 17, which here again is square. In this respect, one half of the LEDs is operated in detection mode, while the other half operates as a radiation source, which is shown by the “light” sensors 16 as sensors 16 in the irradiation mode and by the “dark” sensors 16 in the detection mode. In this respect, the radiating LEDs emit radiation which is reflected by the surface of the item 35 and is measured by the LEDs which are in the detection mode. Taking the example of the planar board, as can be seen for example as the item 35 in FIG. 2, the detector LEDs as sensors 16 would detect significantly more radiation in the centre if all the radiating LEDs radiate with the same intensity. The measured values can be saved in a program. Thereafter, the chessboard pattern can be reversed and in this way the measurement can be completed. After this reversal, the sensors 16 which were previously respectively operating in the detector mode then operate as the radiation source in the irradiation mode, and vice versa. The sensors 16 of the sensor arrangement 15 thereby respectively form partial quantities of sensors 16 which, in the present example, are in fact disjoint, but this does not necessarily have to be the case. Two courses of action are then possible with the device 10: a) In an iterative process, a computer program can adjust the intensity of the radiating LEDs until a distribution of the detector LEDs is measured which is as homogeneous as possible. For this purpose, the radiation intensity would be changed slightly, a new measurement would be carried out using a double chessboard pattern and then the measured values would be checked for homogeneity. This process can be continued until a defined criterion is reached or until similar abort criteria are present which are used in iterative optimisation processes. The result is an intensity distribution of the two overlaid patterns which provide an intensity distribution for the actual irradiation. b) A distribution can be calculated analytically also by means of a computer program. The only difference from the first variant is that here, the iterative procedure is not used. This variant requires a significantly more complex program which represents more complex mathematical connections. Thereafter, the irradiation can take place by the sensors 16 as LEDs with the calculated radiation distribution. Finally, FIG. 4 shows the possibility of homogeneously irradiating three-dimensional structures as the item 35 by the device 10 according to the invention, more specifically without a considerable configuration, calculation or operating expense, FIG. 4 also showing the problem which usually occurs in this respect. If the surface of an item 35 is irradiated by a radiation source, the irradiance E on this surface depends on the radiation intensity and on the distance from the radiation source. The radiation intensity I is constant over the solid angle (the radiation characteristic of the radiation source is not considered for this), so that the irradiance only depends on the distance from the radiation source, since the surface irradiated over a given sold angle increases quadratically with the distance r from the source. Accordingly, the connection can be formulated by E=I/r2. cos(·): the cosine considers the angle between incident radiation and the surface normal, i.e. the projected surface. With an appropriate intensity distribution of the radiators, a homogeneous distribution can, however, also be achieved in the case of three-dimensional surfaces of an item 35. As already described in an exemplary manner by the embodiments of FIGS. 2 and 3, for this purpose, the centre radiators would again have to radiate less intensively than the outer radiators. It is helpful in this respect for the sensors 16 of the sensor arrangement 15 to be as small as possible as radiation sources, i.e. for the realisable grid to be as fine as possible. This is accommodated by the approach of using LEDs as sensors 16/radiation sources. In this regard, FIG. 4 shows an item 35 which is irradiated by sensors 16, which is shown schematically by the regular cones. Due to the distance from the source, a differing intensity distribution is produced on the surface of the item 35, which is indicated by the area 36, coloured in different “thicknesses”, above the surface of the item 35. Calculating an intensity distribution of this type requires a precise knowledge of the shape of the body, and also of the position in the radiation field produced by the radiating sensors 16. The expense for this is not small and requires simulation and calculation software. A distribution, calculated in this way, is then only ever valid for a particular shape of the item 35 in a defined position and has to be re-calculated for every modification. Here, the device 10 according to the invention affords tremendous opportunities for improvements. As a result of the measuring procedure, the reflection behaviour of the item 35 can be characterised and thus an intensity distribution can be calculated which fulfils the required parameters, such as a homogeneous irradiation. The measured values recorded by the sensors 16 in the detector mode are directly connected to the distance of the surface of the item 35 from the sensors 16 used as radiation sources. Furthermore, angles in the surface of the item 35 also have an influence on the measured result. These can also be detected via different arrangements of the sensors 16. The possibilities of use for this are varied, because curved, graduated, interrupted surfaces or combinations of the mentioned characteristics can be irradiated as desired in this way. Thus, the invention includes a device 10 and a method for measuring, characterising and irradiating objects 35, for example in production sequences. An aspect of the invention is based on the use of LEDs in equal measure as sensors and as radiation sources. In a first operating mode, an LED array is used to investigate an item 35. In the second step, calculated from the information which is obtained and also from parameters to be defined by the operator is an intensity distribution which, in the third step, is achieved by the LEDs. Examples of use include, for example, heating geometrically complex items or irradiating particular materials, for example painted surfaces, in order to trigger chemical processes. Although the present invention has been described above on the basis of various embodiments, it is not restricted thereto, but can be modified in many different ways. In particular, the invention can be altered or modified in various ways, without departing from the essence of the invention. While at least one exemplary embodiment of the present invention(s) is disclosed herein, it should be understood that modifications, substitutions and alternatives may be apparent to one of ordinary skill in the art and can be made without departing from the scope of this disclosure. This disclosure is intended to cover any adaptations or variations of the exemplary embodiment(s). In addition, in this disclosure, the terms “comprise” or “comprising” do not exclude other elements or steps, the terms “a” or “one” do not exclude a plural number, and the term “or” means either or both. Furthermore, characteristics or steps which have been described may also be used in combination with other characteristics or steps and in any order unless the disclosure or context suggests otherwise. This disclosure hereby incorporates by reference the complete disclosure of any patent or application from which it claims benefit or priority.
	summary
047568739	claims	1. A single loop nuclear power plant for generation of electricity comprising: a lower pressure vessel; an upper pressure vessel releasably connected to and sealed in a gas tight fashion from said lower pressure vessel; a high temperature gas cooled reactor with an upward coolant flow path and spherical fuel elements, located in said lower pressure vessel; a turbine aligned in said upward coolant flow path above said reactor in said upper pressure vessel; a recuperator connected to said reactor and said turbine in said upper pressure vessel; a radiator connected to said recuperator in said upper pressure vessel aligned above said turbine; a low pressure compressor connected to and aligned above said radiator in said upper pressure vessel; an intermediate radiator connected to and aligned above said low pressure compressor in said upper pressure vessel; a high pressure compressor connected to said recuperator and said intermediate radiator aligned above said intermediate radiator in said upper pressure vessel; a generator aligned above said high pressure compressor; wherein said lower pressure vessel is charged with primary gas and said upper pressure vessel is filled with a protective gas. a container located above said upper pressure vessel, housing said generator. 2. A plant according to claim 1, wherein said generator is located within said upper pressure vessel. 3. A plant according to claim 1, further comprising: 4. A plant according to claim 1, further comprising dry bearings supporting said turbine, high pressure compressor and low pressure compressor. 5. A plant according to claim 1, further comprising magnetic bearing supporting said turbine, high pressure compressor, and low pressure compressor. 6. A plant according to claim 1, wherein said generator is a high speed no topping gear generator exhibiting magnetic bearings. 7. A plant according to claim 1, wherein said protective gas is helium. 8. A plant according to claim 1, wherein said protective gas is nitrogen. 9. A plant according to claim 1, wherein said radiators exhibit sufficient capacity for natural convection decay heat removal. 10. A plant according to claim 1, further comprising a shaft linking said turbine, low pressure compressor, high pressure compressor and generator.
052689549	abstract	A mounting mechanism for a double crystal monochromator or the like has a parallelogram based mounting mechanism in which two of the vertices of the parallelogram are fixed in position, and two vertices are free to translate back and forth in a straight line parallel to the fixed base of the parallelogram. One diffractor is mounting for pivoting at one of the fixed vertices, and the second diffractor is mounted for pivoting at an adjacent movable vertex. The surfaces of the diffractors are maintained parallel as the angle of the diffractors with respect to input and output beams to the monochromator is changed to change the wavelength being passed. The diffractor mounted at the fixed pivot may be connected to a large diameter wheel which in turn is connected by a band to a smaller diameter wheel mounted for rotation at the other fixed vertex of the parallelogram, with a pivotable arm connected to the smaller wheel to rotate therewith. Another larger diameter wheel is connected to the diffractor at the movable vertex and is connected by a band to a smaller diameter wheel at the adjacent movable vertex of the parallelogram, where the smaller wheel is connected by a slider to the pivotable arm. As the movable diffractor is translated in position, the arm pivots to cause the small wheels to move through the same angle of angular displacement as the pivotable arm. The corresponding angular displacement of the diffractors may be one half the angular displacement of the pivotable arm where the larger wheels are twice the diameter of the smaller wheels.
	summary
048636800	description	DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS Now, the embodiment of the present invention will be described below with reference to the accompanying drawings. FIG. 1 illustrates a typical fuel assembly as one embodiment of the present invention. In a fuel assembly 11 of this embodiment, small units 13 are formed each by arranging fuel rods 12 in a square matrix of 3 rows and 3 columns. A total of eight such small units 13 are contained in a channel box 14 of a square cross section. In the central part of the channel box 14, a water rod 15 having an outside diameter equaling one side of the square of the small unit 13 is disposed. The intercentral distance P.sub.2 between two juxtaposed fuel rods 12 belonging one each to two adjacent small units 13 is so larger than the intercentral distance P.sub.1 between two juxtaposed fuel rods 12 belonging to one and the same small unit 13 as to satisfy the relation, P.sub.2 =1.5.times.P.sub.1 for example. The outer diameter of the fuel rod 12 is about 11 mm, for example and the inside diameter of the water rod 15 is about 42 mm. A multiplicity of such fuel assemblies 11 of the construction of the present embodiment described above are arranged in a boiling-water reactor. In the water gap regions intervening between adjacent channel boxes 14, light water which is not boiling is passed. In the water gap regions, a cruciform control rod 16 is inserted. FIG. 2 compares the fuel assembly 11 of the present embodiment with a conventional fuel assembly. As illustrated in FIG. 2 (a), the conventional fuel assembly has all the fuel rods 3 arranged as spaced by a fixed distance P.sub.0. In the fuel assembly 11, the aforementioned intercentral distance P.sub.1 is smaller than P.sub.0 and the aforementioned intercentral distance P.sub.2 is larger than P.sub.0. Thus, in the fuel assembly 11, the resonance escape probability is larger than in the conventional fuel assembly 1 because the mutual effect of the fuel rods 12 in shielding the resonance energy neutrons is larger in the former than the latter. Then, the thermal neutron utilization factor is smaller in the fuel assembly 11 than in the fuel assembly 1 because the thermal neutron flux is high along the boundary between the adjacent small units 13 and low within the small units 13. This decrease in the thermal neutron utilization factor increases in proportion as the ratio between the width of an internal gap to be formed between the adjacent small units 13, namely the distance W indicated in FIG. 2 (b), to the distance of diffusion of thermal neutrons increases. The diffusion length of thermal neutrons increases with the decreasing density of water and decreases with the increasing density of water. The decrease of the aforementioned thermal neutron utilization factor is small during the course of operation and large during the period of cold state. The increase of the resonance escape probability is larger during the course of operation than during the period of cold state because the mutual shielding effect increases in proportion as the density of water decreases. Owing to the effects described above, the ratio increase in the resonance escape probability is amply larger than the decrease in the thermal neutron utilization factor during the course of operation and the decrease in the thermal neutron utilization factor in larger than the ratio of the increase in the resonance escape probability during the period of cold state where small units 13 of the scale of 3 rows and 3 columns to 5 rows and 5 columns are used in the boiling-water reactor, for example. The graphs of FIG. 3 and FIG. 4 show relative changes of resonance escape probability p, thermal neutron utilization factor f, and the product thereof p.times.f, respectively during the course of operation and during the period of cold state, with the horizontal axis taken as the scale of the difference between P.sub.2 and P.sub.1. Since the density of water is small and the diffusion length of thermal neutrons is large during the course of operation, the thermal neutron utilization factor f declines slightly as indicated by the dotted line in FIG. 3 as the difference between P.sub.2 and P.sub.1 widens. During the period of cold state, since the density of water is large and the diffusion length of thermal neutrons is small, the thermal utilization factor f sharply decreases as indicated by the dotted line in FIG. 4 as compared with the change registered during the course of operation. In contrast, as indicated by the alternate one long one short dash line in FIG. 3 and FIG. 4, the resonance escape probability p increases with the widening difference between P.sub.2 and P.sub.1 both during the course of operation and during the period of cold state. The fact remains, however, that the probability of the resonance energy neutrons being moderated by water during their travel between the fuel rods increases with the increasing density of water. The change in the resonance escape probability, therefore, is slightly smaller during the period of cold state than during the course of operation because the mutual shielding effect manifested on the resonance energy neutrons by the fuel rods decreases in proportion as the density of water increases. The effective multiplication factor of the core is directly proportional to the produce of p.times.f as described above. Further, as indicated by the solid line in FIG. 3, the product of p.times.f increases with the widening difference between P.sub.2 and P.sub.1 during the course of operation. It follows that the effective multiplication factor can be increased by widening the difference between P.sub.2 and P.sub.1. If the difference between P.sub.2 and P.sub.1 is increased beyond the range shown in FIG. 3, however, the decrease of f increases and the effective multiplication factor begins to decline even during the course of operation. It is indicated by the solid line in FIG. 4 that the product of p.times.f is decreased by widening the difference between P.sub.2 and P.sub.1 during the period of cold state as observed with respect to the difference between P.sub.2 and P.sub.1 during the course of operation. This means that the effective multiplication factor decreases with the widening difference between P.sub.2 and P.sub.1. Unlike the conventional fuel assembly 1 which has all the fuel rods 3 arranged as spaced with a fixed distance, therefore, the fuel assembly 11 of the present embodiment enables the economy to be improved by heightening the effective multiplication factor of the core during the course of operation and the shut down margin to be increased by lowering the effective multiplication factor during the period of cold state. In the light-water reactor using cruciform control rods, water gap regions having a width approximately in the range of 10 to 20 mm are formed outside channel boxes. As a result, the thermal neutron flux distribution in the horizontal direction increases towards the periphery of the fuel assembly 11 and decreases towards the center thereof. The thermal neutron utilization factor can be further improved and the economy enhanced by flattering this thermal neutron flux distribution. To meet the requirement for extension of burnup, efforts must be paid also to the improvement in this respect. The present embodiment contemplates disposing in the central part of the fuel assembly the water rod 15 having a large diameter equaling one side of the square matrix of 3 rows and 3 columns of fuel rods 12 as illustrated in FIG. 1. In addition to the water inside the water rod 15, the region of water surrounding the water rod 15 is increased by the small units 13. Owing to the water rod 15 and the region of water intervening between the water rod 15 and the small units 13, the thermal neutron flux is heightened in the central part of the fuel assembly 11 so much as to flatten the thermal neutron flux distribution. As a result, the fuel assembly of the present embodiment enjoys a large thermal neutron utilization factor and excels in economy as compared with the conventional fuel assembly. In the fuel assembly 11 described above, the coolant flow path surrounding the fuel rod located at the center of each of the small units 13 is narrow as compared with that surrounding any of the other fuel rods 12 of the same small unit. There is, therefore, the possibility that the thermal restriction imposed upon the fuel rod 12 located at the center of the small unit 13 will be rigidified. To the problem of this nature, a fuel assembly 11a illustrated in FIG. 5 proves to offer an effective solution. In this fuel assembly 11a, the fuel rod 12 at the center of each of the small units 13 is replaced with a fuel rod 20 having a partial length in the direction of axis thereof. This fuel rod 20 may be in the form of a partial length fuel rod 20a having a shortened upper portion, a partial length fuel rod 20b having a shortened lower portion, or a partial length fuel rod 20c having shortened upper and lower portions as illustrated in FIG. 6. The adoption of the partial length fuel rods 20b or the partial length fuel rods 20c in the present embodiment can be realized by the method of retaining fuel rods by means of a spacer which has already found adoption in the existing PWR's. This spacer (consisting of several vertically separated component members) is capable of retaining all the fuel rods contained in a total of eight small units. Otherwise, separate spacers may be used one each for the small units. The problem that the thermal restriction imposed on the fuel rod located at the center of each of the small unit 13 may be otherwise coped with effectively by using fuel rods smaller in diameter than the other fuel rods 12 in the place of the aforementioned partial length fuel rods 20. Alternatively, this problem may be solved by using water rods 30 in the place of all or some of the central fuel rods 12 in the small units 13 as in a fuel assembly 11b illustrated in FIG. 7, by omitting the insertion of the central fuel rods 12 of the small units 13 as in a fuel assembly 11c illustrated in FIG. 8, or using fuel members containing such burnable poisons as gadolinium oxide (Gd.sub.2 O.sub.3) in the place of all or some of the central fuel rods 12 of the small units 13. Incidentally, in the fuel assembly 11 illustrated in FIG. 1, the thermal neutron flux increases towards the periphery of each of the small units 13 and decreases towards the center thereof. It is, therefore, proper as in the case of a fuel assembly 11d illustrated in FIG. 9, to heighten relatively the enrichment of the peripheral fuel rods 12 of the small units 13 for the purpose of extending the burnup and lower the enrichment of the central fuel rods 12 of the small units for the purpose of alleviating the thermal restriction. In FIG. 9, the symbols h, H, M, L, and l denote fuel rods using a fissile substance in enrichment sequentially decreased in the order mentioned. The difference of enrichment between the fuel rods denoted by the symbols h and H and that between the fuel rods denoted by the symbols L and l are smaller than that between the other fuel rods. Then, in a fuel assembly illustrated in FIG. 10, partial length fuel rods 20a having a shortened upper portion are used as fuel rods to be positioned one each at the four corners of the assembly. This modification aims to preclude the possibility that the flux of the coolant (flow volume per unit area) will be smaller in the subchannels (sub-flow paths for the coolant) near the corners of the fuel assembly than the average flux in the overall flow path of the fuel assembly and the critical heat flux or the critical power ratio, one of the magnitudes of thermal restriction imposed on the light-water reactor, will be lowered in the subchannels near the corners. To be specific, the disposition of such partial length fuel rods 20a at the corners of the fuel assembly brings about the following effect upon the critical heat flux (critical power). Firstly, the pressure loss in the subchannels surrounding the partial length fuel rods 20a is decreased and the flux of the coolant in the subchannels is consequently increased and the critical heat flux (critical power) is proportionately increased. Particularly in the boiling-water reactor, the use of partial length fuel rods 20a having a shortened upper portion is effective because the void ratio is large in the upper part (downstream side) of the core and the two-phase flow pressure loss coefficient is consequently large in the upper part of the core. Secondly, in the boiling-water reactor, the critical heat flux is lowered in the upper part of the core because the void ratio is increased in the upper part (downstream side) of the core. Thus, the critical power can be increased by disposing the partial length fuel rods 20a having a shortened upper part as closely juxtaposed to the subchannels of lower critical heat flux thereby alleviating the thermal load upon the subchannels in the upper part of the core. The question as to which subchannels are subject to the most rigorous critical heat flux (critical power) depends, to be exact, upon the pitch of separation of the fuel rods, the distance between the surface of the outermost fuel rods 12 and the surface of the channel box 14, for example. When the fuel assembly constructed as illustrated in FIG. 10 is considered with respect to the sizes of the BWR fuel assemblies, it is most effective to use the partial length fuel rods 20a at the corners or at the (2, 2) positions from the corders. The replacement of at least one of the four corners or near-corner fuel rods with the partial length fuel rod 20 is also effective in improving the critical heat flux (critical power). In this case, it is advantageous to use as the partial length fuel rod 20 the aforementioned partial length fuel rod 20a having a shortened upper portion as compared with the whole length fuel rod 12. The use of the aforementioned partial length fuel rod 20b having a shortened lower portion, however, proves to be effective in decreasing the pressure loss and increasing the flux in the surrounding subchannels. Likewise, the use of the aforementioned partial length fuel rod 20c having shortened upper and lower portion is effective in improving the critical heat flux or the critical power. In the case of the aforementioned partial length fuel rod 20b and partial length fuel rod 20c, the fuel assembly is constructed so as to retain these partial length fuel rods by means of an upper tie plate or a spacer. Since these partial length fuel rods 20b and partial length fuel rods 20c are not required to be supported by the lower tie plate, the pressure loss in the portions of the lower tie plate directly below the partial length fuel rods can be decreased. By this effect, the critical heat flux or the critical power can be proportionately improved. The length of the partial length fuel rods is desired to be in the range of 1/2 to 7/8 of the length of the whole length fuel rods. A fuel assembly 11f illustrated in FIG. 11 represents an embodiment wherein four small units 13a, 13b, 13c, and 13d formed by enlarging the small units 13 of the fuel assembly 11e illustrated in FIG. 10 so as to acquire mutually different shapes are arranged. In this fuel assembly, partial length fuel rods 20a are disposed one each at three positions near each of the corners of the fuel assembly. A fuel assembly 11g illustrated in FIG. 12 has a construction wherein four small units 13e, 13f, 13g, and 13h of mutually different shapes are arranged and a water rod 15b having a diameter equalling one side of a square containing four fuel rods 12 is disposed approximately at the center of the fuel assembly. In this fuel assembly, four partial length fuel rods 20a are disposed in the small unit 13e of the largest size and two partial length fuel rods 20a are disposed in each of the other small units 13f, 13g, and 13h. The embodiments have been described each as forming a fuel assembly composed of 9 rows and 9 columns of fuel rods. They do not prevent the present invention from being embodied in fuel assemblies of other constructions. A fuel assembly 41 illustrated in FIG. 13 is composed of 8 rows and 8 columns of fuel rods 42. Within a channel box 44 having a square cross section, four small units 42 each formed of 4 rows and 4 columns (minus one fuel rod) of fuel rods 42 are arranged and a water rod 45 is disposed at the center of the fuel assembly. In this embodiment, similarly to the embodiments described above, the intercentral distance P.sub.2 is larger than the intercentral distance P.sub.1 A fuel assemble 41a illustrated in FIG. 14 represents an embodiment wherein partial length fuel rods 20a are disposed one each in one inner row and one inner column relative to the four corners of the fuel assembly 41. A fuel assembly 51 illustrated in FIG. 15 represents an embodiment of the present invention in a BWR fuel assembly having a larger size than the conventional countertype. In this fuel assembly 51, 14 small units 53 each composed of 3 rows and 3 columns of fuel rods 52 are disposed inside a channel box 54 and two water rods 55 are disposed approximately at the center. In the case of this fuel assembly 51, since the channel box 54 uses side walls of a large length, richly perforated structural members 56 disposed to interconnect the opposed side walls bring about a notable effect in diminishing possible bulging of the channel box 54. These structural members 56 need not be installed throughout the entire axial length of the channel box but may be disposed as separated into several pieces over part of the axial direction. To decrease the amount of neutrons to be absorbed by the structural members 56, the structural members 56 are desired to possess a large number of through holes. A fuel assembly 51a illustrated in FIG. 16 represents an embodiment wherein three partial length fuel rods 20a are disposed at each of the four corners of the fuel assembly 51. Now, a fuel assembly so constructed that the distance (width of internal gap) between the adjacent small units 13 is varied within the fuel assembly will be described. A fuel assembly 11h illustrated in FIG. 17 represents an embodiment wherein the intercentral distances P.sub.2 and P.sub.3 between two juxtaposed fuel rods 12 belonging one each to two adjacent small units 13 are so larger than the intercentral distance P.sub.1 between two juxtaposed fuel rods 12 belonging to one and the same small unit 13 as to satisfy the relation, P.sub.3 =1.7.times.P.sub.1 and P.sub.2 =1.3 P.sub.1, for example. FIG. 18 compares this fuel assembly 11h and the conventional fuel assembly 1. Unlike the conventional fuel assembly 1 having all the fuel rods 3 disposed as spaced with a fixed intercentral distance P.sub.0 as illustrated in FIG. 18 (a), this fuel assembly 11h has the fuel rods so spaced that the intercentral distances P.sub.2 and P.sub.3 are larger than the distance P.sub.0 and the intercentral distance P.sub.1 is smaller than P.sub.0. In the fuel assembly 11h of such a construction as described above, the ratio of the decline of the thermal neutron utilization factor increases in proportion as the distance between the adjacent small units 13, namely the ratio of the widths W.sub.1, W.sub.2 of the internal gaps illustrated in FIG. 18 (b) to the distance of diffusion of thermal neutrons, increases. The graph of FIG. 19 shows the relative change of the multiplication factor with the sum of W.sub.1 +W.sub.2 during the period of cold state, with the ratio of W.sub.2 to W.sub.1 as a parameter. The decrease of the thermal neutron utilization factor exponentially varies relative to the ratio of the distance between the adjacent small units 13 (widths W.sub.1 and W.sub.2 of the internal gaps) to diffusion length of thermal neutrons in the pertinent region. Even when the sum of W.sub.1. W.sub.2 is fixed, therefore, the proportion of the decrease in the multiplication factor varies with the varying ratio of the widths W.sub.2 and W.sub.1 To be specific, the proportion of the decrease of the multiplication is small where the two widths W.sub.2 and W.sub.1 are equal and large where the ratio of the widths, W.sub.2 to W.sub.1, increases. The graph of FIG. 20 show the same relation as in the graph of FIG. 19, obtained during the course of operation. During the course of operation, since the proportion of the decrease of the thermal neutron utilization factor is fairly small as compared to that during the period of cold state, the change of the multiplication factor due to the change of the ratio of W.sub.2 to W.sub.1 is small. Similarly to the behavior during the period of cold state, however, the multiplication factor decreases, though slightly, in proportion as the ratio of W.sub.2 to W.sub.1 increases. From the standpoint of improving the shut down margin, therefore, the effect in this improvement increases in proportion as the ratio of W.sub.2 to W.sub.1 increases. To be specific, the shut down margin is enhanced when the width, W.sub.2 (P.sub.3), is increased on condition of W.sub.1 =0(P.sub.2 =P.sub.1). The multiplication factor during the course of operation, however, decreases though slightly in proportion as the ratio of W.sub.2 to W.sub.1 is increased. It is, therefore, desirable to fix the ratio of W.sub.2 to W.sub.1 at such a magnitude that necessary shut down margin is secured as balanced with the degree of concentration of the fuel, the amount of burnable poison, the reactivity worth of the control rod, etc. This means that the shut down margin can be secured as desired by suitable varying the intervals between the adjacent small units 13 by relative position within the fuel assembly. A fuel assembly 11i illustrated in FIG. 21 represents an embodiment wherein partial length fuel rods 20a are disposed one each at the (2, 2) positions from each of the corners of the fuel assembly 11h. Even when the intervals between the small units 13 are varied in relative position as illustrated in the diagram, the partial length fuel rods 20a disposed as described above serve the purpose of preventing the otherwise possible decline of the critical heat flux or critical power in the subchannels near the corners. Even the fuel assembly 11h which has the intervals between the adjacent small units 13 varied by relative position within the fuel assembly can be embodied as variously modified as follows. A fuel assembly 11j illustrated in FIG. 22 represents an embodiment wherein the small units 13 are varied is construction. This fuel assembly 11j comprises small units 13i composed of four fuel rods 12 and disposed one each at the four corners of the fuel assembly, small units 13j composed of 10 fuel rods 12 and disposed between the small units 13i, and a small unit 13k composed of 16 fuel rods 12 disposed around a water rod 15. A fuel assembly 11k illustrated in FIG. 23 represents an embodiment wherein a water rod 15 possesses a varied construction. In this fuel assembly 11k, the water rod 15k has a circular cross section of a diameter equaling one side of a square containing four fuel rods 12 and therefore permits an addition to the total number of fuel rods 12 contained in the fuel assembly. A fuel assembly 11l illustrated in FIG. 24 represents an embodiment in which small units 13 and a water rod 15 both have varied constructions. This fuel assembly 11l comprises small units 13l each composed of 16 fuel rods 12 and disposed at each of the corners of the fuel assembly, small units 13m each composed of three fuel rods and disposed between the small units 13l, and a water rod 15l having a square cross section and disposed at the center of the fuel assembly. This invention can be embodied in a construction of 8 rows and 8 columns as in a fuel assembly 41b illustrated in FIG. 25. In this fuel assembly 41b, four small units 43a composed of 9 fuel rods 42 and disposed one each in the four corners and four small units 43b composed of 6 fuel rods 42 and disposed between the small units 43a are contained within a channel box 44. A water rod 45 having a circular cross section is disposed at the center. In the fuel assemblies 11i, 11j, 11k, 11l, and 41b described above, the component fuel rods are so spaced the intercentral distances P.sub.3, P.sub.2, and P.sub.1 have sizes sequentially decreasing in the order mentioned. Now, the fuel assembly to be disposed in the D-lattice core will be described. In the D-lattice core, a wide gap (GW) having a large width for permitting insertion of a control rod 106 and a narrow gap (GN) having a small width not intended to permit insertion of a control rod are formed outside the channel box 104. The width of the wide gap is about twice that of the narrow gap. The power, therefore, issues more readily from the corner on the wide gap side than from that on the narrow gap side. The fuel rods 102 bordering on the gaps issue the power more readily than the fuel rods 102 not bordering on the gaps. In a fuel assembly 101a, therefore, four small units 103 each formed of fuel rods 102 as described above are contained in a channel box 104 in such a manner as to give rise to a cruciform internal gap (gap for boiling water region 108 between the small units 103. A water rod (W) 105 having a large diameter equaling one side of a square of 3 rows and 3 columns is disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow cap (GN). In a fuel assembly 101b illustrated in FIG. 27, a water rod (W) 105a of a medium cross section equaling one side of a square of two rows and two columns of fuel rods 102 and five water rods (W) 105b each of a diameter equaling that of one fuel rod 102 are disposed in the neighborhood of the channel and a crudiform internal gap 108 is disposed as deviated from the center of the channel in the direction away from the two sides of the channel bordering on the wide gap (GW) and toward the two sides thereof bordering on the narrow gap (GN). A fuel assembly 101b' illustrated in FIG. 28 represents an embodiment resembling the fuel assembly 101b of FIG. 27, excepting one water rod (W) 105 of a large diameter is used in the place of the water rod(W) 105a of a medium diameter and the total of 5 water rods (W) 105b having a small diameter. A fuel assembly 101c illustrated in FIG. 29 represents an embodiment wherein one water rod (W) 105a of a medium diameter, 5 water rods (W) 105b of a small diameter, and one cruciform internal gap 108 are disposed as deviated in the direction of the two sides bordering on the narrow gap (GN). A fuel assembly 101c' illustrated in FIG. 30 represents an embodiment resembling the fuel assembly 101c of FIG. 29, excepting one water rod (W) 105 of a large diameter is used in the place of the water rod (W) 105a of a medium diameter and 5 water rods (W) 105b of a small diameter. In the fuel assemblies 101a, 101b, 101b', 101c and 101c' of the embodiments constructed as described above, internal gaps and water rods (W) are disposed invariably inside channel boxes 104. By deviating the positions for the internal gaps 108 and the water rods (W) toward the narrow gaps (GN) side and consequently causing the distribution of the mederators within the channel box 104 to be deviated toward the narrow gap (GN) side as described above, the disadvantage of the D-lattice core that the power from the fuel rods 102 on the wide gap (GW) side tends to increase as compared to that form the fuel rods 102 on the narrow gap (GN) side can be cancelled. The graph of FIG. 31 compares the conventional fuel assembly and the fuel assembly of the present embodiment in terms of reactivity and local power peaking on the condition that the void fraction is set at 40%, the atomic number ratio of water to-fuel in the fuel assembly at a fixed level (4.7), the average enrichment of fuel at a fixed level (4.5%), and the enrichment of the fuel rods except for those disposed in the corners (the fuel rods in the corners have a slightly lower enrichment than the average) at a fixed level. In the graph, the longitudinal axis is the scale of the difference of infinite multiplication factor based on the data obtained by the conventional fuel assembly and the horizontal axis the scale of the difference of local power peaking based on the data obtained by the conventional fuel assembly and the dots a, b, and c represent the results obtained by the fuel assemblies 101a, 101b, and 101c. It is clearly noted from this graph that the fuel assemblies 101a, 101b, and 101c possess an outstanding property of exhibiting large infinite multiplication factors in spite small local peakings. The descriptions given so far have portrayed this invention as embodied in fuel assemblies each using 9 rows and 9 columns of fuel rods. As demonstrated hereinafter, this invention can be embodied similarly in fuel assemblies using 8 rows and 8 columns of fuel rods and producing the same effects and those of 9 rows and 9 columns of fuel rods. In a fuel assembly 121a illustrated in FIG. 32, similarly to the fuel assemblies cited above, four small units 123 formed of fuel rods 122 are contained within a channel box 124 in such a manner as to give rise to a cruciform internal gap 128 between the small units 123. In the small unit 123 two continuous sides of which border on the narrow gap (GN) side, a water rod (W) of a diameter equaling one side of a square containing two rows and two columns of fuel rods 122 is disposed. In a fuel assembly 121b illustrated in FIG. 33, a water rod (W) 125 having a diameter equaling one side of a square containing two rows and two columns of fuel rods 122 is disposed nearly at the center of the channel and a cruciform internal gap 128 is disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). Then, in a fuel assembly 121c illustrated in FIG. 34, four water rods (W) 125a having a diameter equaling that of one fuel rod and a cruciform internal gap 128 are both disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). A fuel assembly 121c' illustrated in FIG. 35 resembles the fuel assembly 121c illustrated in FIG. 34, excepting one water rod (W) 125 is used in the place of the four water rods (W) 125a. Further, the present invention allows various alteration as follows. In a fuel assembly 101d illustrated in FIG. 36, similarly to the embodiments cited above, nine small units 103 formed of fuel rods 102 are contained within a channel box 104 in such a manner as to give rise to an internal gap 108 of the shape of four sides of a square each extended outwardly slightly from the corners and a water rod (W) 105a of a medium diameter equaling one side of a square contain 2 rows and 2 columns of fuel rods 102 and five water rods (W) 105b of a small diameter equaling the diameter of one fuel rod 102 are disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). A fuel assembly 101d' illustrated in FIG. 37 resembles the fuel assembly 101d illustrated in FIG. 36, excepting one water rod (W) 105 of a large diameter is used in the place of the water rod (W) 105a of a medium diameter and the five water rods (W) 105b of a small diameter. In a fuel assembly 101e illustrated in FIG. 38, a water rod (W) 105 of a large diameter is disposed nearly at the center of the channel and an internal gap 108 of the shape of four sides of a square each extended outwardly slightly from the corners is disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). In a fuel assembly 101f illustrated in FIG. 39, a water rod (W) 105 of a large diameter and an internal gap 108 of the shape of four sides of a square each extended outwardly slightly from the corners are both disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). Now, the present invention embodied in fuel assemblies of 8 rows and 8 columns of fuel rods and provided with an internal gap of the shape of four sides of a square each extended outwardly slightly from the corners will be cited below. In a fuel assembly 121d illustrated in FIG. 40, nine small units 123 formed of fuel rods 122 are contained in a channel box 124 in such a manner as to give rise to an intergap 128 of the shape of four sides of a square each extended outwardly slightly from the corners. In this fuel assembly, the internal gap 128 is located as deviated from the center of the channel in the direction away from the two sides bordering the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). In the central small unit 123, a water rod (W) 125 of a diameter equaling one side of a square containing 2 rows and 2 columns of fuel rods 122 is disposed. In a fuel assembly 121e illustrated in FIG. 41, a water rod (W) 125 and an internal gap 128 of the shape of four sides of a square each extended outwardly slightly from the corners are disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). Fuel assemblies 101g to 101k illustrated in FIGS. 44 to 48 represent embodiments wherein internal gaps 108 each of the shape of four sides of a square each extended outwardly slightly from the corners formed in fuel assemblies of 9 rows and 9 columns of fuel rods were altered in shape. A fuel assembly 101g illustrated in FIG. 45 represent embodiments wherein internal gaps 108 are formed each by the 4-2-3 arrangement of fuel rods 102 and are disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). In the fuel assembly 101g, a water rod (W) 105 of a large diameter equaling one side of a square containing 3 rows and 3 columns of fuel rods 102 is disposed near the center of the channel. In the fuel assembly 101h, a water rod (W) 105a of a medium diameter equaling one side of a square containing 2 rows and 2 columns of fuel rods 102 is disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). A fuel assembly 101i illustrated in FIG. 46, a fuel assembly 101j illustrated in FIG. 47, and a fuel assembly 101k illustrated in FIG. 48 are embodiments wherein internal gaps 108 each formed by the 4-3-2 arrangement of fuel rods 102 are each disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). In the fuel assembly 101i, a water rod (W) 105a of a medium diameter equaling one side of a square containing 2 rows and 2 columns of fuel rods 102 is disposed near the center of the channel. In the fuel assembly 101j, a water rod (W) 105a of a medium diameter is disposed as deviated from the center of the channel in the direction away from the two sides bordering on the wide gap (GW) and toward the two sides bordering on the narrow gap (GN). In the fuel assembly 101k, a water rod (W) 105 of a large diameter equaling one side of a square containing 3 rows and 3 columns of fuel rods 102 is disposed near the center of the channel. Then a fuel assembly 10l. illustrated in FIG. 49 is composed of 9 rows and 9 columns of fuel rods. In this fuel assembly 101, an internal gap 108 of the shape of four sides of a square each extended outwardly slightly from the corners is so formed that the narrow gap (GN) side gaps are larger than the wide gap (GW) side gaps. By giving a larger width to the narrow gap (GN) side gaps of the internal gap 108 than to the wide gap (GW) side gaps thereof as described above, the disadvantage of the D-lattice reactor that the power issues more readily from the wide gap (GW) side than from the narrow gap side (GN) can be cancelled. The fuel assembly of this embodiment permits a decrease in the difference of reactivity during the course of power operation and during the period of cold state and secures shut down margin easily as compared with the fuel assembly incorporating therein an internal gap of a fixed width. Fuel assemblies 121h to 121k illustrated respectively in FIGS. 50 to 54 are each formed of 8 rows and 8 columns of fuel rods. They represent embodiments wherein internal gaps are so formed that the narrow gap (NG) side gaps have a larger width than the wide gap (GW) side taps. In each of the fuel assembly 121h and the fuel assembly 121i, a water rod (W) 125 of a diameter equaling one side of a square containing 2 rows and 2 column of fuel rods 122 is disposed as slightly deviated from the center of the channel toward the wide gap (GW) side. In each of the fuel assembly 121j and the fuel assembly 121k, a water rod (W) 125a of a large diameter equaling one side of a square containing 3 rows and 3 columns of fuel rods 122 is disposed as deviated from the center of the channel toward the narrow gas (GN) side. A fuel assembly 101m illustrated in FIG. 54, a fuel assembly 101n illustrated in FIG. 55, and a fuel assembly 101p illustrated in FIG. 56 are each composed of 9 rows and 9 columns of fuel rods. They each represent an embodiment wherein partial length fuel rods (P) having shortened upper portions are closely juxtaposed to an internal gap 108. A fuel assembly 121l illustrated in FIG. 57 is composed of 8 rows and 8 columns of fuel rods. It represents an embodiment wherein partial length fuel rods (P) having a shortened upper portion are closely juxtaposed to an internal gap 128. These fuel assemblies 101m, 101n, 101p, and 121l permit a decrease in the difference of reactivity during the course of power operation and during the period of cold state because the amount of the water surrounding the internal gaps 108 and 128 increase in the part above the upper ends of the partial length fuel rods (P) during the period of cold state. From the standpoint of securing shut down margin, it is important that the upper part of the fuel assembly which has a high thermal neutron flux during the period of cold state should exhibit a small difference of reactivity during the course of output operation and during the period of cold state. From the standpoint of ensuring economy of the fuel, it is desirable that the uranium inventory should be large. Thus, the part below the fuel rods should be filled with uranium. From the standpoint of ensuring the stability of the core, since the pressure loss in the upper part of the fuel assembly destined to constitute a high void region is desired to be as small as possible, it is important that the partial length fuel rods (P) disposed around the internal gaps 108, 128 should allow a decrease in the pressure loss in the upper part of the fuel assembly.
039375137	summary	FIELD OF THE INVENTION The present invention relates to an apparatus for grabbing and lifting differently shaped elongated objects. More particularly this invention concerns an apparatus for inserting control rods and fuel elements into respective holes in a nuclear reactor and withdrawing them therefrom. BACKGROUND OF THE INVENTION In order to change the control rods and fuel elements in a nuclear reactor a device is known having a pair of side-by-side grabs, one adapted to fit and lock onto the top of the fuel element and the other adapted to lock onto the top of a group of control rods. This arrangement allows the adjustment of a reactor by remote control in that this double grab is mounted on the end of a traditional arm operated by remote control with various servo motors. The principal disadvantage of such a device is that it is extremely expensive. A separate set of controls, actuators, motors, sensors and the like is necessary for each of the grabs; thus the expense if elevated. In addition, such a device is relatively bulky and has a limited service life due to its extreme complexity. OBJECTS OF THE INVENTION It is therefore an object of the present invention to provide an improved apparatus for inserting differently shaped elongated objects into and withdrawing same from respective holes. Another object is the provision of an improved double grab usable both to pick up and to set down control rods and fuel elements in a nuclear reactor. Another object of this invention is the provision of such a double grab which is relatively simple and inexpensive to manufacture. A further object is to provide a double grab of the above-described general type which has a long service life. SUMMARY OF THE INVENTION These objects are attained according to the present invention in a grab having a head pivotally mounted on a vertically and horizontally displaceable support. This head has a pair of opposite sides on each of which is provided a respective grab, one adapted to engage and lock on the top of a fuel element and the other adapted to engage and lock on the top of a group of control rods. Means is provided for pivoting the head on the support between a pair of positions corresponding to alignment of a respective grab with a respective object. Thus the overall size of the device is reduced considerably, as the operator need merely turn the head of the grab over in order to change its function. In accordance with yet another object of this invention the pivotal head is provided with a single actuator displaceable against spring force in a single direction and connected to both of the grab means such taht on displacement in this direction against the spring force it opens both of these grabs. This single actuator therefore operates both of the grabs and the overall cost and complexity of the pickup device is considerably lessened. The operating element of this actuating means is according to the present invention a fluid-operated cylinder whose piston is displaceable in a direction which causes grab formations on the grabs to release the respecitve objects by means of fluid pressure and is urged constantly in the other direction by a spring. Thus any failure of fluid pressure cannot cause the grabs to drop their respective objects, a situation which could be extremely dangerous and costly in a nuclear reactor. In accordance with yet another feature of this invention the grab head is generally elongated, having one of the grabs on each of its ends, and is pivotal about a horizontal axis perpendicular to the longitudinal axis of the grab head. The grab is provided at its pivot with means for rotating it through 180.degree. so as to be able to bring either of its ends into a downwardly facing position for engagement of the complementary formation thereon with the top of a control rod assembly or a fuel element. This pivoting means comprises in accordance with the present invention a pinion carried on the grab head at the rotation axis thereof and a rack which is in mesh with the pinion and displaceable so as to rotate the head. According to yet another feature of this invention means is provided for locking the grab head in either of its two positions, that is in the position with one grab facing downwardly and the other upwardly and vice versa. This locking means comprises a locking element engageable with the upper ends of the grab head so as to lock this grab head pivotally. The locking element engages in a pair of slots and is normally urged downwardly toward the grab head into the upwardly directed slots that are provided on each end thereof. Fluid pressure can displace the locking element away from these slots so that once again a failure of fluid pressure will not cause the locking element to release so that the head can wobble and perhaps drop its load. In accordance with features of this invention sensors are provided to detect the position of the operating elements of both the locking mechanism and the actuating mechanism. These sensors give outputs which indicate that the grab head is locked pivotally in place and that the grabs are in the locking or closed positions. The pickup assembly according to the present invention is relatively compact and has a very simple mechanism. It is operated by remote control with relative ease, as only three controls need be provided; the orienting arrangement which determines which grab is facing downwardly toward the control rods or fuel elements, the pressurization apparatus for the operating member of these grabs, and the pressurization apparatus for the piston operating the locking mechanisms that holds the grab head rigidly in place. Otherwise this entire assembly is mounted on a support constituted by the end of the conventional pickup cat or arm of a nuclear power plant.
062185926	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a method and an apparatus for the treatment of radioactive evaporator concentrates from the evaporation system of nuclear plants, which allow efficient separation, as a non-radioactive product, of the sodium sulfate contained therein and, thus, a cost-effective reduction in volume of the concentrates to be decontaminated and disposed. 2. Description of the Related Art In nuclear power plants comprising a boiling water reactor, the main condensate is freed from solid ionic and radiochemical contaminations by means of ion exchange resins. During regeneration of said ion exchange resins with caustic and sulfuric acid, solutions are produced which, after having been neutralized, are conveyed into an evaporating plant, where they are concentrated together with other radioactively loaded effluents to a solids content of about 20%. Thus, depending on the size of the plant, about 20 to 100 m.sup.3 of evaporator concentrates per year are obtained as liquid radioactive waste. The evaporator concentrates contain about 1 to 30% by volume of separable solids (ion exchange resin residues, flocculated detergent residues, heavy metal oxides, fibers etc.). Dissolved components are mainly sodium sulfate, the proportion of which may be about 10 to 25% by weight, and other components, such as incrustation inhibitors (e.g. EDTA), surfactants, organic and inorganic salts of decontaminating agents (e.g. citrates, oxalates, phosphates), activation products and others. In the past, these evaporator concentrates had been further evaporated in a subsequent process step for conditioning and solidified to form a monolithic sodium sulfate block which then proceeded to ultimate waste disposal or interim storage. This, however, has the disadvantage that vast volumes of waste are produced and no decontamination effect is achieved, since radioactive components are encapsulated in the crystals when the sodium sulfate crystallizes. Although 80% or more of the mass to be ultimately disposed of consists of non-radioactive sodium sulfate, according to current practice the latter has to be disposed of in a rather expensive way together with the radioactive components. The suggestions made so far for improving the treatment of evaporator concentrates and similar effluents from nuclear plants were directed to binding the effluent's radioactivity by suitable precipitation or flocculation reactions in such a way that the radioactive products can be removed. The remaining decontaminated solution should then be disposed of as weakly radioactive effluent. The disadvantages thereof were the bad precipitation efficiency, so that said process could not be used in modern nuclear plants due to the associated increase in radioactive release. BRIEF SUMMARY OF THE INVENTION An object of the invention, thus, is to provide a method and an apparatus for the treatment of radioactive evaporator concentrates from nuclear plants, which allow efficient separation of radioactive and non-radioactive components and, thus, a cost-effective reduction in the volume of waste to be disposed. These and other objects are achieved in accordance with the invention by a method for the treatment of radioactive evaporator concentrates from the evaporation system of nuclear plants, wherein (a) the evaporator concentrate is freed from undissolved components, PA1 (b) the sodium sulfate contained in the evaporator concentrate is crystallized as Glauber's salt (Na.sub.2 SO.sub.4.10H.sub.2 O) on an immersion cooler, PA1 (c) the Glauber's salt deposited on the immersion cooler is recrystallized, PA1 (d) optionally, the recrystallization step (c) is repeated once or several times, PA1 (e) essentially inactive sodium sulfate is separated from the process, and PA1 (f) the depleted evaporator concentrate of step (a) is recycled to the evaporation system. The present invention also provides an apparatus for carrying out said method, said apparatus comprising a container for the evaporator concentrate, an immersion cooler having inlet(s) and outlet(s) for a cooling or heating medium, and a cooling or heating aggregate for providing said cooling or heating medium.
	claims	1. A centering pin for centering a nuclear reactor core of a nuclear power plant in a reactor vessel, comprising a central part having a radially inner edge oriented towards the nuclear reactor core and a horizontal thickness along the radially inner edge, an upper hydrodynamic profile, which is disposed above the central part and forms a vertical wing leading edge extending from the central part and having an upper height above the central part and defined between an upper end of the central part and an uppermost point of the upper hydrodynamic profile, and a lower hydrodynamic profile, which is disposed below the central part and forms a vertical wing trailing edge extending from the central part and having a lower height below the central part and defined between a lower end of the central part and a lowermost point of the lower hydrodynamic profile, wherein the upper height of the vertical wing leading edge of the upper hydrodynamic profile has a maximum variation of more or less 25% relative to the horizontal thickness of the central part, and wherein the lower height of the vertical wing trailing edge of the lower hydrodynamic profile has a maximum variation of more or less 25% relative to twice the horizontal thickness of the central part. 2. The centering pin according to claim 1, wherein the upper hydrodynamic profile has a shape of a dihedral. 3. The centering pin according to claim 1, wherein the lower hydrodynamic profile has a shape of a pyramid. 4. The centering pin according to claim 1, wherein the upper height of the vertical wing leading edge of the upper hydrodynamic profile is in an order of 0.87 times the horizontal thickness of the central part, while the lower height of the vertical wing trailing edge of the lower hydrodynamic profile is in the order of 1.93 times the horizontal thickness of the central part. 5. A nuclear power plant reactor comprising a reactor vessel, a nuclear reactor core and centering pins according to claim 1, wherein the nuclear reactor core is centered in the reactor vessel by the centering pins, wherein the radially inner edge of the central part of the centering pins is adjacent to the nuclear reactor core, wherein the central part of the centering pins has a radially outer axial face adjacent to the reactor vessel. 6. The nuclear power plant reactor of claim 5, comprising a cooling fluid circulating in a direction of flow in an annular space situated between the nuclear reactor core and the reactor vessel,wherein at least a part of the centering pins are situated in the annular space,wherein the upper hydrodynamic profile of the centering pins is located upstream in the direction of flow of the cooling fluid,wherein the lower hydrodynamic profile of the centering pins is located downstream in the direction of flow of the cooling fluid. 7. The nuclear power plant reactor of claim 6, wherein the lower hydrodynamic profile of the centering pins located downstream in the direction of flow of the cooling fluid reaches a bottom of the reactor vessel in a plenum of the reactor vessel.
053501616	claims	1. A grid strip for a nuclear reactor fuel assembly, the grid strip having a plurality of loaded metal cantilever springs which grow and relax when subjected to irradiation, each spring having a first end portion connected to the grid strip, a second end portion with an unattached terminal end, a length extending along the spring from the first end to the second end, and a thickness, the first end portion including a first curved portion which is curved in a first direction and the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, the first curved portion having a first layer including a convex exterior surface and a second layer including a concave exterior surface, the first and second layers being opposite each other in a direction parallel to the thickness of the spring, the first layer having different material characteristics than the second layer such that upon irradiation of the grid strip the relaxation of the spring is at least partially offset by differential growth of the first and second layers, the second end portion having substantially uniform material characteristics throughout its thickness. 2. A grid strip according to claim 1, wherein the grid strip is formed from a zirconium alloy. 3. A loaded metal nuclear reactor cantilever grid spring which grows and relaxes as a result of irradiation, the spring having a length, a thickness, a supported first end portion and a second end portion with an unsupported terminal end, the first end portion including a first curved portion which is curved in a first direction and the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, the first curved portion including a first layer having a convex exterior surface and a second layer having a concave exterior surface, the first and second layers being opposite each other in a direction parallel to the thickness of the spring, the first layer having different material characteristics than the second layer such that upon irradiation of the grid spring, any relaxation in the spring is at least partially offset by differential growth of the first and second layers, the second end portion having a substantially uniform composition and substantially uniform material characteristics throughout its thickness. 4. A spring according to claim 3, wherein the first and second layers have approximately equal thicknesses. 5. A grid spring according to claim 3, wherein the spring is formed from a zirconium alloy. 6. A spring according to claim 3, wherein the convex exterior surface is a shot peened surface and the concave exterior surface is a non-shot peened surface. 7. A process for minimizing the loss in load of a zirconium cantilever spring, the spring having a length, a thickness, a supported first end portion and an opposite second end portion with un unsupported terminal end, the first end portion including a first curved portion which is curved in a first direction and has opposite convex and concave exterior surfaces, the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, the process comprising the step of shot peening the convex exterior surface of the first end portion without shot peening the second end portion to form a shot peened layer which extends through only part of the thickness of the first curved portion, the degree of shot peening being sufficient to result in differential growth of the first and second layers of the spring upon exposure to a neutron flux. 8. A process according to claim 7, wherein a portion of the spring is masked during shot peening. 9. In a metal cantilever grid spring for use in a nuclear reactor fuel assembly, the spring being formed from a metal which grows as a result of exposure to radiation and comprising a supported first end portion and a second end portion with an unattached terminal end, the first end portion including a first curved portion which is curved in a first direction and the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, the first curved portion including a concave exterior surface and an opposite convex exterior surface, the improvement wherein the first curved portion of the spring includes a first layer including the convex exterior surface and a second layer including the concave exterior surface, the first and second layers each having a length and being opposite each other in a direction parallel to the thickness of the curved portion, the first layer having been specially treated to experience an increase in length upon exposure to a neutron flux which is sufficiently different from the increase in length of the second layer upon exposure to the neutron flux to result in a lower rate of radiation-induced loss in spring load of the spring than the rate of radiation-induced load loss of a spring which does not have a curved portion with a specially treated first layer, the second layer and the second end portion remaining untreated. 10. A spring according to claim 9, wherein the metal comprises a zirconium alloy. 11. A spring according to claim 9, wherein the first layer has a first thickness and the second layer has a second thickness, the first and second thicknesses being approximately equal. 12. A component for use in a nuclear reactor fuel assembly having a cantilevered portion, the cantilevered portion being formed from a metal which grows as a result of exposure to a neutron flux, the cantilevered portion having a length, a thickness, a supported first end portion, and an opposite second portion with an unsupported terminal end, the first end portion including a first curved portion which is curved in a first direction and the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, the first curved portion including a first layer having a convex exterior surface and a second layer having a concave exterior surface, the first and second layers being opposite each other in a direction parallel to the thickness of the spring, the first and second layers having different material characteristics such that upon irradiation of the grid spring any relaxation in the spring is at least partially offset by differential growth of the first and second layers, the second end portion having a substantially uniform composition and substantially uniform material characteristics throughout its thickness, the degree of curvature of the first curved portion changing upon exposure to a neutron flux due to differential growth of the first and second layers. 13. A method of making a cantilevered metal component for use in a nuclear reactor, comprising: obtaining a component which is formed from metal that grows as a result of exposure to radiation, the component having a length, a thickness, a supported first end portion and an opposite second end portion with an unsupported terminal end, the first end portion including a first curved portion which is curved in a first direction and has opposite concave and convex exterior surfaces, the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, and cold working the convex exterior surface of the first curved portion without cold working the second end portion to result in the first curved portion having a first degree of curvature and to impart to the first curved portion characteristics sufficient to cause the degree of curvature of the first curved portion to increase to a second degree of curvature upon exposure to a neutron flux. an attached first end portion which is connected to a support, an opposite second end portion with an unattached terminal end, the first end portion including a first curved portion which is curved in a first direction and the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, the first curved portion including a shot peened first layer including a convex surface, and a second layer overlapping the first layer in a direction which is parallel to the thickness of the curved portion and including a concave surface, the first layer growing at a different rate than the second layer upon exposure to a neutron flux, the second end portion having substantially uniform material characteristics throughout its thickness. (a) obtaining a grid spring having a length, a thickness, a supported first end portion and an opposite second end portion with an unsupported terminal end, the first end portion including a first curved portion which is curved in a first direction and has opposite concave and convex exterior surfaces, the second end portion including a second curved portion which is curved in a second direction which is opposite to the first direction, and (b) shot peening the convex exterior surface of the first curved portion without shot peening the concave exterior surface and the second end portion to form a shot peened layer which extends through only part of the thickness of the first curved portion. (c) prior to step (b), masking a portion of the spring surrounding the convex exterior surface of the first curved portion. 14. A method according to claim 13, wherein the step of cold working comprises shot peening. 15. A fuel assembly for use in a nuclear reactor, comprising a metal component, the component having a cantilevered portion including: 16. A fuel assembly according to claim 15, wherein the cantilever portion of the metal component is formed from a zirconium alloy. 17. A process for producing a pre-load retaining nuclear reactor grid spring for supporting a fuel rod, comprising the steps of: 18. A process according to claim 17, further including the step of:
042344492	claims	1. A method of treating radioactive aklali metals or radioactive solid salts thereof, said method comprising mixing particulate silica substrate material having a particle size such that the majority of the substrate material passes through a 200 mesh sieve and the radioactive material in a rotary drum calciner and converting the radioactive material to alkali metal monoxide by reaction with oxygen present in a diluent at a temperature sufficient to initiate the reaction thereby forming particulate substrate particles coated with alkali metal monoxide, said reaction temperature being controlled by the amount of oxygen present in the diluent to ensure the reaction product remains flowable for easy handling. 2. The method set forth in claim 1, wherein the alkali metal is sodium or potassium. 3. The method set forth in claim 1, wherein the temperature is at least as high as the melting point of the highest melting alkali metal present but not greater than about 200.degree. C. 4. The method set forth in claim 1, wherein the weight ratio of silica to alkali metal monoxide is about seven to one. 5. The method set forth in claim 1, wherein the particulate substrate material and the radioactive material are continually mixed during the conversion to the monoxide. 6. The method set forth in claim 1, wherein oxygen is present in an amount up to about 20% by volume. 7. A method of treating radioactive sodium or potassium metals or radioactive solid salts thereof, said method comprising mixing silica particles most of which are of a size to pass through a 95 mesh screen and the radioactive material, and oxidizing the radioactive material to the monoxide in a rotary drum calciner by passing oxygen in a diluent over the mixture to form particulate silica coated with sodium monoxide or potassium monoxide or mixtures thereof, the reaction temperature being controlled by the amount of oxygen present in the diluent and being maintained at about 250.degree. C. or less. 8. The method set forth in claim 7, wherein most of the silica present passes through a 200 mesh screen. 9. The method set forth in claim 7, wherein the diluent is argon. 10. A method of storing radioactive waste as glass, comprising providing radioactive alkali metals or solid salts thereof, mixing particulate silica having a particle size the majority of which passes through a 200 mesh screen with the radioactive material, oxidizing the radioactive material in a rotary drum calciner at a temperature less than about 200.degree. C. by passing oxygen in a heavier than air diluent over the mixture to form particulate silica coated with alkali metal monoxide which is easily flowable and fusing said alkali metal monoxide coated silica to form glass. 11. The method set forth in claim 10, wherein the radioactive material has a sodium cation or potassium cation or mixtures thereof. 12. The method set forth in claim 10, wherein the weight ratio of silica to alkali metal monoxide is about five to one. 13. The method set forth in claim 10, wherein the diluent is argon and oxygen is present in an amount not greater than about 20% by volume.
	summary
048204722	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings wherein like reference characters designate like or corresponding parts through the several views, there is shown in FIGS. 1 and 2 a spent fuel rack module 15 which consists of an array of containers or cells 16 each of a size and configuration for holding a nuclear fuel assembly (not shown). An 11.times.11 array is shown but any number of cells, including a rectangular array may be used depending on the design and configuration of the pool. The fuel assembly may be either a fresh or spent fuel assembly since both are of the same size and the fuel racks must be designed to meet Nuclear Regulatory Commission criteria for storage of spent or new nuclear fuel. The fuel racks are of modular design and the module shown in FIGS. 1 and 2, is one of many arranged to be located in a spent fuel storage pool at the site of a nuclear reactor. Although storage pools vary in size, they generally range in depth from about 20 to 40 feet and hold anywhere from two hundred to about sixteen hundred fuel assemblies. The pool walls are formed of reinforced concrete and are particularly designed in accordance with NRC specifications to withstand seismic forces. To remove heat which continues to be generated by the fuel assemblies, water or other coolant is circulated in heat exchange relationship with fuel rods in the assemblies in a manner well known in the art. The spent fuel rack module shown includes a support including a base structure 18 arranged to be supported from the pool liner 20 (FIGS. 9, 10, 11) by leveling pads 22. The base structure of stainless steel is of sufficient thickness to carry the full weight of the cells 16 and fuel assemblies without distorting and still maintain vertical alignment of the cells positioned thereon. To achieve the desired horizontal spacing and vertical alignment of cells 16, multiple X-axis box beams 24 and Y-axis box beams 26 are mounted on and welded to base structure 18 to form lower grid 28. As shown in FIG. 1, X-axis box beams 24 extend unbroken from one end across base structure 18 to the other side of the module, while x-axis beams 26 include short sections which extend between and are welded to the unbroken parallel beams 24. FIG. 4 illustrates the location and extent of welds 30 made at the intersection of X-axis and Y-axis beams and the base plate 18. An upper grid structure 32 vertically displaced from lower grid 28, is constructed of similar X-axis and Y-axis box beams 34, 36. This interlocking arrangement of box beams in both lower and upper grid structures form multiple openings of square configuration aligned vertically to receive the stainless steel cells 16. The cells have walls approximately 0.10-inch thick and are open at both ends. Although the cells may be made of any size or configuration, the design chosen to illustrate this invention has an inner dimension of 8.75 inches in length and width directions and is approximately 14 feet high. the bottom end of each cell welds to lower grid structure 28. The base structure 18 under the grid structure 28 is equipped with openings between its plates 120 through which coolant is adapted to flow upwardly through the fuel assemblies to carry away generated heat (FIGS. 2 and 3). To impart reasonable rigidity to the complete module and to maintain the same uniform distance between all cells in the module while obtaining parallelism between cell centerlines, each cell is fixedly secured by welding on all sides to adjacent box beams in both the lower and upper grid structures as more fully described hereafter. FIGS. 1, 2 and 3 show a lower side plate 40 which extends completely around the module and is attached to be plate structure 18 by welds 42 (shown in FIG. 3), and includes in the FIG. 3 modification only a weld 44 between the side plate and cell walls. These welds run along the base structure length and along the sides of the peripheral cells in the module. This arrangement imparts strength and rigidity to the lower outside areas of the module. Likewise, the upper part of the module includes an upper side plate 46 which encompasses the complete module and is welded to each cell on the module periphery. Both peripheral plates 40, 46 encompass the cell module at low and high elevations to accurately define the module outer limits and to help impart squareness and strength to the complete module. The plates 40 in the FIG. 2 modification and 46 in both the FIG. 2 and FIG. 3 modifications which surround the module are welded respectively at 51, 53 along their top edges 54, 56 (FIG. 2) to the dimples 49 on the cell walls which face outwardly from the module. As illustrated in the preferred embodiment of the invention in FIG. 2, the dimples 49 are formed in the walls of each cell at an elevation near the top of box beams 34, 36, 24, 26. These dimples may be of the design shown, or of other configuration, such as a continuous deformation of the cell wall, which projects outwardly a distance at least equal to the thickness of neutron absorbing material 64 and wrapper plates 66. Welds 51, 53 are made at the interface of the dimples and lower and upper side plates 40, 46, while welds 55, 57 are made between box beams 24, 26 and 34, 36 and dimples 49. In the alternative design of FIG. 3 , in lieu of using dimples which project into the inter-cell space at both the bottom and top of the module, plates 48 are welded at 52, 130 to the cell surface but only near the upper end of cells 16. The space between cells on box X and Y axes near the base plate 18 is occupied by box beams 24, 26 of slightly larger size than box beams 34, 36 located thereabove. Welds 60 secure the lower box beams to the sides of cells 16 while welds 62 secure the upper box beams to the surface of plate 48. In both modifications which utilize dimples and protective plates 48 to which the side plates 46 and box beams are welded, the depth of dimple and protective plate thickness respectively is chosen or made to a greater dimension than the combined thickness of the neutron absorbing material 64 and wrapper plate 66. The protective plates or dimples will therefore extend into the inter-cell space for a distance greater than the combined depth of neutron material and wrapper plate in order to protect the latter when the cells are installed in the network of lower and upper box beams which comprise the basic structure of the spent fuel racks. FIG. 1 shows how the bottom end of cells 16 snugly fit in the aligned square openings formed by the lower and upper grid assemblies 28, 32. As shown in FIG. 3, welds 58 attach the bottom edges of box beams 24, 26 to the plates 120 of the base structure 18 while welds 60 secures the upper edges of the box beams to the lower sides of cells 16. The box beams 34, 36 in the upper grid structure are welded at 62 along their upper edges to the protective plates 48 attached to all four sides of each cell. By welding both the beams and upper grid structures to the cell surfaces in this manner, a strong relatively rigid module is formed which not only will provide parallelism and vertical alignment between all cells but will also accommodate seismic disturbances. In order to assure that fuel stored in the cells will not reach a critical mass, neutron absorbing material 64 mounted on the cell surface together with the space between cells which is occupied by water, or borated water, will effectively minimize neutron activity. The cross-sectional views in FIGS. 2 and 3 of a cell shows that the neutron absorbing material, preferably "Boraflex" which is boron carbide in an elastomeric silicon polymer matrix manufactured by Brand Industrial Services, Inc., of Park Ridge, Ill., is attached to all sides of each cell. Other equivalent materials may be used if desired. As shown in FIGS. 2 and 3, the material 64 may be in the form of sheets of material which cover substantially the full surface area of the cell walls but terminates just short of the sides and the top and bottom grid structures. A wrapper plate 66 of a size slightly larger than the material is welded to the sides of the cell by tack welds 68 to retain the neutron absorbing material. Water tightness is not essential since the neutron absorbing materials used is not adversely affected by contact with the pool environment. The total thickness of material 64 and wrapper plate 66 is less than the thickness of the dimples 49 (FIG. 2) or protective plate 48 (FIG. 3), the purpose being that when the cells are loaded into or removed from the module, the greater thickness of dimple 49 or protective plate 48 will permit that part of the cell having the wrapper plates thereon to pass freely through the upper grid structure without damaging the wrapper plates and material 64 surfaces. One Nuclear Regulatory Commission requirement for spent fuel racks is that they must withstand seismic forces. In the present invention, this is accomplished by utilizing the interconnected box beam and side plate arrangements described above. Since the upper and lower grid structures 32, 28 are welded to the cells in the manner shown, they impart substantial rigidity to the complete fuel rack module and thus meet the NRC seismic criteria. In those geographical areas where seismic activity is relatively high, additional rigidity may be incorporated in the module by welding shear plates which extend substantially the complete height of the module, to adjacent cells around the complete periphery of the module. These shear panels may be welded at selected points along their length or along the complete length as desired. The foregoing discussion indicates the need to have the center line; i.e., the longitudinal axis, of each fuel assembly cell perpendicular to the base plate on which the cells are arranged to be positioned. In the present invention, leveling of the base plate 18 is accomplished by utilizing leveling pads 22 more fully described hereafter, positioned under base structure 18 at each of the corners of the module, and beneath the selected section of base structure 18, depending on the loads carried by the base plate. Each cell in the module is of square cross-section and of a size to fit into the complementary and vertically aligned square openings formed by the lower and upper grid structures 28, 32. As more clearly shown in FIGS. 1 and 2, the cell upper walls have funnel cell flanges 39 which flare outwardly to help guide a fuel assembly into the cell during the loading process. The upper ends terminate just above the upper grid 32 and a brace 38 (FIG. 1) shaped to the same configuration as the flared sections, is welded to the outer wall of the peripheral cells in the module. The brace serves a support function and helps keep the cells in proper alignment. Additional rigidity may be imparted to the structure by welding the shear plates mentioned above to adjacent cells 16 having their surfaces on the modular periphery. Each plate is of a width sufficient to bridge the gap 70 between adjacent cells and thereby overlap the cell walls which face outwardly. Preferably, the longitudinal edges of the shear plates terminate short of wrapper plate 66 and the vertical edges on each end of the shear plates are then welded to the cell walls. This construction is repeated on adjacent cells in the outer rows in the module to provide a degree of rigidity to the complete module and, if necessary, help meet NRC seismic criteria for spent fuel racks. It is essential that the spaced cells comprising the fuel rack have the cell walls aligned vertically to help assure unimpeded loading and removal of fuel assemblies from the fuel rack cells. This is accomplished in the apparatus disclosed herein by providing adjustability to the base structure 18 on which the fuel assemblies rest. The pool liners rarely are exactly flat and level and leveling means is therefore necessary to adjust base structure 18 to a horizontal condition. FIGS. 4 through 11 show the structure needed for leveling purposes. FIG. 4 is a plan view of a portion of base structure 18 and shows the box beams 24, 26 welded to its upper surface. There are openings 73 between the plates 100 of the base structure 18 through which coolant flows upwardly for cooling the fuel assemblies and for providing access to leveling pads located therebeneath. FIG. 5 shows leveling pad 22 located at the four corners of the base structure 18 and at selected other positions beneath the base structure as necessary to adequately support the load on the upper surface of the base structure. FIGS. 5-7 show lifting plates 76 welded to the underside of the base structure 18 for module lifting purposes. The lifting plates are about 1" thick and have rectangular openings 78 which extend upwardly through the base structure 18. Multiple stop bars 79 at each corner of opening 78 extend downwardly from the plate underside so that when a lifting lug is moved downwardly through rectangular opening 78 and rotated 90.degree. to enable lifting upwardly on the underside of the plate, the stop bars serve to preclude inadvertent movement of the lug to a position where it could slip upwardly through the plate opening. Referring more specifically to the arrangement for leveling base structure 18, FIG. 9 shows a pool liner 20 having a pedestal 80 mounted for free unrestricted movement on the pool liner surface. The pedestal includes an accurate surface 82 which merges into upstanding cylindrical walls 84. A leveling screw 86 having a spherical surface 88 formed on its bottom end is complementary to pedestal surface 82. After the end of leveling screw 86 is placed in the pedestal, a circular plate 87 is welded to the walls 84 thus leaving a space 89 into which the end of the screw may move if necessary for leveling purposes. External threads 90 on the screw mesh with similar threads 92 on a support pad 22 so that when the screw is rotated by a tool in slot 96, the support pad is caused to be moved vertically. The support pad 22 includes four radially spaced support arms 97 attached to the underside of base plate 18 by welds 98. In operation, to adjust base structure 18 to a level condition, leveling pads 22 are located beneath base structure 18 corners and at selected positions under the base structure central area (see FIG. 5). To level the base structure and cells thereon, each pedestal 80 is moved in an amount and direction to have its bottom surface tilt or conform to the slope of the floor on which pool liner 20 rests. If the liner floor is uneven, the complementary spherical surfaces on the pedestal and leveling screw are adjusted to each other until the axis of leveling screw 86 lies in a vertical plane. By inserting a tool in slot 96 through opening 73 in the base structure and rotating the leveling screw, which then acts as a bearing, the support pad 22 will move vertically and thus raise or lower the base structure 18 to a desired position. This action is repeated for each corner and central area support pad until the base structure 18 is adjusted to a horizontal position. The support pad of FIG. 10 is used in those spent fuel racks already in place which have shear studs 100 embedded in the pool floor. The leveling pad parts are otherwise the same and include a pedestal 80 modified to include a central opening 102 through which stud 100 projects. The diameter of opening 102 is sufficiently large to accommodate non-verticality in the stud 100 and variations in slope in the pool floor. Since the stud serves to maintain the position of a leveling pad in the pool floor area and subjected to shear only in the event of a large seismic disturbance, it is designed to a length to extend upwardly into the leveling screw 86 to a relatively short distance. As in the case of the pedestal opening, a space 104 is provided between the stud and leveling screw walls to provide flexibility in fitting the parts to each other. As a tool in slot 96 rotates the leveling screw, the coacting threads on the screw and support pad 22 cause the support pad and base structure 18 to move vertically until a horizontal position is reached. The support pad of FIG. 11 is likewise used in those fuel rack installations where studs are already embedded in the pool floor. In this design, the stud 100 extends the full length of leveling screw 86 and is topped by a spherical anchor nut 106 and anchor washer 112 which restrains the rack vertically. The upper end of the stud has external threads 108 to accept the threads of nut 106. The spherical washer 112 bears between the upper end of leveling screw 86 and the nut to provide spherical surfaces which help to compensate for the potential misalignment of the stud embedded in the pool floor. As in the previous modifications, the pedestal seeks the floor slope and the spherical surfaces on the pedestal and leveling screw permit the leveling screw to lie in a vertical plane, all within the range of spaces 89 and 110 provided in the pedestal assembly. Adjustment of the support pad 22 vertically by the leveling screw 86 causes variation in the horizontal position of base structure 18. It will be apparent that many modifications and variations are possible in light of the above teachings. It therefore is to be understood that within the scope of the appended claims, the invention may be practiced other than as specifically described.
053352521	summary	BACKGROUND OF THE INVENTION The present invention relates to heat exchange apparatus transferring heat from a reactor primary coolant, typically helium or carbon dioxide, to a secondary fluid medium, typically water and steam, and more particularly to a novel superheating arrangement in which a reheater tube bundle located within a nuclear pressure vessel works in conjunction with superheater tube bundles which are located outside of the nuclear pressure vessel. The reheater absorbs sufficient heat from the reactor gas coolant to supply required reheat steam to the reheat turbine, and in addition the reheater absorbs excess heat at higher temperature than required to meet the reheat turbine inlet steam conditions. The excess heat contained in the reheat steam flow is transferred regeneratively to the external superheater tube bundles to raise the superheat temperature of the main steam flow to the temperature required by the main steam turbine. While it is understood that various fluids can be used for the reactor primary coolant and the secondary fluid medium, the descriptions which follow shall employ the terminology reactor gas coolant to describe the reactor primary coolant and water and or steam to describe the secondary fluid medium. It is desirable to remove heat from gas cooled nuclear reactors by circulating superheated steam at maximum temperature to maximize volumetric and thermal efficiency. This is typically done with tubular heat exchangers specifically referred to as steam generators. A steam generator is comprised of a series of high pressure main steam tube bundles which supply steam to the high pressure main steam turbine, and a lower pressure reheat tube bundle which supplies steam to the lower pressure reheat turbine. Within the nuclear pressure vessel the main steam tube bundle is comprised of an economizer/evaporator tube bundle stage in which feedwater is raised in temperature and evaporated to steam, and an initial superheater tube bundle stage in which the main steam flow is superheated to a desired level for exit from the nuclear pressure vessel. The intermediate superheater and the finishing superheater tube bundles are contained in separate pressure vessels located outside of the nuclear pressure vessel. Steam exiting from the initial superheater tube bundle stage is raised in temperature in the intermediate superheater tube bundle until stress limitations on the heat transfer tube material require a higher grade tube material. Accordingly, the finishing superheater tube bundle, pressure vessel and other components are constructed of materials having design stress limits high enough to accomodate the final steam temperature required by the main steam turbine. A bi-metallic weld is provided between the intermediate superheater tube bundle and the finishing superheater tube bundle. Inlet and outlet penetrations in the walls of the various pressure vessels provide for passage of water and steam flow to and from the respective tube bundles. A steam generator can be designed to make steam at subcritical (less than 3206.2 psia) pressure or supercritical (greater than 3206.2 psia) pressure. In a subcritical system water changes to steam with heat addition at constant temperature and with water density exceeding steam density, while in a supercritical system the phase change is temperature dependant, occuring without a change in density. By employing a supercritical main steam system reheat steam pressure can be raised above reactor gas coolant pressure such that radiation bearing reactor gas coolant cannot leak into reheat steam. Because of space limitations in the nuclear pressure vessel a once-through steam generator is preferred over a drum type steam generator in gas cooled reactors. However, once-through steam generators have certain inherent problems when utilized in gas cooled reactors. In prior designs utilizing once-through steam generators in gas cooled reactors parallel tube circuits were continuous from feedwater inlet to finishing superheater outlet so that steam temperature could not be equalized among tube circuits by the use of mixing headers. Also the lack of intermediate mixing headers and confinement in the nuclear pressure vessel precluded the use of water recirculation to provide flow stability (positive upward flow in all tube circuits) during low load and start-up operation of the plant. As a result flow resistance in the form of orifices at tube circuit inlets had to be provided. Orifices imposed a large pressure drop penalty and had a predisposition to foul by build-up of deposits from impurities in the feedwater. Special feedwater demineralizer systems had to be employed to reduce fouling of the otherwise non-maintainable orifices. Another problem with the use of once-through steam generators in gas cooled reactors was protection of the bi-metallic weld which had to be located within the nuclear pressure vessel in the tubing connecting the intermediate and finishing super-heater stages. Because the bi-metallic welds were not maintainable the use of special insulation and temperature sensors was required. It has been accepted practice with once-through steam generators in gas cooled reactors to plan for plugging of failed tubes because access for replacement of these tubes was not available. The potential for tube failure was high due to vibration and wear of tubes, blockage of tubes from orifice fouling, thermal stress at bi-metallic welds, and over heating due to low flow instability, poor gas and or water/steam flow distribution, and gas hot streaks and unmixed tube side flow. The inability to provide recirculation flow during low load and start-up operation also limited main steam outlet pressure such that it was substantially lower than reactor gas coolant pressure. High safety gas cooled reactor designs eliminated reheating from the steam generator system because of potential leakage of radiation bearing reactor gas coolant into reheat steam, leading to further reduction of main steam outlet pressure. As a result the plant was deprived of several economic advantages including thermal and volumetric efficiency and the use of standard turbine equipment. In general the difficulties with once-through steam generators and the lack of reheaters have prevented gas cooled reactors from realizing the very high temperature capability of the graphite core. The advantages of the present invention, namely a steam generator heat removal system having once-through capability at normal load combined with capability for water recirculation at low load and during start-up, working in conjunction with a balanced pressure reheater will avail gas cooled reactors of highest temperature potential. SUMMARY OF THE INVENTION One of the primary objectives of the present invention is to provide a novel steam generator for gas cooled reactors which can maximize thermal and volumetric efficiencies of the plant, is relatively compact, and provides greater ease of fabrication, installation and inspection than heretofore obtainable with known gas cooled reactor steam generator heat removal systems. A more particular object of the present invention is to provide a novel steam generator for transferring heat from a reactor gas coolant to a secondary fluid medium which may be at subcritical (less than 3206.2 psia) pressure or supercritical (greater than 3206.2 psia) pressure. The reheater portion of the steam generator is located inside of the nuclear pressure vessel, and is capable of absorbing sufficient heat to meet the requirements of the reheat system as well as the heat requirements of the finishing superheater and the intermediate superheater tube bundles, which are located outside of the nuclear pressure vessel. The excess heat absorbed by the reheater which is over and above the heat required to produce steam for the reheat turbine is transferred to the finishing superheater and intermediate superheater tube bundles by flowing steam initially at the maximum temperature attained in the reheater tube bundle, first through the shell side of the finishing superheater tube bundle, then through the shell side of the intermediate superheater tube bundle, with superheated steam meeting the requirements of the main steam turbine being produced at the tube side outlet of the finishing superheater tube bundle, before shell side steam flow continues to the reheat turbine. The initial superheater tube bundle stage, located inside of the nuclear pressure vessel, and the intermediate superheater tube bundle located, outside of the nuclear pressure vessel, are sized relative to each other so as to produce desired steam temperature and moisture requirements at the tube side outlet of the initial superheater tube bundle stage. Also, superheated steam can be diverted from the tube side outlet of the initial superheater tube bundle stage to plant feedwater heaters during continuous plant operation as a means to increase the ratio of reheat steam flow to main steam flow through the intermediate superheater and the finishing superheater tube bundles, thereby reducing the maximum steam temperature requirement at the tube side outlet of the reheater tube bundle. Reheat steam pressure is selected to be higher than reactor gas coolant pressure during continuous plant operation to prevent leakage of radioactive material into reheat steam. In summary the steam generator of the present invention includes a reheater tube bundle designed for subcritical pressure just above reactor gas coolant pressure, located inside of the nuclear pressure vessel, and a main steam system designed for subcritical or supercritical pressure comprising an economizer/evaporator tube bundle stage and an initial superheater tube bundle stage located inside of the nuclear pressure vessel, and a finishing superheater and an intermediate superheater tube bundle, contained in separate pressure vessels, located outside of the nuclear pressure vessel. A feature of the steam generator in accordance with the present invention lies in the ability to design the main steam system for subcritical or supercritical pressure operation. Another feature of the steam generator in accordance with the present invention lies in the ability to employ a reheater, thereby absorbing more heat from the reactor gas coolant and operating with higher main steam pressure than with prior steam generator designs. The reheater is designed for very high temperature to regeneratively superheat main steam flow through the intermediate superheater and the finishing superheater tube bundles which are located outside of the nuclear pressure vessel. Because reheat pressure is approximately equal to reactor gas coolant pressure, creep stresses in the reheater heat transfer tubes and other reheater pressure parts are negligible. Also, the very low pressure differential across reheater heat transfer tubes and other reheater pressure parts makes modern high temperature materials, such as graphite, ceramics and special alloy steels, feasible fop fabrication of the reheaters. The inclusion of a reheater in the steam generator system allows for significantly higher main steam system pressure and for the use of standard reheat and main steam turbines, thereby improving overall plant economics. Another feature of the steam generator in accordance with the present invention lies in the provision of water recirculation for tube side flow stability (positive up flow in all parallel main steam tube circuits) during start-up and low load operation, in which water is pumped from the initial superheater tube bundle stage outlet to the economizer/evaporator tube bundle stage inlet. This feature allows for enlargement or total elimination of flow stabilizing orifices as used in prior steam generators, resulting in reduction of steam side pressure loss at full load flow. Another feature of the steam generator in accordance with the present invention lies in the provision of a water cooling heat exchanger at the recirculation pump inlet to produce desired water temperature and to condense excess steam at the recirculation pump inlet. Another feature of the steam generator in accordance with the present invention lies in the ability to divert superheated steam from the outlet of the initial superheater tube bundle stage to plant feedwater heaters to increase the ratio of reheat steam flow to main steam flow through the finishing superheater and intermediate superheater tube bundles, thereby reducing the maximum steam temperature requirement at the reheater tube bundle outlet. Another feature of the steam generator in accordance with the present invention lies in the ability to size the heat transfer tube surface area of the initial superheater tube bundle stage with respect to the heat transfer tube surface area of the intermediate superheater tube bundle to produce liquid flow at the tube side outlet of the initial superheater tube bundle stage during low load and start-up operation of the plant. Another feature of the steam generator in accordance with the present invention lies in the ability to size the heat transfer tube surface area of the initial superheater tube bundle stage with respect to the heat transfer tube surface area of the intermediate superheater and the finishing superheater tube bundles to achieve the desired reheater tube bundle heat duty. Another feature of the steam generator in accordance with the present invention lies in the addition of steam side mixing locations in the piping between the initial superheater tube bundle stage and the intermediate superheater tube bundle, and between the intermediate superheater and the finishing superheater tube bundles, which act to equalize the temperature of steam emerging from the intermediate superheater and from the finishing superheater tube bundles. Equalized steam temperatures at tube bundle outlets promotes steam side flow stability and reduces tube overheating. Another feature of the steam generator in accordance with the present invention lies in the provision of a bypass system utilized during plant start-up, in which water is passed through a pressure reducing valve to a flash tank from which low pressure superheated steam is diverted to the tube side of the reheater tube bundle, and water is diverted to other plant systems. Still another feature of the steam generator in accordance with the present invention lies in locating the finishing superheater and the intermediate superheater tube bundles outside of the nuclear pressure vessel. The bi-metallic weld is then located outside of the nuclear pressure vessel where it is readily monitored and maintained. Tube replacement instead of tube plugging or tube bundle replacement is possible. Furthermore, reliability of steam generator components including the bi-metallic weld is improved. Further objects, advantages and features of the present invention, together with the organization and manner of operation thereof, will become apparent from the foregoing detailed description of the invention when taken in conjunction with the accompanying drawing wherein like reference numerals designate like elements throughout the several views.
	abstract	A patient's lesion is localized for the purpose of administering radiation treatment by obtaining a beam shape representation along one or more beam directions of a radiation treatment device. An image corresponding to the lesion is obtained from each beam direction, and the beam shape and image are fixed to a common coordinate system to facilitate alignment.
050664522	summary	FIELD AND BACKGROUND OF THE INVENTION The present invention relates in general to testing equipment and methods in nuclear power plants, and in particular to a new and useful method and apparatus for detecting and measuring wear on the control rods for the fuel assemblies for nuclear power plants. More and more frequently, control rod assemblies (CRA) are being thought of as components of pressurized water reactor (PWR) systems which require periodic non-destructive evaluation (NDE). In light of recent observations of breach-of-cladding, plant owners are under increasing regulatory pressure to determine the condition of their control components and to insure that shutdown margins are within specifications. Damage mechanisms include vibratory contact with support or guide components and absorber swelling due to neutron irradiation. As a result of either gross diameter increase in the cladding or leeching of the poison from the cladding, the ability to scram the reactor could be jeopardized. During the operation of a PWR, the control rods are suspended above the fuel assembly with the individual pins contained within the brazement guide structure. The tips of the pins are captured in guide tubes located in the fuel assemblies. Coolant flow through the guide tubes causes the pins to vibrate, and the resulting mechanical contact between the control rod pins and the support structure/fuel assembly guide tubes induces localized wear on the outside surface of the individual pins. There are at least two different flaw types in the CRA. The first flaw type is a large volume wear mark caused by contact with the guide tube nut. The second type of defect is a small volume, axial groove caused by contact with the brazement support structure above the fuel assembly. In the past, inspection of control rods, the primary component of rod cluster control assemblies (RCCAs) in pressurized water reactors, has been performed non-destructively, using various eddy current techniques. These eddy current techniques have been used to look for and measure cracks and wear marks that, if large enough, could render an RCCA unusable. The use of eddy current coils has been successful in determining the existence of a breach in cladding or the amount of material left on a control rod. In some situations, measuring the percent of material remaining has provided sufficient information for determining the useability of a control rod. In fact, pass/fail criteria has largely been based upon the kind of data (i.e. location, quantity and accuracy of data points) collected by eddy current inspection equipments. Inherent in the making of eddy current measurements and especially in the interpretation of the measurements is the problem of material variability. Because eddy current measurements are based on the electrical properties of the material of which the control rods are made, a control rod calibration standard must be produced to replicate the actual control rod and damage mechanisms as closely as possible. Any difference in material or defect geometry has the potential to be a source of error in data collected on actual control rods. If a very precise method of measurement that is not dependent upon material properties could be integrated in a system to inspect control rods, the accuracy of measurement data could be improved. SUMMARY OF THE INVENTION The present invention utilizes an ultrasonic non-destructive technique to detect wear on the control rods. The system is comprised of a rotating UT transducer mounting fixture which rests on a storage rack in the spent fuel pool of a nuclear power plant. This fixture houses and rotates a series of transducers while the control rod cluster is run through the fixture. The transducers maintain a constant distance from the control rods at all times as the transducers float to preserve transducer standoff. The system uses high frequency focussed transducers, multichannel ultrasonic thickness gauging instrumentation, a probe/electronics switching network, and a calibration standard. A conventional pulse-echo ultrasonic technique is used by the invention in conjunction with an immersion transducer to measure water path. Taking advantage of the velocity (longitudinal) difference between steel and water, a water path measurement provides a profile of the outside surface of the control rod by gauging the spacing between transducer and the test surface. This ultrasonic profilometry system uses the multichannel ultrasonic thickness gauge and multiple test stations to examine several control rods simultaneously. The transducers are rotated continuously around the control rods, maintaining constant spacing between the transducer and the individual control rods. The control rods are raised and lowered through the inspection fixture by the existing refueling mast in the plant, effecting a helical scan which provides maximum coverage. Both the pulse voltage and reflected signal are transmitted through slip rings, and the resulting data indicative of rod profile are recorded on computer-based oscillographic strip chart via the analog output of the thickness gauge. This ultrasonic technique has advantages over an electromagnetic method for this particular application for several reasons. It is not hindered by variations in the electrical properties of coatings applied to the outside surface of the control rods; it is unaffected by the presence of conductive material inside the control rods; and it is not dependent on representative calibration standard flaw geometry for sizing accuracy. The system can inspect 25% of the rods during one pass of the rod cluster control assembly (RCCA) through the UT transducer mounting fixture. To inspect the remaining rods the RCCA is removed from the fixture and rotated 90.degree.. The RCCA is then lowered back into the fixture and the rods inspected. This is repeated until all control rods have been inspected. The entire process is supported by any additional support tooling which itself, does not form a part of this invention. On-site activities can be performed off the side of the spent fuel pool or off a bridge in the spent fuel pool. Accordingly, an objective of the present invention is to provide an apparatus and method for scanning the outer profile of control rods, in particular, for nuclear fuel assemblies, comprising an ultrasonic transducer for sending an ultrasonic beam to, and for receiving an ultrasonic echo signal from the surface of a control rod, rotation means connected to the ultrasonic transducer for rotating the transducer around the axis of the control rod, and control rod translation means for translating the control rod with respect to the transducer, parallel to the axis of the control rod, for scanning the surface of the control rod along its length. A further object of the present invention is to provide an apparatus which can be used to measure the outer profile of control rods, in particular for detecting wear and surface defects, which is simple in design, rugged in construction and economical to manufacture. The various features of novelty which characterize the invention are pointed out with particularity in the claims annexed to and forming a part of this disclosure. For a better understanding of the invention, its operating advantages and specific objects attained by its uses, reference is made to the accompanying drawings and descriptive matter in which a preferred embodiment of the invention is illustrated.
042347980	description	SPECIFIC DESCRIPTION The shielding transport and storage container illustrated in the drawing and represented generally at 1 is formed with a vertically elongated compartment 1a adapted to receive irradiated nuclear fuel elements. Naturally, the chamber can also receive other radioactive wastes as may be required. Basically, the container or receptacle comprises a container wall or shell 2, a bottom 3 and a cover as will be described in greater detail below. According to the invention, the wall 2 laterally surrounding the chamber 1a, and the bottom 3 of the container are cast in one piece from cast iron, especially spherolytic cast iron, cast steel or the like. The receptacle is generally prismatic and of the configuration of a rectangle parallele-pipedon with rounded edges, i.e. has vertical faces 2a and 2b which are parallel to one another and vertical faces 2c and 2d which are also parallel to one another but are perpendicular to the faces 2a and 2b. The upper and lower end faces of the receptacle are also flat. The chamber 1a for receiving the radioactive waste, is formed with a shoulder 1b at its upper end upon which rests a lateral flange 4a of a shielding cover 4. Surrounded by this flange is a block 4b of shielding material which plugs the upper end of the chamber 1a, the cover 4 being recessed in the body formed by the wall 2 and the bottom 3 so that its upper surface 4c lies flush with an end face 2e of this body. According to the invention, the vessel wall 2 is provided in situ, i.e. during the casting process, with a tubular passage 5 which communicates between the low point 5a of the chamber 1a and the face 2e of the body 2, 3. To close this passage 5 at its upper end, which can be formed with an internally threaded formation 5b for connection to a pipe, the receptacle 1 is provided with a passage cover which is generally represented at 6. The latter is wholly received in a recess 2f formed at the upper end of the container. In the embodiment illustrated, the cover 6 not only closes the passage 5b, 5, but also overlies the cover 4. Bolts or screws 6a are threaded into the body 2, 3 and screws 4d passing through the flange 4a secure the cover 6 and 4 in place. In the bestmode embodiment of the invention, a further tubular passage 7 is cast in situ within the body 2, 3, the passage 7 communicating at 7a with the chamber 1a at its upper end, i.e. just below the plug 4b. At its opposite end, the passage 7 opens into a compartment 8 recessed in this body and terminates flush with the bottom 8a of this compartment. A valve 9 can have a flange 9a which is screwed to the body so as to sealingly couple this valve with the passage 9. The valve 9 is wholly received in the compartment 8 which, in turn, is closed by the cover 6. Thus the cover 6 is screwed to the body 2, 3 outwardly of the regions in which the passages 5 and 7 open at the surface 2e of the body. The valve 9 may be a fluid control valve or a pressure relief or safety valve which can be vented, e.g. through an additional tube. As is apparent from FIGS. 1 and 2, the passages 5 and 7 lie within the inner half of the thickness T of the wall 2. This permits the outer half of the wall thickness to receive passages 10 which, as shown in FIG. 2, are of oval cross section and extend the full length of the vessel. These passages receive radiation shielding material 10a, e.g. graphite or some other neutron moderator or a heavy metal such as lead. The additional moderator 10a is especially advantageous when the radioactive wastes have a high neutron emission. The channels 10 can also be closeable by the cover 6. Of course, the cover 4, 6 can be connected together e.g. by screws. The outer periphery of the receptacle is provided with cooling ribs 11 which are cast in situ and in one piece with the body 2, 3. The ribs 11 are shown to extend along generatrices of the body and to have gaps or cutouts 12 which facilitate expansion and contraction. Naturally, the ribs may also lie in horizontal planes as desired.
048308161	abstract	A getter trap to remove hydrogen and oxygen from a liquid metal, such as liquid sodium, includes an elongated, closed housing having an inlet at one end thereof and an outlet at the other end. A getter material is randomly diposed within the housing comprising a zirconium-containing substrate of hollow, tubular sections having a coating thereon of a gettering alloy of zirconium, vanadium and iron. As a liquid metal flows through the inlet into the housing and through the getter material, and is discharged from the housing through the outlet, hydrogen and oxygen impurities are removed from the liquid metal. The getter trap is particularly useful in an improved liquid metal cooled nuclear reactor system.
	claims	1. An apparatus, comprising: at least one heat radiation activated deployable structure; and a heat shield assembly that shields the deployable structure in a first position and that exposes the deployable structure to heat radiation in a second position. 2. The apparatus according to claim 1, wherein the heat radiation activated deployable structure is a solar radiation activated deployable structure. 3. The apparatus according to claim 1, wherein the heat radiation activated deployable structure has a base, and wherein the heat radiation activated deployable structure has a plurality of petals, each of the petals of the plurality of petals having a first end operatively coupled to the base. 4. The apparatus according to claim 3, wherein the petals are in a reduced size configuration in a first shielded position and in an extended configuration in a second exposed position. 5. The apparatus according to claim 4, wherein the petals in the first shielded position are shielded from heat radiation and in the second exposed position absorb heat radiation. 6. An apparatus, comprising: a plurality of structural elements that are operatively coupled to a base structure; and a heat shield assembly that shields the structural elements in a first position and that exposes the structural elements to heat radiation in a second position. 7. The apparatus according to claim 6, wherein the structural elements are petals, and wherein the petals and base structure form a star shade. 8. The apparatus according to claim 7, wherein the petals are in a reduced size configuration in a first shielded position and in an extended configuration in a second exposed position. 9. The apparatus according to claim 8, wherein the petals in the first shielded position are shielded from heat radiation and in the second exposed position absorb heat radiation. 10. The apparatus according to claim 7, wherein each of the petals are approximately 25 meters long and may taper to a point in an extended configuration. 11. The apparatus according to claim 7, wherein each of the petals is approximately 0.15 inches thick. 12. The apparatus according to claim 7, wherein each of the petals has a curved shape across a width thereof to provide a degree of structural stability. 13. The apparatus according to claim 6, wherein each of the structural elements is formed from a thin elastic memory composite material. 14. A method, comprising: heating up at least one structural element beyond a change state temperature thereof; changing the configuration of the structural element from an extended configuration to a reduced size configuration; cooling the structural element to below the change state temperature thereof; and covering the structural element with a thermal protection device; removing the thermal protection device to expose the structural element to heat radiation; and heating, via the heat radiation, at least a portion of the structural element to thereby cause the structural element to change from the reduced size configuration to the extended configuration. 15. The method according to claim 14, wherein the step of changing the configuration of the structural element from an extended configuration to a reduced size configuration comprises rolling the structural element onto a mandrel. 16. The method according to claim 15, wherein the step of heating, via the heat radiation, at least a portion of the structural element to thereby cause the structural element to change from the reduced size configuration to the extended configuration comprises successively heating different areas of the structural element such that the structural element unrolls from the mandrel. 17. The method according to claim 14, wherein the structural element in a first shielded position is shielded from heat radiation by the thermal protection device and in a second exposed position, by at least partial removal of the thermal protection device, absorbs heat radiation. 18. The method according to claim 14, wherein the structural elements are petals, and wherein the petals are attached to a base structure and form a star shade in the extended configuration. 19. The method according to claim 18, wherein each of the petals is approximately 0.15 inches thick. 20. The method according to claim 18, wherein each of the petals has a curved shape across a width thereof to provide a degree of structural stability. 21. An apparatus comprising:a heat activation deployable structure formed of a material having the properties of an Elastic Memory Matrix (EMC) material; anda heat shield assembly adapted and constructed to shield the heat activation deployable structure in a first position and to expose the heat activation deployable structure to heat radiation in a second position, wherein the heat activation deployable structure is deployably and selectively connected to the heat shield assembly.
	summary
043354661	summary	In order to comply with various safeguards agreements, inspection organizations such as NRC (Nuclear Regulatory Commission) and IAEA (International Atomic Energy Agency) need a capability of very quickly and accurately monitoring in a non-destructive manner the fissile content of spent fuel assemblies in storage pools. Presently, measurements of the content of residual and produced fissile material are not directly measured but rather are inferred by measuring particular data which is correlated to burnup (which is a measure of nuclear reactor fuel consumption, expressed either as a percent of fuel atoms that have undergone fission or as the amount of energy produced per unit weight of fuel). It is known in the art that the amounts of certain fission products which are present in a fuel assembly, such as Cs-137, .sup.144 Ce-Pr, and Ru-106, are proportional to burnup and can be used as burnup monitors. See, for example, S. T. Hsue et al., Los Alamos Scientific Report LA-6923 (ISPO-9) (1978). It is known in the art that the gross gamma activity of a spent fuel assembly depends both upon the cooling time (i.e., time measured from discharge from the reactor) and upon the intensity of various fission product gamma rays. The gamma rays from a spent fuel assembly can be divided into two categories, (1) gamma rays from direct fission products and (2) gamma rays from isotopes resulting from neutron capture on direct fission products. The number of the type (1) gamma rays is known to be proportional to the reactor neutron flux; and the number of the type (2) gamma rays is known to be approximately proportional to the square of the reactor neutron flux. However, only type (1) gamma rays have been found to be proportional to burnup. Thus, a gross gamma activity measurement of a spent fuel assembly will not in general be expected to give an accurate measurement of burnup due to the possible interference of the type (2) gamma rays, described above. In all detectors in which gross gamma activity is measured, the detector response is proportional to the sum of the gamma rays emitted, which depends upon cooling time and intensities of the fission products (which depend upon burnup and operating history). However, corrections to the data for these factors is not generally possible because the relative contribution of each factor is not known. Thus, it generally cannot be known prior to experimental determination or complicated calculations when, if ever, a gross measurement of total emitted gamma rays will agree with the true burnup. Typically, the preferred method for measuring relative burnup has been to use high resolution gamma ray spectroscopy (HRGRS) and to perform a series of measurements of the intensity of gamma rays having a particular energy at various points along the length of the fuel assembly and then to use the integrated area of that profile and an established calibration curve of calculated burnup vs. integrated Ge detector response (measuring, for example, the 661 keV gamma ray of Cs-137) to provide the corresponding burnup value. The use of a germanium detector to monitor the intensity of the 661 keV gamma ray of Cs-137 as a function of axial position along a fuel assembly provides a very accurate (2-6%) measure of relative burnup but takes a long period of time and requires a multichannel analyzer system, a mechanical scanning system, and a collimator assembly. An alternative scanning technique is to employ a cadmium telluride detector for the profile measurements and then to calibrate the profile by use of a germanium detector for a gamma-ray absolute intensity measurement, normally at one point in the center of the profile. Although both of these techniques provide statistically satisfactory data, both require a quite long period of time for the measurements, often one hour or longer per assembly, and both require collimators. A 1965 publication entitled Richard J. Nodvik, "Evaluation of Gamma Scanning as a Tool for Determining Fuel-Burnup Distribution in Large Power-Reactor Cores," Transactions, 1965 Annual Meeting, American Nuclear Society, described the use of a miniature ion chamber inserted in-core during reactor operation for gross gamma scanning, (a technique which was being evaluated as a tool for determining burnup distributions within large power reactor cores). However, although that reference initially mentioned the term "distribution," there was no further discussion of the subject. And it was found that the gamma activity generally overestimated the burnup in assemblies that occupied the central region of the core (where higher burnup normally occurs) and underestimated the burnup in assemblies that formed the periphery of the core (where lower burnup normally occurs), implying that a measured profile would be flatter than the true burnup profile. The agreement between the gross gamma intensity and burnup was not good, deviations having ranged from -16 to +13%. The agreement must be good at every point in the profile in order to get a good measure of burnup. Therefore, in view of the above, a single ionization chamber would probably not be expected to be very useful in a method of accurately measuring burnup. Furthermore, the uses of the apparatus of this invention in rapidly measuring burnup and rapidly measuring an identifying characteristic which is used to determine whether a fuel assembly has been tampered with would be unobvious. And although two ionization chambers (each anode having a plurality of wires) have been used in measuring profiles in two coordinates of particle beams, (as described in C. K. Hargrove et al., "A Multiwire Proportional Chamber System for Monitoring the Position and Profile of a Charged Particle Beam," Nuclear Instruments and Methods, 113 (1973), pp. 141-145), the more versatile and less cumbersome apparatus of this invention has not previously been known. SUMMARY OF THE INVENTION An object of this invention is an apparatus for and another object is a method for measuring data directly correlatable with the burnup profile of a reactor fuel assembly in a period of measurement time which is less than 10 seconds, rather than nearly an hour or more as is required in the prior art apparatus described above. Other objects of this invention are a method and apparatus for determining within 10 seconds whether a fuel assembly has been tampered with. Additional objects, advantages, and novel features of the invention will be set forth in part in the description which follows and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be utilized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects and in accordance with the purposes of the present invention, as embodied and broadly described herein, the apparatus of this invention may comprise: a multiplicity of spaced apart substantially identical ionization chambers or proportional chambers, the individual chambers being operably connected so as to provide a multielement detector having a capability of substantially instantaneously obtaining a profile of data which is directly correlatable with burnup as a function of axial position. Further according to the invention in another embodiment, the multielement detector of the invention is used to substantially instantaneously and nondestructively measure a profile directly correlatable with the burnup profile of an object, for example a spent fuel assembly, with an accuracy equivalent to that of a germanium detector by measuring the gross gamma activity profile with the detector located outside the core of the reactor after a cooling time as short as 9 months and at a voltage such that saturation of the chambers does not occur. In yet another embodiment, the profile substantially instantaneously obtained by using the multielement detector of the invention is used to determine whether a particular object, for example a fuel assembly, has been tampered with. The apparatus according to the invention exhibits the following combination of advantages. It has the capability of being used to obtain a relative gross gamma activity profile measurement (which can be used to identify a particular fuel assembly, much like a fingerprint) in a very short period of time, less than 10 seconds. And unexpectedly, it has been found that the integrated area of the normalized gross gamma activity profile obtained with the multielement detector agrees to within the statistics of the normalized profile obtained by employing a germanium detector to measure the intensity of the 661 keV gamma ray of Cs-137 at a multiplicity of axial positions, using a cooling time as short as 9 months, provided that the detector is used out-of-core and provided that saturation of the detector does not occur. And, if desired, an absolute burnup profile can be obtained in a few minutes using the multielement detector if a germanium detector is additionally used to make one measurement for calibration of the normalized profile (referred to above). The apparatus of the invention, furthermore, can operate in both the ionization range and in the proportional range. And, furthermore, no problems which are intrinsic in mechanical scanning are encountered with the apparatus of the invention. The multielement detector is quite versatile, allowing one to measure long and short fuel assemblies with one convenient device, adjustable by varying the number of individual chambers and by varying the spacings between chambers. This device is less cumbersome than a large fixed-size detector employing one chamber with an anode made from a multiplicity of wires. Unlike multiwire detectors, no sophisticated construction techniques are required; and because individual detectors are used, repair is made easier. Additionally, the electronics setup which is used in cooperation with the multielement detector is much simpler than that needed with HRGRS, can be made portable, and may even be battery powered.
056278659	summary	FIELD OF THE INVENTION The present invention relates generally to nuclear fuel assemblies for nuclear reactors, and more particularly to fuel rod configurations for fuel assemblies that have a square cross-sectional area. BACKGROUND OF THE INVENTION Current operating light water reactors (LWR) utilize fuel assemblies that have a square cross-sectional area in which the nuclear fuel rods are located. Light water reactor designs employ a square array for the layout for control rod drives and consequently the area allocated for fuel assemblies is square. The fuel rods are distributed in the available square area so that there will be an approximately uniform distribution of coolant/moderator area for each fuel rod. The approach has been to arrange the fuel rods within the available square area so that there was an equal number of rows and columns of fuel rods with a uniform center-to-center distance (i.e. pitch) between fuel rods. This arrangement is referred to as a square lattice, as lines drawn through adjacent fuel rod centers divide the area into a number of uniform squares. The reactor power and power distribution (axial, radial and local peaking) set the volumetric power density generated in the fuel rods. The minimum spacing between fuel rods to assure adequate cooling of adjacent fuel rod surfaces, which has been determined by heat transfer tests, must be provided with allowance for manufacturing tolerances and predicted fuel rod bowing during operation. For a uniform array of fuel rods, the required minimum rod-to-rod spacing limits the maximum allowable fuel rod diameter for that array. Uniform distribution of uranium fuel and coolant moderator (i.e. water) has been typically obtained by selecting an equal number of rows and columns of fuel rods in a square lattice array and positioning the center of the nuclear fuel rods at the corners of the squares. Thus, the number of rows of fuel rods equal the number of fuel rods in a row. The fuel rod array is sized to obtain sufficient heat transfer area for the volume of nuclear fuel in a fuel rod to enable the removal of the heat generated by the fuel within temperature limits of the materials used for the fuel rod. Boiling water reactor (BWR) fuel assemblies typically have such a fuel rod array in which the fuel rods are arranged in rows with the same number of fuel rods in each row as there are rows in the array. In adjacent rows, fuel rods are located with their centers at the corners of squares. Such square rod arrays or lattices are commonly named according to the number of rows of rods and number of rods in a row such as 8.times.8, 9.times.9, 10.times.10, etc. Regardless of the number of rows of rods, each array is constrained to fit within a standard size fuel assembly channel. The use of a square lattice whereby fuel rods are located with their centers at the corners of squares results in a larger flow area at the center of the square formed by four fuel rods than is necessary. This is an inefficient use of the cross-sectional area within a fuel assembly channel. It is desirable to reduce the fuel rod linear heat generation rate and the internal fuel rod temperature for a given fuel assembly power level by increasing the number of fuel rods. This is done, for example, by changing from a 10.times.10 fuel rod array to an 11.times.11 array. Since the fuel rod array is constrained to fit within the fixed dimensions of a standard fuel assembly channel and is required to have a certain minimum fuel rod surface to surface and fuel rod surface to channel wall surface spacing, increasing the number of rows of fuel rods and number of fuel rods in a row necessitates a decrease in the fuel rod diameter. The fuel rod diameter must be reduced to maintain surface to surface spacing since the fuel rod center to center distance is reduced. The spacing between rods to allow for adequate cooling and to accommodate fuel rod bow cannot be reduced in proportion to the rod-to-rod pitch. As the quantity of the fuel rods is increased in a square lattice, the increased number of fuel rods will not compensate for the required fuel rod diameter reduction with the result that the uranium loading in the fuel assembly is reduced in the finely divided array. For example, a 10.times.10 square lattice array would have a rod pitch of approximately 0.51 inch and a minimum rod surface to rod surface space that would allow for manufacturing tolerances, and rod bow to maintain adequate cooling throughout the operating life. If such a space were 0.114 inch, then the maximum rod diameter could be 0.396 inch. If the square lattice array was more finely divided to an 11.times.11, then the rod pitch would be approximately 0.464 inch. The maximum rod diameter would be limited to 0.35 inch to maintain the required 0.114 inch space between rods. The amount of space for fuel is proportional to the number of rods and their cross sectional area. The relative fuel cross sectional area for the two arrays would be ##EQU1## In BWR fuel assemblies, a number of fuel rod locations are reserved for use instead as water rods or a water channel to selectively increase neutron moderation for more efficient fuel utilization. If the square fuel rod array is more finely divided and if the number of reserved water rod locations remains constant, then the amount of moderating water within the water rods or water channel becomes smaller because of the smaller allowable diameter for both the fuel rods and water rods. If the number of reserved rod locations for water rods is increased as the array size is more finely divided, then the uranium loading for the fuel assembly is decreased even further. Thus, as the square fuel rod array is more finely divided and the number of water rods either increases or remains unchanged, inefficient fuel utilization as well as high fabrication costs result. A triangular lattice array in which the centers of fuel rods are located at the vertices of a triangle is more desireable than the square lattice array in that it provides a more efficient arrangement of fuel rods while also maintaining required rod-to-rod spacing. For a specified fuel rod diameter and minimum rod-to-rod spacing, the triangular lattice allows a tighter packing of fuel rods within the given cross sectional area of the fuel assembly channel, resulting in a better allocation of area for flow of coolant water among fuel rods. The higher density of fuel rods will permit a higher loading of uranium, and better heat transfer characteristics as the coolant water is on the average in closer proximity to the fuel rod surfaces. In addition, more fuel rod heat transfer surface can be incorporated in a unit area than in a square lattice array of the same pitch, and greater flexibility for internal moderation using water rods and inner water channels can be obtained. Since the higher density of fuel rods permits a higher loading of uranium as the number of fuel rods in the assembly is increased, more fuel rod locations can be reserved for water rods or water channels without causing a decrease in the uranium loading in comparison to a square lattice array which will have fewer fuel rod positions. A triangular lattice however cannot be made to fit into a square cross-sectional area by having an equal number of rows and columns of fuel rods. It is an object of the invention to have a fuel rod arrangement in which a triangular lattice is utilized for fuel assemblies that are square. SUMMARY OF THE INVENTION In accordance with one aspect of the invention, a nuclear fuel assembly for boiling water reactors is provided having a plurality of elongated parallel fuel rods supported between a lower tie plate positioned toward the bottom of the assembly and an upper tie plate positioned toward the top of the assembly, an outer channel surrounding the plurality of fuel rods and having a substantially square cross-sectional area for conducting coolant/moderator about the fuel rods from the bottom of the assembly toward the top of assembly, at least one spacer for positioning and retaining the fuel rods in a predetermined configuration, and the fuel rods being arranged with a predetermined pitch in an array where the centers of the fuel rods are located at the vertices of isosceles triangles.
	summary
	summary
	claims	1. A method of measuring a dimension of a circuit pattern, which is formed on a substrate, using a scanning electron microscope (SEM), comprising the steps of:(a) inputting a center coordinate of an SEM image area and design information of a circuit pattern;(b) setting a measurement object area including an edge of the circuit pattern having a dimension to be measured, using the center coordinate of the SEM imaging area and the design information input, and setting an imaging area and imaging condition for imaging an area including the measurement object area with the scanning electron microscope;(c) setting an imaging sequence for imaging the imaging area with the scanning electron microscope for measuring the dimension of the circuit pattern;(d) imaging the circuit pattern formed on the substrate with the scanning electron microscope based on the imaging condition and the imaging sequence; and(e) processing the image obtained by imaging to measure the dimension of the circuit pattern,wherein, step (b) includes the steps of setting, as the measurement object area, an area including the edge of the circuit pattern in the vicinity of the position at which the dimension of the circuit pattern is measured, and setting in accordance with a direction of the edge of the circuit pattern included in the area, a direction of continuous scanning of an electron beam scanned in the scanning electron microscope. 2. An apparatus adapted to measure a dimension of a circuit pattern formed on a substrate using a scanning electron microscope, comprising:input means for inputting a center coordinate of an SEM image area and design information of a circuit pattern;imaging condition setting means including a measurement object area setting section adapted to set a measurement object area including an edge of the circuit pattern having a dimension to be measured, using the center coordinate of the SEM imaging area and the design information input, and an area/condition setting section adapted to set an imaging area and imaging condition for imaging an area including the measurement object area with the scanning electron microscope;imaging sequence setting means for setting an imaging sequence for imaging the imaging area, which is set by the imaging condition setting means for measuring the dimension of the circuit pattern, with the scanning electron microscope;scanning electron microscope means for imaging the circuit pattern formed on the substrate based on the imaging condition set by the imaging condition setting means and the imaging sequence set by the imaging sequence setting means; andimage processing means for processing the image obtained by imaging with the scanning electron microscope means to measure the dimension of the circuit pattern,wherein, the measurement object area setting section of the imaging condition setting means sets, as an area including a position at which the dimension of the circuit pattern is measured, an area including the edge of the circuit pattern in the vicinity of the position at which the dimension of the circuit pattern is measured, and the imaging condition means further includes a scanning direction setting section adapted to set a direction of continuous scanning of an electron beam scanned in the scanning electron microscope in accordance with a direction of the edge of the circuit pattern included in the area set by the measurement object area setting section.
	summary
	abstract	Described is a system and method for in situ sample preparation and imaging. The system includes a multi-axis stage 100 having a bulk stage 110 and a grid stage 150 with various degrees of freedom to allow for sample preparation. In some embodiments, a focused ion beam system is used to prepare a lamella on the bulk stage 110. The lamella can then be transferred to the grid stage 150 from the bulk stage 110 without needing to move the multi-axis stage 100 from the focused ion beam system.
	description	This invention relates generally to manufacturing methods of semiconductor devices, and more particularly to the control of implantation processes. Ion implant is a critical technology in the fabrication of semiconductor devices. Ion implanters are typically used for performing ion implantation processes. The ion implanters are used to provide doping for the semiconductor devices, wherein impurity atoms are introduced to change the electrical properties of semiconductor materials. A typical ion implantation process includes the steps of creating ions of the desired impurity atoms, using electric fields to accelerate the ions to a required energy, transporting the ions down a beam line to a silicon wafer, and scanning the beam, or moving the silicon wafer, or both, such that uniform dosage in the silicon wafer is accomplished. All semiconductor fabrication processes use many (typically 15 to 25) steps of ion implantation to create a completed semiconductor device. The primary parameters of ion implantation are species, energy and dose. The species are the types of atoms being implanted, and there are two main categories, N-type and P-type, which are denoted by the electrical activity of the impurity in the semiconductor materials. N-type dopants are usually arsenic or phosphorus and P-type dopants usually include boron. The energy determines how deep into the silicon wafer the ions will go: high energy implants are deep, while low energy implants are shallow. The dose determines how conductive the implanted region will be. All of these parameters are chosen by the transistor designer for each implant step to optimize the device characteristics. A basic ion implantation system is schematically shown in FIG. 1, which includes an implanter comprising an ion beam generator, a beam manipulation unit, and a process chamber. In the ion beam generator, plasma of the desired implant species is produced, from which ions are extracted. The resulting ion beam may then be passed through a magnet (not shown), so that ions of a particular mass to charge (m/e) ratio are selected. The ion beam that emerges from the magnet is further accelerated during the beam manipulation stage. The ion beam is then projected on the silicon wafer in the process chamber. Since ions are electrically charged, and only ions with a certain number of charges are selected in a given implantation process, by determining the charges landing on the wafer, the ion dosage can be determined. The dosage determination is typically performed by a dose integrator electrically connected to the wafer, wherein the dose integrator determines the accumulated dosage implanted on the wafer. A schematic diagram of a dose integration process performed by a dose integrator is shown in FIG. 2, wherein a wafer current, which is generated due to the charges carried by ions, is input into the dose integrator and integrated with respect to time. The accumulated charge amount is converted to an accumulated dosage and is used to control the implanter. In a typical implantation process, a target dosage is predetermined. The dose integrator determines the accumulated dosage by multiplying the current flowing through the dose integrator by the duration for which the current flows. The accumulated dosage is then compared to the target dosage. As soon as the target dosage is reached, the implantation process stops. The dose integrator typically comprises common circuit components such as resistors, capacitors and operational amplifiers. Therefore, with time, the circuit components may degrade and the detected accumulated dosage may drift from the actual value. As a result, the implanted dosage will deviate from the predetermined value. For example, assume one coulomb of charges is expected by the dose integrator to provide the target dosage, and the charges are supplied by one-ampere current flowing through the dose amplifier for one second. If the dose integrator drifts, and the one-coulomb charges are wrongfully determined to be 0.8 coulombs, the implantation will then last 1.25 seconds instead of one second to make up the difference. As a result, 1.25 coulomb charges actually pass the dose integrator, which means that the corresponding dosage is also 25 percent more than necessary. This will cause degradation of, or even failure of, the resulting integrated circuits on the wafer. Although dosage integrator drift does not occur frequently, when it occurs, the respective cost is high. Conventionally, periodic monitoring is the only way to catch this problem. However, periodic monitoring cannot catch the problem in real time, and many wafers may be damaged during the period of time from when the drift occurs to when the problem is found by e.g., daily monitoring. For example, up to about 1000 wafers may be manufactured and damaged due to the drift of the dosage integrator. Therefore, a method for catching dosage drift in real time to prevent high loss is needed. In accordance with one aspect of the present invention, an implantation system includes a first dose integrator and a second dose integrator. The first dose integrator includes a first input configured to receive a first current generated from charges carried by implanted ions in a wafer, and a first output configured to output a first accumulated dosage value. The second dose integrator includes a second dose integrator including a second input configured to receive a second current generated from the charges carried by the implanted ions in the wafer, and a second output configured to output a second accumulated dosage value. The implantation system further includes a processing unit comparing the first accumulated dosage and the second accumulated dosage to determine a drift in one of the first and the second dose integrators. In accordance with another aspect of the present invention, an implantation system includes an implanter, a current divider having a current input and a first divider current output and a second divider current output, a first dose integrator coupled to the first divider current output and having a first accumulated dosage output, a second dose integrator coupled to the second divider current output and having a second accumulated dosage output, and a processing unit coupled to the first accumulated dosage output and the second accumulated dosage output. In accordance with yet another aspect of the present invention, a method for detecting a drift of a dose integrator in an implantation system includes connecting a current divider into a current path wherein the current path is connected to a wafer in an implanter, implanting ions into the wafer, conducting a wafer current generated from ions in the wafer to a current divider wherein the current divider divides the wafer current into a first portion and a second portion, inputting the first portion of the wafer current into a first dose integrator wherein the first dose integrator outputs a first accumulated dosage output, inputting the second portion of the wafer current into a second dose integrator wherein the second dose integrator outputs a second accumulated dosage output, and comparing the first accumulated dosage output and the second accumulated dosage output to determine a drift in one of the first and the second dose integrators. With two dose integrators connected to one wafer current source, if one dose integrator drifts, the problem can be caught in real time. The possible loss due to over-dosage or under-dosage is thus avoided. The making and using of the presently preferred embodiments are discussed in detail below. It should be appreciated, however, that the present invention provides many applicable inventive concepts that can be embodied in a wide variety of specific contexts. The specific embodiments discussed are merely illustrative of specific ways to make and use the invention, and do not limit the scope of the invention. Dose integrators are key elements for dosage control. The dosage provided by an implanter is determined by the dose integrator. Any dosage monitoring based on the value provided by the dose integrator is unlikely to serve its purpose if the dose integrator itself drifts, and a reference device external to the sole dose integrator must be used to monitor the performance of the dose integrator. A schematic diagram of the preferred embodiment of the present invention is shown in FIG. 3. Note that the diagram is only used for explaining the concept of the present invention, and the present invention may have a variety of different embodiments, hence different connections. In the preferred embodiment, the wafer current I, which is generated from charges carried by ions implanted into the wafer, is input into a current divider. The current divider divides the current I into two substantially equal currents, a current IA and a current IB, with the currents IA and IB having a difference ΔI of less than about one percent of the wafer current I. Current IA and current IB are input into a dose integrator A and a dose integrator B, respectively. Accumulated dose output DA and accumulated dose output DB, which are the outputs of the dose integrators A and B, respectively, are then processed, preferably by a processing unit. The processing unit is signally connected to the implanter. Although the processing unit is shown as a separate device, it may be a functional component integrated into the implanter or a dose integrator, etc. A schematic diagram of an exemplary workflow of a dose integration process is shown in FIG. 4, which illustrates that after the wafer current I is divided, each of the currents IA and IB are continuously sent through separate paths and are accumulated by the dosage integrators A and B, respectively. During the entire implantation process, the accumulated dosage outputs DA and DB are compared. If no drift occurs in either of the dosage integrator A and dosage integrator B, the accumulated dosage outputs DA and DB will be substantially identical, and the match between the accumulated dosage outputs DA and DB depends on the precision of the current divider and the precision of the dosage integrators A and B. In the preferred embodiment, the accumulated dosage outputs DA and DB are added by the processing unit. The combined dosage output (DA+DB) represents the total dosage implanted into the wafer up to the moment. The total dosage (DA+DB) is then compared to a target dosage. When the target dosage is reached, the implantation process will be stopped. If, however, one of the dosage integrators A and B drifts, the accumulated dosage outputs DA and DB are likely to have a difference that is greater than a maximum allowed error. The maximum allowed error is pre-determined by the user of the implanter, and may be determined based on the design requirements. The difference in accumulated dosage outputs DA and DB will be detected by the processing unit, which is signally connected to the implanter. The implantation process will then be stopped, and the cause of the difference will be determined. The dosage integrator having drift will be replaced. If the difference in accumulated dosage outputs DA and DB is less than the maximum allowed error, the implantation process will continue. Although the maximum allowed error can be an absolute dosage value specified for a particular implantation process, a relative maximum allowed error is more preferable. In an exemplary embodiment, the maximum allowed error is about one percent. Accordingly, a relative dosage difference, which is preferably represented by a percentage, is calculated. In an exemplary embodiment, the relative dosage difference is expressed as \|DA−DB\|/(DA+DB). In the previously discussed embodiment, two dose integrators are used, and each acts as a reference to the other. As dose integrators are relatively reliable devices, the likelihood of both dose integrators drifting at the same time is very small, thus the reliability of an implantation system using two dose integrators is very high. Assuming at any moment the probability of one dose integrator drifting is 1/P, wherein P is typically a large number, the probability of two dose integrators drifting at the same time is 1/(PP). Furthermore, unless two dose integrators drift in the same direction (either higher than the actual value or lower than the actual value) and have the same magnitude, the drifting is still likely to be detected. Therefore, the probability of undetected drift is less than about 1/(PP). Assuming P is 10,000, 1/(PP) is as high as 1/108. One will realize that the preferred embodiment of the present invention has many variations. In other embodiments, more than two dose integrators are used. FIG. 5 illustrates three dose integrators used in the implantation system, although more dose integrators can be used. The wafer current I is preferably divided into three equal shares, each being fed into a dose integrator, namely A, B and C. The accumulated dosage outputs DA, DB and DC are then combined as the total dosage. Similarly, by comparing the accumulated dosage outputs DA, DB and DC, any drift in the dosage integrators A, B and C will be found. It is appreciated that the reliability of an implantation system having three dose integrators is even higher than an implantation system having two dose integrators. However, cost is higher also. One skilled in the art will realize that although not shown in the drawing, implantation systems may have three or more dose integrators if very high reliability is required. In yet other embodiments, after the current I is divided, currents IA and IB(refer to FIG. 3) are amplified by the current divider to the original value of I, and the amplified currents are input into the dosage integrators A and B. Preferably, only the output of one of the dosage integrators is used to determine the process end point, and the other dosage integrator is only used for monitoring purposes. Therefore, in these embodiments, the current divider may generate currents that, when added, do not equal to the wafer current I. In yet other embodiments, the current divider may divide current I into non-equal shares. A calibration process is preferably performed by the processing unit, which adjusts the comparison process accordingly. For example, a current divider divides the current I into a current equaling 0.49I, which is input into the dose integrator A, and a current equaling 0.51I, which is input into the dose integrator B. In the calibration process, the ratio of the unevenly divided currents is identified. The relative difference of the accumulated dosages DA and DB is then adjusted accordingly to be \|DA−0.49/0.51DB\|/(DA+DB). An advantageous feature of the non-equal current dividing is that with a calibration process being performed, low precision and low cost current dividers can be used without compromising the precision of the implantation system. The preferred embodiments of the present invention significantly improve the reliability of implantation systems. The cost of the extra dose integrator is low compared to the likely loss due to the damaged wafers. For example, if daily monitoring is performed, an implanter may produce about 1000 defective wafers due to the dose integrator drift between two monitoring, which will result in a loss of over one hundred times the cost of the additional dose integrator. Although the present invention and its advantages have been described in detail, it should be understood that various changes, substitutions and alterations can be made herein without departing from the spirit and scope of the invention as defined by the appended claims. Moreover, the scope of the present application is not intended to be limited to the particular embodiments of the process, machine, manufacture, and composition of matter, means, methods and steps described in the specification. As one of ordinary skill in the art will readily appreciate from the disclosure of the present invention, processes, machines, manufacture, compositions of matter, means, methods, or steps, presently existing or later to be developed, that perform substantially the same function or achieve substantially the same result as the corresponding embodiments described herein may be utilized according to the present invention. Accordingly, the appended claims are intended to include within their scope such processes, machines, manufacture, compositions of matter, means, methods, or steps.
062360558	claims	1. An article irradiation system, comprising a radiation source positioned for scanning a target region with radiation, a conveyor system, including a process conveyor, positioned for transporting articles in a substantially closed loop including the target region, radiation shielding material defining a chamber which substantially encloses the substantially enclosed loop and which encloses the radiation source, the target region and a portion of the conveyor system, wherein the radiation source is disposed on a particular axis inside the substantially closed loop defined by the conveyor system and is adapted for scanning the articles being transported in the substantially closed loop including the target region with radiation scanned in a plane transverse to the direction of transport of the articles by the process conveyor in the target region, and an intermediate wall of radiation shielding material positioned within the substantially closed loop in a direction transverse to the particular axis, the intermediate wall being separated in the transverse direction by air gaps from the radiation shielding material defining the chamber and being provided with dimensions in the transverse direction to inhibit radiation from the radiation source from reaching the radiation shielding material defining walls of the chamber. wherein the chamber-defining radiation shielding material includes a lateral wall that is disposed outside the substantially closed loop and that defines with the chamber-defining radiation-shielding material the path outside of the substantially closed loop; wherein the lateral wall inhibits any radiation in the path outside of the chamber from flowing past the lateral wall. a beam stop of a material for absorbing electrons and for converting the energy of the absorbed electrons into photons that are emitted from the beam stop, wherein the beam stop is disposed in a particular wall of said chamber-defining radiation shielding material adjacent the target region, and wherein the intermediate wall is positioned between the beam stop another wall of said chamber-defining radiation shielding material on the opposite side of the chamber from the wall adjacent the target region and is provided with dimensions in the direction transverse to the particular axis so that photons emitted into the chamber from the beam stop are inhibited from impinging upon the other wall. wherein a second portion of the conveyor system is positioned for transporting articles in a path that is outside of the substantially closed loop but continuous with the substantially closed loop; wherein the chamber-defining radiation shielding material includes a lateral wall that is disposed outside the substantially closed loop and that defines with the chamber-defining radiation-shielding material the path outside of the substantially closed loop; and wherein the lateral wall inhibits any radiation in the path outside of the chamber from flowing past the lateral wall. a chamber defined by walls made from a radiation shielding material, a radiation source constructed to provide radiation in the chamber, a conveyer system constructed to carry the articles in a loop through the chamber for the reception of the radiation in the chamber by the articles, first means disposed in the chamber for receiving radiation from the source and for converting the radiation to photons movable into the chamber, and second means disposed within the loop in the chamber and separated by air gaps from the walls defining the chamber and provided with dimensions relative to the walls defining the chamber and disposed relative to the first means for inhibiting the photons from the first means from impinging on the walls defining the chamber, thereby providing for a reduction in the thickness of the walls defining the chamber. the second means is disposed within the loop in the chamber to minimize the intensity of the photons and includes an intermediate wall separated by air gaps from the walls defining the chamber. the radiation source extends through the second means and wherein the chamber has opposite side walls and wherein the second means extends through most of the distance between the opposite side walls of the chamber. the chamber includes a ceiling and wherein the second means supports the ceiling. the radiation source extends through the second means, the chamber includes a ceiling and wherein the second means supports the ceiling. the second means includes an intermediate wall made from a radiation shielding material and wherein the intermediate wall is separated by air gaps from the walls defining the chamber and wherein one of the walls defining the chamber is on the opposite side of the chamber from the radiation source and wherein a beam stop is disposed in the one of the walls defining the chamber. a chamber defined by walls made from a radiation shielding material, a radiation source constructed to provide radiation in the chamber, a conveyor system constructed to carry the articles in a loop through the chamber for the reception of the radiation in the chamber by the articles, ozone being derived in the chamber from the radiation source, and an intermediate wall disposed within the loop in the chamber and separated by air gaps from the walls defining the chamber and made from a radiation-shielding material and provided with dimensions relative to the walls defining the chamber for restricting the flow through the chamber of the ozone derived from the radiation source. means disposed in the chamber for removing the ozone from the chamber. the radiation source extends through the intermediate wall. the walls of the chamber are made from a radiation shielding material and wherein means are disposed in the chamber for removing ozone from the chamber and wherein the chamber has opposite side walls and wherein the intermediate wall extends in a direction transverse to the opposite side walls of the chamber. means are disposed in the chamber for receiving radiation from the source and for converting the radiation to photons in the chamber and wherein the intermediate wall inhibits the photons from impinging on the walls defining the chamber, thereby providing for a reduction in the thickness of the walls defining the chamber. the intermediate wall is separated from the walls defining the chamber and wherein one of the walls defining the chamber is on the opposite side of the chamber from the radiation source and wherein a beam stop is disposed in the one of the walls. the chamber includes a ceiling and wherein the flow-restricting means including the intermediate wall provides a support for the ceiling. the radiation source extends through the means for restricting the flow of the ozone through the chamber and wherein means are disposed in the chamber for removing ozone from the chamber and wherein the flow-restricting means constitutes a first means and wherein second means are disposed in the chamber for receiving radiation from the source and for converting the radiation to photons in the chamber and wherein the first means including the intermediate wall inhibits the photons from impinging on the walls defining the chamber, thereby providing for a reduction in the thickness of the walls defining the chamber and wherein the chamber includes a ceiling and wherein the flow-restricting means provides a support for the ceiling. a chamber defined by walls, a radiation source constructed to provide radiation in the chamber, a conveyor system constructed to carry the articles through the chamber for the reception by the articles of radiation in the chamber, a beam stop disposed in the chamber for absorbing electrons from the radiation source and for converting energy from the absorbed electrons into photons and for emitting the photons, and the beam stop being disposed relative to a particular one of the walls of the chamber to provide for a reduction in the intensity of the photons in the chamber by the particular one of the walls, and means disposed within the loop in the chamber and separated by air gaps from the walls defining the chamber for inhibiting the photons from impinging on the walls defining the chamber, thereby providing for a reduction in the thickness of the walls defining the chamber. a chamber defined by walls made from a radiation shielding material, a radiation source constructed to provide radiation in the chamber, a conveyor system constructed to carry the articles in a loop through the chamber for the reception by the articles of radiation in the chamber, a beam stop disposed in the chamber for absorbing electrons from the radiation source and for converting energy of the absorbed electrons into photons and for emitting the photons, the beam stop being disposed relative to a particular one of the walls of the chamber to provide for a reduction in the intensity of the photons in the chamber by the particular one of the walls, means disposed within the loop in the chamber and separated by air gaps from the walls defining the chamber for inhibiting the photons from impinging on the walls defining the chamber, thereby providing for a reduction in the thickness of the walls defining the chamber, ozone being derived in the chamber from the radiation source, and the photon-inhibiting means being operative to restrict the flow of ozone through the chamber. the photon-inhibiting means includes an intermediate wall disposed in the chamber and separated by the air gaps from the walls defining the chamber. the intermediate wall is made from a radiation shielding material and wherein the radiation source extends through the intermediate wall and wherein one of the walls defining the chamber faces the radiation source and the intermediate wall and wherein the beam stop is disposed in the one of the walls defining the chamber. a chamber defined by walls, a radiation source disposed to provide radiation, a loading area for the articles, an unloading area for the articles, a conveyor system constructed to move the articles in a loop within the chamber, a first path extending from the loading area to the loop within the chamber, a second path extending from the loop within the chamber to the unloading area, the first and second paths being disposed in adjacent relationship to each other and in communicating relationship with the chamber and being separated from the chamber for at least a portion of their lengths by a particular one of the walls defining the chamber, an intermediate wall disposed within the loop in the chamber and made from a radiation-shielding material, and an additional wall disposed outside of the chamber, the first and second paths being confined between the particular wall and the additional wall. the walls defining the chamber and the additional wall are made from a radiation shielding material and wherein the intermediate wall is separated in the chamber from the walls defining the chamber. the walls defining the chamber and the intermediate wall are made from a radiation shielding material and wherein the particular wall and the additional wall are disposed relative to the loading area and the unloading area to prevent radiation from the source from reaching the loading area and the unloading area and wherein the radiation source extends through the intermediate wall and wherein the intermediate wall is spaced by air gaps from the walls defining the chamber. the particular wall has a limited length to provide for a communication between the chamber and each of the first and second paths and wherein one of the walls defining the chamber is on the opposite side of the chamber from the radiation source and wherein a beam stop is disposed in the one of the walls defining the chamber. means disposed in the chamber for receiving radiation from the source and for converting the radiation to photons movable into the chamber, and means including the intermediate wall disposed within the loop in the chamber for inhibiting the photons from impinging on the walls defining the chamber, thereby providing for a reduction in the thickness of the walls defining the chamber. ozone being derived in the chamber from the radiation source, and means including the intermediate wall disposed in the chamber for restricting the flow of ozone through the chamber, the ozone-restricting means including the intermediate wall being disposed within the loop in the chamber in the spaced relationship to the walls defining the chamber and being made from a radiation shielding material. the particular wall and the additional wall are disposed relative to the loading area and the unloading area to prevent radiation from the source from reaching the loading area and the unloading area and wherein the particular wall has a limited length to provide for a communication between the chamber and each of the first and second paths and wherein ozone is derived in the chamber from the radiation source and wherein means are disposed in the chamber for restricting the flow of ozone through the chamber and wherein the ozone-restricting means includes the intermediate wall disposed in the chamber in the spaced relationship to the walls defining the chamber and made from the radiation shielding material and wherein one of the walls defining the chamber is disposed opposite in the chamber from the radiation source and the intermediate walls and wherein a beam stop is disposed in the one of the walls on the opposite side of the chamber from the radiation source. providing a chamber defined by a plurality of walls, providing a loading area for the articles at a position displaced from the chamber, providing an unloading area for the articles at a position displaced from the chamber and from the loading area, providing a source of radiation in the chamber, the source having properties of producing photons in the chamber, providing a conveyor path for a movement of the articles in a loop within the chamber from the loading area to the unloading area and for the irradiation of the articles by the source during the movement of the articles in the loop within the chamber, and providing a member within the loop in the chamber for inhibiting the movement of the photons to the walls defining the chamber, thereby minimizing the thickness of the walls defining the chamber, the member being spaced by air gaps from the walls defining the chamber. the member is an intermediate wall disposed within the loop in the chamber in the spaced relationship to the walls defining the chamber and wherein the chamber has opposite sides and wherein the intermediate wall extends through most of the distance between the opposite sides of the chamber. the walls in the plurality and the intermediate wall are formed from a radiation shielding material. a first path extends from the loading area to the loop within the chamber and wherein a second path extends from the unloading area to the loop within the chamber in adjacent relationship to the first path and wherein an additional wall is disposed outside of the chamber in a cooperative relationship with a particular one of the walls defining the chamber to define a confining relationship for the first and second paths. the particular one of the walls constitutes a first particular one of the walls and wherein the walls defining the chamber and the member and the additional wall are made from a radiation shielding material and wherein a second particular one of the walls defining the chamber is opposite in the chamber from the radiation source and wherein a beam stop is disposed in the one of the walls defining the chamber. providing a chamber defined by a plurality of walls, providing a conveyor path for a movement of the articles in a loop within the chamber and for an irradiation of the articles by a radiation source during the movement of the articles in the loop within the chamber, providing a loading area for the articles at a position displaced from the chamber, providing an unloading area for the articles at a position displaced from the chamber and the loading area, the conveyor path including the loading area and the unloading area, providing the source of radiation in the chamber, the source having properties of deriving ozone in the chamber, and providing a member within the loop in the chamber for restricting the flow of the ozone in the chamber. the member is an intermediate wall disposed within the loop in the chamber and separated by air gaps from the walls defining the chamber and wherein the radiation source extends in the chamber through the intermediate wall. the intermediate wall an d the walls defining the chamber are made from a radiation shielding material and wherein one of the walls defining the chamber is on the opposite side of the chamber from the radiation source and the intermediate wall and wherein a beam stop is disposed in the one of the walls defining the chamber . providing a chamber defined by a plurality of walls, providing a conveyor path for the movement of the articles in a loop within the chamber and for the irradiation of the articles by a radiation source in the chamber during the movement of the articles in the loop within the chamber, providing a loading area for the articles at a position displaced form the chamber, providing an unloading area for the articles at a position displaced from the chamber and the loading area, providing a first path from the loading area to the chamber, providing a second path from the chamber to the unloading area in adjacent relationship to the first path, the first and second paths being included in the conveyor path and being disposed in adjacent relationship to a particular one of the walls defining the chamber, disposing within the loop in the chamber an intermediate wall made from a radiation shielding material and separated by air gaps from the walls defining the chamber, and providing an additional wall on an opposite side of the first and second paths from the particular wall. the walls defining the chamber and the additional wall and the intermediate wall are made from a radiation shielding material, the first and second paths are substantially parallel and are contiguous and wherein the particular wall and the additional wall are substantially parallel to each other and to the first and second paths and are respectively contiguous to the first and second paths on opposite sides of the first and second paths and wherein one of the walls defining the chamber is on the opposite side of the chamber from the radiation source and wherein the radiation source extends through the intermediate wall and wherein a beam stop is recessed in the one of the walls defining the chamber. 2. A system according to claim 1, wherein the intermediate wall has an aperture through which the radiation source is disposed on the particular axis. 3. A system according to claim 1, wherein the chamber-defining radiation shielding material includes a ceiling section that is supported in part by the intermediate wall and wherein the substantially closed loop defines the path of movement of the article through the chamber. 4. A system according to claim 1, wherein a second portion of the conveyor system is positioned for transporting articles in a that is outside of the substantially closed loop but continuous with the substantially closed loop; 5. A system according to claim 1, wherein the radiation source is an electron beam source, the system further comprising 6. A system according to claim 5, wherein the intermediate wall is positioned relative to the radiation shielding material defining the chamber, and is provided with dimensions in the transverse direction relative to the radiation shielding material defining the chamber, for restricting flow through the chamber of ozone derived in the target region from the radiation source and wherein the substantially closed loop defines the path of movement of the articles and wherein the chamber defined by the radiation shielding material has opposite side walls transverse to the wall adjacent the target region and transverse to the other wall and wherein the intermediate wall extends most of the distance between the opposite side walls of the chamber to prevent the photons from impinging upon the other wall of the chamber and from impinging upon substantial portions of the side walls closest to the other wall. 7. An irradiation system as set forth in claim 6 8. An irradiation system for irradiating articles, including: 9. An irradiation system as set forth in claim 8 wherein 10. An irradiation system as set forth in claim 8 wherein 11. An irradiation system as set forth in claim 8 wherein 12. An irradiation system as set forth in claim 9 wherein 13. An irradiation system as set forth in claim 8 wherein 14. An irradiation system for irradiating articles, including, 15. An irradiation system as set forth in claim 14, including, 16. An irradiation system as set forth in claim 14 wherein 17. An irradiation system as set forth in claim 14 wherein 18. An irradiation system as set forth in claim 14 wherein 19. An irradiation system as set forth in claim 14 wherein 20. An irradiation system as set forth in claim 14 wherein 21. An irradiation system as set forth in claim 14 wherein 22. An irradiation system for irradiating articles, including, 23. An irradiation system for irradiating articles, including, 24. An irradiation system as set forth in claim 23 wherein 25. An irradiation system as set forth in claim 24 wherein 26. An irradiation system for irradiating articles, including, 27. An irradiation system as set forth in claim 26 wherein 28. An irradiation system as set forth in claim 27 wherein 29. An irradiation system as set forth in claim 26 wherein 30. An irradiation system as set forth in claim 26, including, 31. An irradiation system as set forth in claim 26, including, 32. An irradiation system as set forth in claim 30 wherein 33. A method of providing an irradiation of articles, including the steps of: 34. A method as set forth in claim 33 wherein 35. A method as set forth in claim 34 wherein 36. A method as set forth in claim 34 wherein 37. A method as set forth in claim 36 wherein 38. A method of providing an irradiation of articles, including the steps of: 39. A method as set forth in claim 38 wherein 40. A method as set forth in claim 39 wherein 41. A method of providing an irradiation of articles, including the steps of: 42. A method as set forth in claim 41 wherein
	abstract	A nuclear reactor module includes a reactor vessel containing coolant, a reactor core submerged in the coolant, and a heat exchanger configured to remove heat from the coolant. The nuclear reactor module further includes one or more heaters configured to add heat to the coolant during a startup operation and prior to the reactor core going critical.
	description	The present invention relates to a method for suppressing corrosion of pipes, devices, machineries, and the like that constitute a plant and also relates to a plant. A thermal power plant and a nuclear power plant generally provided with a system which drives a turbine with steam generated by a steam generator and returns condensate water to the steam generator. Since, pipes and/or devices of such system may be damaged due to corrosion during operation, countermeasures to such damage have been taken for reducing the corrosion. For example, in secondary systems of current pressurized water nuclear power plants, such countermeasures as makeup water management and management of water treatment chemicals are being taken for preventing infiltration of impurities into the system in order to prevent corrosion troubles in steam generators and turbines. In order to suppress corrosion of devices and pipes that constitute a system, countermeasures are taken to obtain a deoxidized and reducing atmosphere by pH control with use of pH adjusters and injection of hydrazine. Furthermore, various other countermeasures or procedures have been taken, such as installation of a desalting device and the proper operation thereof for removing infiltrated impurities out of the system, installation of a clean-up system and a steam generator blowdown collection system, and installation of a deaerator for reducing dissolved oxygen. The deaerator is placed to deaerate circulating water of the system and to reduce oxygen from transferring to the steam generator. The deaerator acts to suppress increase in corrosion potential of structural members due to oxygen contribution. As oxygen concentration increases, cracking such as intergranular corrosion cracking and stress corrosion cracking occurs due to the potential increase. Meanwhile, elution of metal ions from pipes and the like is a typical phenomenon that occurs in high temperature hot water. Elution of metal ions causes operational problems attributed to corrosion of structural members as well as pipes and other members, and exerts various influences such as increase in frequency of maintenance. Moreover, eluted metal ions are deposited and crystallized as an oxide at high temperature portions in the system, such as pipe surfaces and the steam generator, which causes a phenomenon of corrosion cracking due to potential increase. Since the adhering oxide causes deterioration in heat transfer, the oxide needs to be removed on a periodic basis by chemical cleaning. Thus, such phenomena as metal elution and corrosion may gradually be accumulated during a long-term plant operation or running and may possibly cause disaster at some point without notice. In order to obviate such phenomena, chemicals such as ammonia and hydrazine are injected for pH control to implement deaeration so as to reduce iron elution from the system as a countermeasure to prevent inflow of iron into the steam generator. In order to eliminate alkali concentration in a clevis portion, various suggestions have been made for water quality control, such as chloride ion concentration management and dissolved oxygen concentration control. Patent Document 1: Japanese Patent Laid-Open Publication No. 2010-96534 Patent Document 2: Japanese Patent No. 3492144 As described in the forgoing, conventional corrosion suppressing methods not only need various devices such as a deaerator and chemical injection and control devices for suppression of corrosion, but also require execution of chemical concentration control and strict water chemistry control. Consequently, equipment is enlarged and operation control is complicated, which causes increase in equipment costs and operating costs of the plants. The present invention has been made in consideration of the circumstances mentioned above, and an object thereof is to provide a method for suppressing corrosion in a plant and a plant in which a structural member of a system having a steam generator and a turbine is deposited with a protective substance so as to achieve reduction in equipment costs and running costs. In order to solve the problem in the conventional art mentioned above, the present invention provides a method for suppressing corrosion in a plant including a system which is provided with a steam generator, a turbine, a condenser and a heater and in which non-deaerated water circulates, wherein depositing a structural member of the system which comes into contact with the non-deaerated water with a protective substance. In the method for suppressing corrosion in a plant and the plant according to the present invention, equipment costs and running or operating costs of the plant can be reduced. Hereunder, an embodiment of the present invention will be described with reference to the accompanying drawings. (Constitution) An example in which a method for suppressing corrosion of the present embodiment which is applied to a secondary system of a pressurized water nuclear power plant will be explained with reference to FIGS. 1 through 7. As shown in FIG. 1, the secondary system includes a nuclear reactor 1, a steam generator 2, a high pressure turbine 3, moisture content separation heater 4, a low pressure turbine 5, a condenser 6, a low pressure heater 7, a high pressure heater 8, a high temperature desalting device (purification equipment) 9, and a high temperature filter (purification equipment) 10. The condenser 6 may include a condenser unit having a low-temperature purification device (desalting device+filter) provided in the downstream side of the condenser 6. Since the secondary system of the structure mentioned above, is not provided with a deaerator provided in the secondary system of the conventional pressurized water nuclear power plant, non-deaerated water circulates inside the secondary system. The non-deaerated water is the circulating water which is neither subjected to deaeration processing by a deaerator nor subjected to injection of chemicals such as hydrazine for deaeration by a chemical injection device. In the present embodiment, surfaces of pipes and devices that constitute the system, such as the steam generator 2, the low pressure heater 7 and the high pressure heater 8, i.e., surfaces of structural members which come into contact with non-deaerated water, are deposited with a protective substance by a conventionally known method. The structural member may be made of one or more of a steel material, a non-steel material, a nonferrous metal, or a weld metal corresponding to types or location of devices, machineries or like. Examples of the protective substance include an oxide, a hydroxide, a carbonate compound, an acetic acid compound, and an oxalic acid compound of a metallic element selected out of Ti, Y, La, Zr, Fe, Ni, Pd, U, W, Cr, Zn, Co, Mn, Cu, Ag, Al, Mg, and Pb. Further, although one type of the protective substance may be formed on the pipes and various devices, the protective substance may be formed in combination of two or more types. For example, in the present embodiment, as shown in FIG. 2, the surface of the steam generator 17 is deposited with a titanium based protective substance (such as titanium oxide (TiO2)) 18, the surface of a pipe 13 is deposited with a yttrium based protective substance 14 (such as yttria (Y2O3)), and the surface of the heater 15 is deposited with a lanthanum based protective substance 16 (such as lanthana (La2O3)). FIG. 2 is a concept view showing a protective substance 12 that is deposited on the surface of the structural member 11. As a method for depositing with the protective substance 12, various publicly known methods may be used, such as depositing by spray and application, and depositing by bringing a fluid containing a protective substance into contact with the pipes and the devices. Further, such depositing is suitably performed before a plant operation or at the time of periodical inspections depending on a degradation level of the deposit. (Operation and Function) As described in the foregoing, a deaerator disposed in a conventional secondary system is placed to deaerate circulating water in the system for the purpose of reducing transfer of oxygen to a steam generator. The deaerator performs a function of suppressing increase in corrosion potential in structural members by oxygen contribution. Accordingly, if devices or equipment such as the steam generator including pipes would not be damaged by corrosion without deaeration processing applied to the circulating water in system water, it is not necessary to locate the deaerator itself, making it possible to achieve downsizing of equipment and reduction in equipment costs and running or operating costs. Inventors of the present invention focused attention on this point and employed the above described constitution. As a result, it was newly found out that the deaerator in the secondary system which was conventionally needed could be saved. More specifically, in the present embodiment, a protective substance that deposits the pipes and the devices of the secondary system serves as a barrier against oxygen diffusion in the water of the system, thereby reducing the amount of oxygen reaching the surface of the structural member. This reduction eliminates increase in corrosion potential by the oxygen contribution and makes it possible to keep the surface of the structural member at low voltage. As a result, it becomes possible to use non-deaerated water as circulating water of the system. Hereinafter, effect confirmation tests performed on the protective substance of the present embodiment will be explained with reference to FIGS. 3 through 7. (Effect Confirmation Test 1) FIG. 3 is a view showing a corrosion amount ratio of structural members 11 of the present embodiment deposited with the protective substances 12 with respect to a structural member (pure material) not deposited with the protective substances. As a result of a test conducted in neutral non-deaerated water of 180° C., considerable reduction in the corrosion amount was confirmed in all the structural members 11 deposited with respective protective substances 12 (TiO2, Y2O3 and La2O3 in this example) as shown in FIG. 3. (Effect Confirmation Test 2) FIG. 4 is a view showing a corrosion amount ratio between a pure material and structural members 11 deposited with protective substances 12 of the present embodiment in the case of using high temperature hot water different in water quality (neutral, acid and alkaline). FIG. 4 indicates that corrosion due to oxidation progressed in the pure material, whereas the structural members 11 deposited with the protective substances 12 of the present embodiment provided a corrosion suppressing effect regardless of water quality. (Effect Confirmation Test 3) FIG. 5 is a view showing a corrosion amount ratio between a pure material and the structural members of the present embodiment in the case of varying temperatures of the system water. FIG. 5 indicates that corrosion due to oxidation progressed in the general pure material, whereas the structural members deposited with the protective substances of the present embodiment provided a corrosion suppressing effect by the suppression of the oxygen diffusion. Furthermore, in a low temperature region, since the corrosion did not occur, a corrosion weight ratio to the pure material showed almost no change, whereas as the temperature increases, the oxidation reaction progressed and the corrosion amount increased. This fact indicates that a diffusion barrier function of the protective substances became stronger. Thus, even under water quality conditions with the deaerator being saved, the corrosion suppressing effect by the protective substances becomes notable with a higher temperature. This effect is exhibited in the respective substances. Therefore, it is found that the protective substances of the present embodiment exhibit a remarkable corrosion suppressing effect at operating temperature of the plant. (Effect Confirmation Test 4) FIG. 6 is a view showing an adhering amount ratio between a pure material and the structural members deposited with the protective substances of the present embodiment in the case where system water contains particulate clads or ions. Generally, in adhesion of clads, zeta potential in clad particles contributes to the adhesion. General metal oxide takes a positive value in an acid region, reaches an isoelectric point (0) around a neutral region, and takes a negative value in an alkaline region. The Confirmation Test 4 was conducted under alkaline water conditions, and therefore, the clad provided a negative potential. The protective substances also had negative potential in the alkaline region. As a result, the protective substances had electrostatic repulsion with the clad. Since the corrosion potential on the surfaces of the structural members acted as an oxygen diffusion barrier because of the protective substances depositing the surfaces, a corrosion potential stabilizing action was also implemented. As shown in FIG. 6, adhesion or crystallization of ions was notably influenced by the oxygen concentration on the surface of the members. That is, the oxygen concentration contributes to both the crystallization by the reaction between the ion and the oxygen, and by the variation in corrosion potential. The adhesion or crystallization of the ions is reduced by such an effect of suppressing oxygen from transferring to the surface of the structural member. It is also known that roughness on the surface of the structural member affects the clad adhesion. Further, since the depositing of the protective substances fills the processing traces on the surface of the structural member, and hence, the surface becomes smooth. As a result, the adhesion of clads can be suppressed. (Effect Confirmation Test 5) FIG. 7 is a view showing a corrosion amount ratio between a pure material and the structural members 11 deposited with the protective substances 12 of the present embodiment in the case of using deaerated water and non-deaerated water at a temperature of about 185° C. as the system water. As shown in FIG. 7, the structural members 11 deposited with the protective substances 12 of the present embodiment do not attain a strong corrosion suppressing function in the case of using the deaerated water with a low dissolved oxygen concentration. On the other hand, it is indicated that the structural members 11 deposited with the protective substances 12 provide a remarkable corrosion suppressing effect in the case of non-deaerated water with a high dissolved oxygen concentration. (Effect) As can be understood from the above effect confirmation tests 1 to 5, the effect confirmation tests indicate that the protective substances of the present embodiment provide a remarkable corrosion suppressing effect in the system using non-deaerated water at a plant operation temperature. It is also indicated that the protective substances of the present embodiment provided a remarkable corrosion suppressing effect regardless of the water quality of the system water and regardless of the clads and ions contained in the system water. Accordingly, as mentioned above, by forming a depositing of the protective substance according to the present embodiment on the surfaces of structural members of pipes and system devices, non-deaerated water can be used as system water. As a result, it becomes possible to save a deaerator and a chemical injection device or like. The method for suppressing corrosion and the plant according to the present embodiment can achieve downsizing of the plant and reduction in equipment costs and can also eliminate the necessity of deaerator control, dissolved oxygen control in operation, and various chemical concentration control, so that the substantial reduction in running costs or operating costs can also be achieved. It is to be noted that although examples of using TiO2, Y2O3, and La2O3 as a protective substance have been explained in the present embodiment, the same operational effects can be obtained by using metallic elements other than those described hereinbefore. The same operational effects can also be obtained by using a hydroxide, a carbonate compound, an acetic acid compound, or an oxalic acid compound of the above metallic elements as a protective substance. Furthermore, it is to be noted that although an example of applying the invention to a secondary system of a pressurized water nuclear power plant has been explained in the present embodiment, the present invention is not limited thereto, and is applicable to secondary systems of other plants such as fast reactors and to primary systems of thermal power generation plants. 1 - - - nuclear reactor, 2 - - - steam generator, 3 - - - high pressure turbine, 4 - - - moisture content separation heater, 5 - - - low pressure turbine, 6 - - - condenser, 7 - - - low pressure heater, 8 - - - high pressure heater, 9 - - - high temperature desalting device, 10 - - - high temperature filter, 11 - - - structural member, 12 - - - protective deposit.
	description	The present application claims priority from Japanese application JP 2005-307584 filed on Oct. 21, 2005, the content of which is hereby incorporated by reference into this application. 1. Field of the Invention The present invention relates to a semiconductor inspection equipment for inspecting semiconductors or analyzing defects of semiconductors, and in particular, it relates to a semiconductor inspection equipment having a navigation function with respect to defective positions. 2. Background Art When a defective position of a semiconductor is logically determined, it is necessary to physically clarify the defective position for performing an observation through probing, processing (FIB: Focused Ion Beam), or a microscope. Generally, a process of clarifying such defective position is automatically carried out through CAD navigation utilizing CAD data indicating circuit design used when a semiconductor is manufactured. Patent Document 1: JP Patent Publication (Kokai) No. 2000-164659 A In recent years, with an increase in capacity and miniaturization of semiconductors and with an increase in accuracy, layer, and complexity of circuits, accuracy on the order of nm is being demanded for determining defective positions. CAD data indicating such minute information is increasing to approximately 100 G bytes. Searching enormous quantities of CAD data for a defective position, and loading/displaying CAD data are beginning to require a considerable amount of time. Meanwhile, attempts have been made to solve the above problems by deleting data unnecessary for analysis beforehand. However, this method is also beginning to require time since it is necessary to determine how necessary data and unnecessary data should be separated, and initial data is enormous. Basically, in CAD data indicating circuit design, even a single wire is accurately displayed for manufacturing a semiconductor, resulting in a large amount of data as described above. However, as data indicating defective positions of semiconductors, such CAD data does not have necessary information but has too much unnecessary information. Further, if probing of a semiconductor inspection equipment is taken for example, when a plurality of possibly defective positions are found and one of such defective positions is determined through probing (a test for measuring electrical characteristics by bringing an electrode probe into contact with an object to be measured) with a semiconductor inspection equipment, it is desirable to put only the thus determined defective position into CAD data. Particularly, when electrical characteristics are measured by bringing a probe into contact with a plug on the order of several dozen nm, it is very stressful to recognize a target plug among similar plug patterns with an electron microscope image alone. As a problem when new CAD data is created, failure to easily display the repetition number of important positions, mats and plugs, results in display of incorrect positions. Further, there are cases in which CAD data for circuit design cannot be obtained from a viewpoint of security, and counting the repetition number of mats and plugs by visual observation can be easily the cause of false recognition. It is an object of the present invention to provide a navigation system by which defective positions can be easily determined when a semiconductor device is inspected with an electron microscope image. In the present invention, in semiconductor inspection, logical information for the purpose of indicating defective positions is created in a CAD format, instead of CAD data of physical information indicating circuit design, when CAD navigation to defective positions is performed. For example, by attaching marks such as rectangles, characters, or lines, on an electron microscope image with software, quick navigation is performed with required minimum information. Preferably, by using created CAD data, re-navigation with the same equipment and CAD navigation to a heterogeneous equipment are performed. Namely, as an example, a semiconductor inspection equipment of the present invention comprises a movable stage while holding a semiconductor sample, a sample image acquisition unit that irradiates the semiconductor sample held on the stage with a charged particle beam, so as to obtain a sample image through a sample signal discharged from the sample due to irradiation of the charged particle beam, a memory unit for storing CAD data concerning the semiconductor sample, and a display unit capable of displaying CAD data such that the CAD data is superimposed on the sample image obtained by the sample image acquisition unit. Further, the semiconductor inspection equipment of the present invention has a grid creation means that generates a grid composed of vertical and horizontal lines having a constant distance, so as to superimpose the grid on the sample image displayed on the display unit, and a means of generating systematic numbers indicating vertical and horizontal positions of a plurality of rectangular areas contained in the grid created by the grid creation means, so as to display the numbers in relevant areas in the grid displayed on the display unit. Preferably, the grid creation means has a function of generating another grid having a narrower distance between the lines thereof in one rectangular area of the previously created grid. The grid generated by the grid creation means and then superimposed on the sample image displayed on the display unit is stored as CAD data. Further, the semiconductor inspection equipment of the present invention preferably has a means of generating characters, symbols, lines, figures, and/or painted figures, so as to superimpose them on the sample image displayed on the display unit. In the present invention, additional information of CAD data for indicating defective positions used for semiconductor inspection/analysis is newly created on a microscopic observation image of a semiconductor inspection equipment with simple operation, without using CAD data of circuit design used with a semiconductor manufacturing equipment as in a conventional technology. While a CAD data format according to the present invention is the same as that for circuit design, the purpose is to indicate defective positions of a semiconductor. By using a CAD data format as a data format, it becomes easier to find a defective position repeatedly with the data, and also such format can be used as a means of cooperating with another type of semiconductor inspection/analysis equipment. For example, in the case of probing using the semiconductor inspection/analysis equipment, the type of plug can be identified by its initial letter or color; a tick indicating completion of each plug is displayed in accordance with the status of inspection/analysis; wires, cut surfaces, and target areas are displayed by using solid/dashed lines or changing the thickness of the line; a display method by which layers are divided into hierarchies is used; and same-sized markers are displayed irrespective of display magnification. These can be registered in CAD data as additional information indicating defective positions with simple operation of clicking and dragging individual operation buttons. Further, in order to create additional information with simple screen operation, if functions such as copying, cutting, x-axis inversion, and y-axis inversion are provided, usability can be improved. In accordance with the present invention, quick and accurate positioning can be achieved, whereby usability for equipment users is improved. Embodiments of the present invention will be hereafter described with reference to the drawings. FIG. 1 shows a constitutional example of an embodiment of a semiconductor inspection equipment 1 of the present invention. In FIG. 1, the semiconductor inspection equipment 1 comprises in a sample chamber 7 a probe stage 6, which includes a stage including a sample holder 2 for holding a sample and a sample holder receiver 17 for holding the sample holder 2, and a probe unit 33. While a sample is fixed on the sample holder 2, since it is a thin piece, it is not shown for convenience of drawing figures. In order to inspect the sample, an electro-optical device 4 (charged particle device) provided with an ion pump 44, such as SEM (Scanning Electron Microscope) or FIB, is provided opposite to the sample holder 2 on the housing of the sample chamber 7. A sample image formed based on a sample signal, such as secondary electrons, discharged from a sample by an electron beam or an ion beam emitted from the electron optical device 4 is displayed on a display 15 of a display unit 14. The display unit 14 comprises a display body 16 having a function of controlling a SEM image, a CCD image from a CCD (Charge Coupled Device) camera, and data from a CAD navigation body and a function of controlling a stage, probe, exhaust system, and optical system through a GUI (Graphical User Interface) screen, in addition to a display function. Further, the display body 16 includes a CPU (Central Processing Unit), a memory, and a hard disk. Further, a probe coarse adjustment image acquisition device 10 is provided in the vicinity of the electro-optical device 4. From the electro-optical device 4, a charged particle beam (electron or ion beam) for observing a sample surface or movement of probes 3 is emitted in the direction of the sample holder 2. The probe coarse adjustment image acquisition device 10 arranged in parallel and in the vicinity of the electro-optical device 4 on the upper surface portion of the housing of the sample chamber 7 comprises a probe coarse adjustment optical microscope and a CCD camera for acquiring an image. With this device 10, a state of coarse adjustment with respect to the sample via the probes 3 is observed, and it can be acquired as image information. With regard to the probe coarse adjustment image acquisition device 10, not only a longitudinal direction type 10A but also a lateral direction arrangement type 10B for a cross-shape arrangement is used. With this cross-shaped arrangement, the probes 3 are observed longitudinally and laterally, whereby the state of coarse adjustment is reliably grasped. The magnification of an image through coarse adjustment in the lateral direction is made greater than that of an image through coarse adjustment from above. For measurement, first, a coarse adjustment for bringing the probes 3 closer to the horizontal direction is performed with the use of the above probe coarse adjustment image acquisition device 10A, and at this point, it is necessary to capture the plurality of probes 3 on an image through coarse adjustment. With regard to coarse adjustment in the lateral direction, the probes 3 are brought down so that they become in proximity to the sample while examining an image through coarse adjustment in the lateral direction. Next, an operation of bringing the probes 3 into contact with the sample is performed while examining adjustment of the focus of the tips of the probes 3 and the sample, using the electro-optical device 4. If the distance between the probes 3 and the sample is small when the coarse adjustment in the lateral direction is performed, time necessary for the operation of bringing the probes 3 close to the sample using the electro-optical device 4 can be shortened. For this reason, the magnification of an image through coarse adjustment in the lateral direction is made greater than that of an image through coarse adjustment from above. The stage comprises the sample holder 2 for holding a sample, a sample stage 50 on which the sample holder 2 is mounted, a large stage 49 on which the sample stage 50 is mounted, and a base 48 on which the large stage 49 is moved. At a place immediately below the probe coarse adjustment image acquisition device 10 where the positioning of coarse adjustment is performed, the positional relationship between the probes and the positional relationship between each probe and a sample are measured, and each probe or a set of probes is positioned with respect to the sample, such that the distance between the probes becomes within the range of vision of the electro-optical device 4 or within the region seen in the thickness direction of a sample that is under the possible measurement range. In addition, the distance between each probe and the sample or the distance between the set of probes and the sample is made less than a predetermined value and more than zero seen in the direction perpendicular to the sample thickness direction, such that each probe and the set of probes become slightly separated from the sample seen in the direction perpendicular to the sample thickness direction (the distance is made as short as possible while a minimum distance or gap between the probes and the sample is maintained). The probe stage 6 comprises a probe unit 33 provided with a probe holder 31 for holding the probes 3, a probe unit base 34 for holding the probe unit 33, and a probe unit support 35 for connecting the probe unit base 34 to the large stage 49. The probe unit 33 comprises x, y, and z tables (not shown), and it can move the probes 3 three-dimensionally. The base 48 is fixed on a side wall of the sample chamber 7 via a fixed member 47. The sample chamber 7 is provided with a sample exchange chamber 8 and a probe exchange chamber 9. The sample chamber 7 is provided with a field-through on a face plate 71, in order to send a signal for controlling operation of the x, y, and z tables of the probe unit 33 and a signal for controlling operation of x, y, and z tables of the sample stage 50 from the outside of the sample camber 7. FIG. 2 shows a diagram of a basic concept concerning the operation of software of the present invention. A plurality of workstations and personal computers may constitute physical components of the CAD display and the semiconductor inspection equipment shown in FIG. 2. The present invention includes any physical components for creating CAD data as a semiconductor inspection/analysis system as a whole. A SEM image of a semiconductor is displayed on the display 14 of the semiconductor inspection equipment. For the purpose of inspecting the semiconductor, a mark representing position, a mark representing electrical characteristics, and a mark representing completion of measurement are newly created and displayed, so that the marks are superimposed on the SEM image, using a mouse and keyboard. (1) to (5) of FIG. 2 show the flow of data. In (1) of FIG. 2, marked information is transferred to the CAD navigation body 105. In (2) of FIG. 2, CAD data is created in a general CAD data format, for example, in the GDSII stream format, in the CAD navigation body, and it is stored in a file or a database. Namely, mark information is converted into coordinates and wiring information, so as to create data in accordance with a CAD data format for manufacturing semiconductors. While such CAD data format is originally for data for creating a semiconductor, the marks created in the present case are for inspecting a semiconductor, thereby differing in the purpose. Thus, grids, figures, and markers are created in accordance with a format for creating CAD data. For example, a CAD data format for storing grid data is created with array information indicating repetition. A CAD data format for storing characters, rectangles, and painted figures can be used without change. Circles, ovals, and painted circles are available depending on a CAD data format. If a format for circles is not available, polygons with multiple points are created and stored instead. Circle-attached characters can be composed of a combination of circles and characters. Since markers are not available in any CAD data format, an advanced function format (user-specific unique format) of a CAD is used. When a user-specific format is used, generally, there is a disadvantage that it cannot be used for CAD navigation between heterogeneous types. A SEM image is not contained in stored data. Only figures drawn are stored in accordance with a CAD data format for creating semiconductors. Coordinate positions are calculated based on the position of a SEM image and stored in CAD data. By unifying with the format used when creating semiconductors, an overlay with a CAD display used when creating semiconductors is enabled and only one kind of conversion format becomes necessary. For such reasons, the present format is used. After (3) of FIG. 2, the flow of data is the same as that of a conventional CAD navigation. In (3) of FIG. 2, CAD data is obtained, coordinate conversion is performed, and conversion into a format for an overlay on a SEM image is performed. In (4) of FIG. 2, converted data is transferred to the display of the semiconductor inspection equipment. (5) of FIG. 2 indicates cooperation of CAD data with the FIB, which is another type of semiconductor inspection/analysis equipment. As a specific example, in the case of analysis on an SRAM, even when a defect is determined in one bit, since one bit is composed of six transistors, it is impossible to determine which transistor is defective. By examining electrical characteristics (such as VdId characteristics, VgId characteristics, or resistors) of transistors, using semiconductor inspection equipment, the defective transistor in question can be determined. In order to clarify the cause based on such transistor characteristics, it is often the case that a defective position is cut with the FIB. It is efficient if such defective position and cut surface are expressed by means of CAD data in the semiconductor inspection equipment. CAD data created with the semiconductor inspection equipment can be used for mere SEM observation or for a semiconductor inspection/analysis equipment such as the FIB. FIG. 3 shows an example of a display screen of the semiconductor inspection equipment of the present invention. The rectangle composed of 401 to 235 on the left side of the screen representing CAD data is overlay-displayed on an electron microscope image. On the right side of the screen, operation buttons 201 to 228 and operation property information 229 to 234 are displayed, so as to create CAD data with a simple mouse operation. The operation buttons and property information will be hereafter described. Area Selection Button 201 After clicking this button, a cursor is moved over to an electron microscope image and a start point and an end point are determined with mouse dragging, whereby a rectangular area composed of the two points can be selected. In this way, original data is determined, and therefore an operation, such as desired copying or mirror inversion, can be easily performed. Rotation Button 202 With respect to the area selected with the area selection button 201, when this rotation button 202 is clicked, the area selected is rotated by 90 degrees clockwise. Another click on the button rotates the area further by 90 degrees. Further, by clicking the right mouse button, another window is displayed so that an arbitrary degree can be inputted, thereby rotating the entire CAD data by the designated degrees. This rotation operation is realized by rotating CAD data on the display without change in the contents of CAD data, and it can be realized with an easy operation as described above. Conventionally, while rotation has been handled only by performing raster rotation on the electron microscope image side, an overlay display is also enabled by rotating the CAD data side, too, thereby offering more alternatives. X-axis Mirror Inversion Button 203 A dashed-line area 507 shown in FIG. 9 is selected with the area selection button 201. Next, after clicking an x-axis mirror inversion button 203, by performing dragging from a start point of a vertical line 509 in FIG. 9 to an end point on the screen, an area 508 that is mirror-inverted with respect to the x-axis is created based on the dashed-line area 507. In this way, the operation of an x-axis mirror inversion for creating a semiconductor with the same structure can be easily performed. Y-axis Mirror Inversion Button 204 The dashed-line area 507 shown in FIG. 9 is selected with the area selection button 201. Next, after clicking a y-axis mirror inversion button 204, by performing dragging from a start point of a horizontal line 511 in FIG. 9 to an end point on the screen, an area 510 that is mirror-inverted with respect to the y-axis is created based on the dashed-line area 507. In this way, the operation of y-axis mirror inversion for creating a semiconductor with the same structure can be easily performed. Grid Creation Button 205 Repetitive patterns of a start point 401, an end point 402, and a unit vector 403 are found in FIG. 5. By creating this grid, it becomes possible to count the repetitive patterns on the screen, whereby the transfer to a defective position can be performed accurately. After a grid creation button 205 is clicked, by performing dragging from 401 to 402 shown in FIG. 5 on the screen, a grid is created that has a unit rectangle represented with an arrow 403 composed of the two points of the drag start point 401 and end point 402. Upon such operation, grid vertical/horizontal numbers 409, grid vertical lines 410, and grid horizontal lines 411, which are shown in FIG. 6, are automatically displayed. Further, when a new grid is created in one rectangular area of the previously created grid (in 2-2 of FIG. 3), a grid (243, 244) for a higher magnification display is created in the rectangular area. Further accurate transfer to a defective position can be easily performed with the grid for a higher magnification display. Since individual rectangular areas formed in the grid are provided with systematic numbers 236 for representing vertical/horizontal arrangement, it is possible to easily determine which grid contains an intended defective position. With regard to the numbers indicating grid vertical and horizontal positions, while two ways of representation, 5-7 and (5, 7), are possible when the grid is horizontally the fifth and vertically the seventh, for example, a desired expression can be designated with an ini file. Grid Correction Button 206 A grid correction button 206 is used for correcting a grid created with the grid creation button 205. The grid correction button 206 is clicked, and dragging is performed from 407 to 406 of FIG. 6 on the grid of the screen. By performing a calculation from the grid start point 401 to the end point 406, the size of the grid is corrected so that the grid end point 406 becomes a grid point without changing the grid start point and the number of grids. As a result, the grid displayed as shown in FIG. 6 is corrected to a grid size that accords with an SEM image as shown in FIG. 7. Further, by right-clicking the grid correction button 206 with respect to the grid created with the grid creation button 205 and by dragging the start and end points on the grid, the entire grid is translated. There are cases in which the entire grid is shifted to the lower left, for example, even when the grid size is matched. With this operation, the entire grid including the grid original points can be slid (parallel translation), whereby misalignment of the entire grid can be handled. Generally, even when a grid unit is specified, error from the start point becomes larger as the number of grids is increased, and thus it becomes impossible to perform counting. Thus, with the two corrective operations of grid correction and parallel translation, fine adjustment of the entire grid can be performed, so as to freely create a grid that accords with a SEM image. Copy Button 207 By selecting an area with the area selection button 201 and clicking a copy button 207, a copy destination rectangle corresponding to the mouse position is displayed. A copy destination is next determined with a mouse click. In this way, a desired copy operation can be easily performed. Cut Button 208 By selecting an area with the area selection button 201 and clicking a cut button 208, a copy destination rectangle corresponding to the mouse position is displayed. A cut destination is next determined with a click. The difference from the processing with the copy button 207 is that data in the area selected with the area selection button 201 remains without change in the case of copying whereas data in the area selected with the area selection button 201 disappears and is moved to the copy destination in the case of cutting. For example, in cases in which data is copied on a position slightly different from an expected position, a desired copying operation can be easily performed again by moving the position with this cutting processing. Circle Button 209 After clicking a circle button 209, dragging is obliquely performed from a start point to an end point on the screen. A maximum circle inscribed in a rectangular area composed of the start point and the end point is displayed. With this operation, it becomes possible to easily create a desired circle superimposed on a SEM image, so as to express a single plug. Oval Button 210 After clicking an oval button 210, dragging is obliquely performed from a start point to an end point on the screen. A maximum oval inscribed in a rectangular area composed of the start point and the end point is displayed. With this operation, it becomes possible to easily create a desired oval superimposed on a SEM image. Painted Circle Button 211 After clicking a painted circle button 211, dragging is obliquely performed from a start point to an end point on the screen. A maximum painted circle inscribed in a rectangular area composed of the start point and the end point is displayed. With this operation, it becomes possible to easily create a desired painted circle superimposed on a SEM image. Painted Oval Button 212 After clicking a painted oval button 212, dragging is obliquely performed from a start point to an end point on the screen. A maximum painted oval inscribed in a rectangular area composed of the start point and the end point is displayed. With this operation, it becomes possible to easily create a desired painted oval superimposed on a SEM image. Alignment Button 213 Since a click on an alignment button 213 displays a crosshair cursor in the middle of the screen on the system side, a point that can easily be a reference for alignment is adjusted, and again, the alignment button is clicked, whereby a mark (a box+a crosshair cursor) for alignment is displayed. This mark for alignment is stored in CAD data. It is important that this alignment is present at an easily recognizable position on the SEM image when CAD data and an electron microscope image are overlaid. In order to match a SEM image and CAD data, three alignment points that are not in a line are necessary. With this operation, matching of an electron microscope image and CAD data can be easily performed when an overlay display is carried out next. Character String Button 214 After clicking a character string button 214, characters are inputted in a character input 231. By mouse-clicking a start position for the character string, the character string is displayed on the screen. Since such character string is freely written, designers and inspectors, for example, can write comments, and information concerning defects can be shown on CAD data. Circle-attached Minimum Character Button 215 After clicking a circle-attached minimum character button 215, a character is inputted in the character input 231 and a position that is to be the center of a circle on a SEM image is clicked, whereby a circle-attached minimum character is displayed. Since the size of a plug is gradually reduced, the explanation of such plug needs to be expressed with small characters. With this operation, recognizable minimum characters can be easily expressed. Circle-attached Character Button 216 After clicking a circle-attached character button 216, a character is inputted in the character input 231 and dragging is obliquely performed from a start point to an end point on the screen. A maximum circle inscribed in a rectangular area composed of the start point and end point and the character are displayed. In this case, there is a limit to minimizing characters for display, and since the size of the circle differs from that of characters, the characters may not always be contained in the circle. With this operation, it becomes possible to unify the size of the circle and that of the plug on a SEM, easily write a desired circle and character at once, and easily perform the operation of creating a circle-attached character. Marker Buttons 217 and 218 The difference between marker buttons 217 and 218 is only their design. After clicking the marker buttons (217, 218), by performing dragging with the left button of a mouse on the screen, a maximum design inscribed in a rectangular area composed of the two points of drag start and end points is displayed. In this case, a marker with a constant size irrespective of display magnification is displayed on the screen. FIG. 12 shows examples in which the same CAD data is overlay-displayed with different display magnification. Wires 602 and 603, and a marker 604 are displayed on a high-magnification screen 601. When the display magnification is changed to a lower magnification, wires 606 and 607, and a marker 608 are displayed on a low-magnification screen 605. Because of the change in display magnification, the distance between the wires 602 and 603 is more separated, compared with the distance between the wires 607 and 608. However, the size of the marker 604 at high magnification remains unchanged with respect to that of the marker 608 at low magnification. This means that the coordinate positions only recognizable at high magnification can be easily recognized at low magnification. However, since a CAD data format used for the purpose of manufacturing semiconductors does not have a function of indicating such size that does not change depending on display magnification, it can be realized by using an advanced function format of a CAD. A general CAD navigation that does not have such advanced function has a disadvantage that these markers cannot be displayed. If markers with a constant size are displayed on the screen irrespective of magnification, a target can be easily found at low magnification. Thus, they are most suitable as alignment marks and marks attached to target defective positions. Rectangular Area Designation Button 219 After clicking a rectangular area designation button 219, by performing dragging with the left button of a mouse on the screen, a rectangular area composed of the two points of drag start and end points is displayed in the form of a frame. In this case, the frame is displayed with the attributes of a thickness 233 and solid line/dashed line 234 being effective. With this operation, rectangles can be easily drawn. Since horizontal/vertical lines are much more frequently used than diagonal lines, the operation for the rectangular area using horizontal/vertical lines is user-friendly and easily carried out. Painted Rectangular Area Designation Button 220 By clicking a rectangular area designation button 220 and performing dragging with the left button of a mouse on the screen, a painted rectangular area composed of the two points of drag start and end points is displayed. For example, when displaying an area on which measurement has been completed, the processing for a painted rectangular area can be substituted for a tick. With this operation, a desired rectangular area painting processing can be easily realized. Tick Button 221 By clicking a tick button 221 and mouse-clicking a tick center on the screen, a tick is displayed. With this operation, it becomes possible to easily recognize a plug on which measurement has been completed, and occurrence of measurement on an erroneous plug will be eliminated even after raster rotation is performed. By designating a layer different from those of other information, the display or the nondisplay of the tick can be easily changed. A sign such as “x” or “/” can be substituted for the tick. Anything will do as long as plugs on which measurement has been completed can be easily expressed. Arrow Button 222 By clicking an arrow button 222 and performing dragging with the left button of a mouse on the screen, an arrow pointing from the drag start point to the end point is displayed. This operation can be used for drawing a lead line when it is difficult to attach characters or signs to a plug due to a narrow area. An arrow 240 shown in FIG. 3 is an example of such arrow displayed with the arrow button 222. Horizontal/vertical Line Button 223 By clicking a horizontal/vertical line button 223 and performing clicking with the left button of a mouse on the screen, a straight line using the first click point as a start point and using an x-coordinate or a y-coordinate of the next click point as an end point coordinate (the value of the y-coordinate or x-coordinate of the end point is the same as that of the start point) is displayed either vertically or horizontally, depending on closeness. If clicked continuously, horizontal lines and vertical lines connected in the form of a polygonal line are alternatively displayed. This operation is mainly used for displaying a wire as a landmark, and by indicating wires, it becomes easier to determine a target position. Further, since the thickness, type (solid/dashed), and color of the line can be changed, the line can be used as a landmark. Lines 239 and 245 of FIG. 3 are figures drawn with this horizontal/vertical line button 223. Straight Line Button 224 By clicking a straight line button 224 and performing dragging with the left button of a mouse on the screen, a straight line connecting the drag start point to the end point is displayed. Since the thickness, type (solid/dashed), and color of the line can be changed, the line can be used as a landmark. Polygonal Line Button 225 By clicking a polygonal line button 225 and performing clicking with the left button of a mouse on the screen, a line connecting the first click point to the next click point is displayed. Another click displays another line connecting the point clicked second last to the point clicked last. If this processing is repeated, a polygonal line can be displayed. Further, since the thickness, type (solid/dashed), and color of the line can be changed, the line can be used as a landmark. Redo Button 226 By left-clicking a redo button 226, the last processing is cancelled. By right-clicking the redo button 226, the last cancel processing is restored. For example, when a position slightly different from an intended position is obtained with an operation of y-axis mirror inversion, by using the redo button 226, the operation of y-axis mirror inversion can be easily redone. Polygon Button 227 By clicking a polygon button 227 and performing sequential clicking with the left button of a mouse on the screen, gaps among the click points are sequentially connected with a straight line. A polygonal line is displayed by repeating this processing. By performing double-clicking on the last point or by clicking the polygon button 227 again, a straight line connecting the last point and the first point is drawn, whereby a closed polygon is drawn. Further, since the thickness, type (solid/dashed), and color of the line can be changed, the line can be used as a landmark. A polygon 242 (see FIG. 3) drawn with the polygon button 227 can be used when a target is indicated by area. Painted Polygon Button 228 By clicking a painted polygon button 228 and performing sequential clicking with the left button of a mouse on the screen, gaps among the click points are sequentially connected with a straight line. A polygonal line is displayed by repeating this processing. By performing double-clicking on the last point or by clicking the painted polygon button 228 again, a straight line connecting the last point and the first point is drawn. Painting is carried out when a closed convex-type polygon is formed. In this way, a desired painted polygon can be obtained, so as to create an area as a landmark. Layer Selection Box 229 By inputting numbers in a layer selection box 229, layers divided into display hierarchies can be designated. The layer selection is effective in all the drawing processing on the screen. Smaller numbers indicate lower layers. Necessary layers can be displayed by dividing layers and storing them in CAD data. Color Selection Part 230 When selecting red, blue, or green, the color to be drawn is designated by clicking a radio button. When designating colors other than the above colors, selection is made using a color pallet. Character Input 231 When inputting characters for the character-related buttons 214 to 216, the region of a character input 231 shows an input region of characters from a keyboard. Specifically, by inputting D in the character input 231 and creating the circle-attached character 215, character D with a circle is displayed on the screen. Grid Number Selection 232 By ticking two checkboxes indicating an x-axis grid number and a y-axis grid number, numbers such as x- and y-grid number 236 (see FIG. 3) are displayed. Thickness Selection 233 It is used for buttons for line-drawing processing, and the thickness of a line can be selected. Line Type Selection 234 It is used for buttons for line-drawing processing, and either a solid line or a dashed line can be selected. By using the input part composed of the area selection button 201 to the line type selection 234 for selecting attributes, it becomes possible to draw figures, symbols, characters, and the like on an electron microscope image, as shown on the left side of FIG. 3. After CAD data indicating defective positions are created on the screen, a saving-to-a-file button on a different screen or a pull-down menu is selected, so as to create a CAD file in which figures, symbols, characters, and the like drawn on the electron microscope image are stored as CAD data. In FIG. 3, a grid is contained in a grid 2-2, thereby expressing a plurality of hierarchies. FIG. 4 shows an enlarged view of a portion of the hierarchies. FIG. 4 is a diagram in which probing is being performed on individual plugs, a drain 307, a well 308, a gate 309, and a source 310, using probes 301 to 304. A grid is indicated with dashed lines 305 and 306. In the figure, by using the circle-attached character button 216, circle-attached character D is drawn superimposed on the position of the drain on an electron microscope image, circle-attached character W on the position of the well, circle-attached character G on the position of the gate, and circle-attached character S on the position of the source. If an initial letter is attached to a plug in this way, it becomes clear that which plugs have already been probed. FIG. 5 shows a diagram for explaining a method for creating grids. After the grid creation button 205 shown in FIG. 3 is clicked, dragging is performed from position 401 to position 402. The arrow 403 is displayed during dragging. Upon completion of dragging at the position 402 shown in FIG. 5, grids 410 and 411 are drawn as shown in FIG. 6. The drag start position 401 becomes the grid origin, and a grid having a unit grid rectangle with a diagonal line connecting the position 401 and the position 402 is drawn. Referring to the figure, while a grid ┌1-1┘ in the bottom-left corner seems correctly displayed, when a grid ┌10-5┘ in the top right is referred to, it can be seen that an intended grid is not accurately created at the positions of the plugs. In this case, by left-clicking the grid correction button 206 shown in FIG. 3 and clicking the top-right grid point of the grid ┌10-5┘ as if grabbing it, a grid point 407 starts to move in accordance with the mouse operation. During this operation, an additional line 408 from the origin is also displayed. By performing dragging as indicated by an arrow 405 and completing mouse-dragging at a correct position 406, correction of the distance between grid lines is performed such that the grid size is equalized from the origin, whereby the grids 401 and 411 as shown in FIG. 7 are displayed. Generally, both the origin 401 of FIG. 6 and the grid point 407 used for error determination are not contained in one screen of an electron microscope, and therefore, the entire grid size from the origin is equally corrected without the origin 401 on the screen. When an accurate grid is created by repeating this processing several times, correction is completed by left-clicking the grid correction button 206 of FIG. 3. Further, in cases where the entire grid needs to be equally slid because the origin position is inaccurate while the grid size is adjusted, the grid correction button 206 of FIG. 3 is right-clicked and the entire CAD data grid alone is moved by grabbing an intersection point of the grid, whereby correction is enabled. As a specific example 1, regarding a semiconductor memory, particularly a DRAM in which similar plugs are regularly arranged, it is not easy to attach types to plugs that need to be measured, using an electron microscope image alone. FIG. 8 shows an example of copy processing with respect to plugs. After the area selection button 201 is clicked, a copy source area 504 is designated by creating a rectangle with a mouse drag. An example in which copying 207 is performed from the copy source 504 to the areas of copy destinations 505 and 506 is shown. Further, an example in which cutting processing from 502 to 503 is performed using the cut button 208 is shown. Of course, when the copy source area is made larger, more copies can be made at once. These copy processing and cutting processing are performed such that a circle-attached character S is superimposed on the position of a source of an electron microscope image displayed on the screen, a circle-attached character W on the position of a well, a circle-attached character D on the position of a drain, and a circle-attached character G on the position of a gate. In this way, a large number of symbols corresponding to the position of each plug on an electron microscope image can be efficiently attached. Further, regarding constitution of a semiconductor plug, mirror inversion is often seen. FIG. 9 shows an example of processing with respect to a device that is mirror-inverted. Plugs formed symmetrically with respect to boundaries 509 and 511, which are mirrors, appear on an electron microscope image. In the present example, an operator first draws a group of circle-attached characters such that they are superimposed on the positions of the sources, drains, wells, and gates of the electron microscope image, as shown in an area 507. By using the group as a copy source, a mirror-inverted copy is created in a copy destination area 508 symmetrical with respect to the boundary 509, using the x-axis mirror inversion button 203. Further, by using the group of symbols drawn in the area 507 as a copy source, a mirror inverted copy is created in a copy destination area 510 symmetrical with respect to the boundary 511, using the y-axis mirror inversion button 204. Thus, by using the processing buttons for copying, cutting, and mirror inversion, usability is significantly improved, compared with cases in which identification symbols are individually attached. As a specific example 2, FIG. 10 shows an example of how an area and a tick are used. An area 512 is an area where electrical characteristics need to be measured by bringing the probes of the semiconductor inspection equipment into contact therewith. If completion of measurement per transistor is marked with a tick 513, the next transistor to be measured can be clarified. Further, the lower diagram in FIG. 10 shows a representation of the upper diagram rotated by 90 degrees with the rotation button 202. Rotation is indispensable due to ease of probe application. However, rotation makes it difficult to determine to what extent measurement has been completed. Particularly, when magnification is changed and a similar pattern is repeated, mistakes can be often made. However, as shown in the figure, if the measurement range is clear, the type of each plug is recognizable, and the completion of measurement is indicated with a tick, it is very obvious where to measure next. Further, a method for painting areas that have already been measured with a painted polygon button 228 can clarify measurement completed portions in the same sense. As a specific example 3, as an example of an actual semiconductor, there are cases in which a single transistor has two meanings. FIG. 11 shows such a case. While numeral 516 denotes a single transistor and numeral 517 also denotes a single transistor, the problem is that a plug 518 refers to the gate of the transistor 516 and refers to the drain of the transistor 517. In this case, if an initial letter and a tick are inserted in the same way as described above, it is impossible to determine which plug is overlapped. In such case, by separating a layer into two and displaying either of the two, plugs are clearly determined. Further, regarding ticks in this case, too, it is clear that determination is easier if ticks are attached to individual layers. FIG. 12 shows an example of displaying markers. The purpose of displaying markers is not for expressing the actual size thereof but for using them as landmarks. Namely, even when a mark with an accurate size is attached to a defective position, it is very troublesome to find the mark when displayed at low magnification. Further, if a larger mark is attached, the accuracy is decreased because it becomes too large when displayed at high magnification or it interferes with information in question. Thus, by displaying the individual markers 604 and 608 with a constant size irrespective of magnification between the high-magnification display screen 601 and the low-magnification display screen 605, the above problems are solved. Numerals 602, 603, 606, and 607 in FIG. 12 refer to wires. Of course, the distance between the wires is changed depending on magnification. Each of the examples described above merely shows an example of the present invention, and the present invention is not limited to such examples. Further, each of the methods of operation also indicates a mere example, and thus the inventive methods of operation are not limited to such methods of operation.
052001440	summary	BACKGROUND OF THE INVENTION The present invention relates generally to electrical simulators of nuclear fuel assemblies, and more particularly to a heater element design for electrically-powered heater assemblies having concentric fuel tubes. The heating produced in nuclear fuel assemblies can be simulated in theory by electrically powered assemblies which are designed to represent all or part of an actual fuel assembly. Current machining technology has been successfully applied in the construction of cylindrical rod heater elements for simulation of nuclear fuel rods. However, for other nuclear fuel assembly geometries, such as the concentric cylindrical fuel tubes in the Savannah River Laboratory (SRL) production reactors, current machining technology cannot be used to build electrically powered cylindrical heater tube elements. A major reason is the small thicknesses of the components in a cylindrical heater tube which do not have the material structural strength to withstand machining. Additionally, the design of cylindrical heater elements using only MONEL (Trademark of a nickel-copper alloy and a nontypical nuclear material) and based on current machining technology requires power leads to enter and exit from both ends of the cylindrical heater element. Such designs also contain mechanical components associated with the power leads that perturb the coolant flow at the entrance and exit of the electrically powered simulators. The need for electrical powered simulators of the production reactor fuel assemblies originates from studies on the safety aspects of these reactors. In accident conditions, such as the loss-of-coolant accident from a failure (break) in a primary reactor coolant system component, the loss of coolant that removes the nuclear generated heat from the core causes the fuel assemblies to go into a sustained heatup. If coolant is not restored the core heatup will result in fuel melting temperatures being reached which is the initial phase of what is termed a "servere reactor accident" in which fuel damage and release of radioactivity occurs. In order to prevent such severe accidents from occurring, or as a minimum reducing the probability of such an accident to a low value (1 in 10.sup.6), accident recovery systems and operator procedures must be developed that would contain or limit the fuel temperatures in accident conditions to the non-damage region. The successful development of systems and procedures depends in part on experimental data obtained from experiments simulating the accident conditions. Understanding the physics involved provides the means to design the methods to limit the fuel temperatures to non-damage values. Thus, the need for accurate simulation of the fuel assemblies with electrically powered simulators becomes apparent. Without very good experimental data the analysis of such accident conditions in the reactors in the reactors is entirely theoretical. Theoretical analyses alone are not sufficient to obtain licensing to operate nuclear plants. Accordingly, it is an object of the present invention to provide a nuclear fuel assembly electrical simulator having a full scale cross-section with power leads connected only at the top of the simulator. A further object of the present invention is to provide a simulated nuclear reactor heater assembly which provides prototypical fuel assembly entrance and exit geometry. Yet another object of the present invention is to provide a simulated nuclear reactor heater assembly with preservation of fuel assembly surface materials that transfer heat to the coolant. SUMMARY OF THE INVENTION This invention provides an apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and including a plurality of concentric heater tubes having electrical circuitry connected to power leads. The heater tubes are radially spaced from each other. An outer target tube (unheated) and an inner target tube (unheated) is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly in generally open and allows for the electrical power leads to be routed into the assembly interior to the inner target tube. The bottom of the assembly is generally open and includes means for completing the electrical circuitry, connecting all of the heater tubes in a series resistance circuit. The electrical resistance in each heater tube is mechanically designed to provide electrical heat with the same power profile as in a nuclear reactor. The heater tubes are aluminum. Each heater tube is connected to the power source by an embedded layer of an electrical conductor along the length of the tube. The conductor layer is surrounded by first and second ceramic insulating layers. The inner aluminum layer is a machined tube defined as the "base" tube. The ceramic insulating layers, the conductor layer, and the outer aluminum layer are all thermal sprayed in succession onto the base tube to form the electrical simulator tube. Each heater tube electrical conductor layer consists of two semicircular elements which are isolated from each other by an electrical insulating ceramic extending substantially along the entire length of the tube but not the full length thereof, such that each tube is joined at one end and is not joined at the other end. Each heater tube also includes a multiplicity of thermocouples embedded at various locations within the heater tubes.
048083186	summary	BACKGROUND OF THE INVENTION The invention relates to a method for immobilizing radioactive wastes for long term storage. More specifically, the invention relates to a method for recovering cesium from solutions containing cesium together with other metal values and for immobilizing radioactive cesium in a highly stable, nonvolatile and insoluble product suitable for long-term storage. The principal long-term problem caused by nuclear reactor accidents is contamination of the environment with radioactive material as was evidenced by the Chernobyl nuclear reactor accident in 1986. Cesium is a particular problem in this respect because it is very volatile and can be carried in the upper atmosphere for long distances. Therefore, suitable methods of decontaminating the environment, i.e. water and soil, of cesium before its ingestion by animals or humans are highly desirable. Also desirable are methods for recovering radioactive cesium from the gastrointestinal tracts of contaminated animals. Certain cation exchange resins and various cation exchangers are available which are selective for the recovery of cesium from solution. These include clay minerals and zeolites, both naturally occurring and synthetic. Furthermore, naturally occuring mordenite zeolite has been mixed with animal feed to remove ingested cesium-137 from the gastrointestinal tracts of contaminated animals. Other natural zeolites include clinoptilolite, erionite and chabazite. In addition to recovering the ions, the radioactive ions must be immobilized in a form suitable for longterm storage so that they cannot be released back into the environment or leached from the storage medium into the surrounding soil or groundwater. Many methods and materials have been developed over the years for immobilizing various nuclear wastes, and especially cesium, for long-term storage. For example, U.S. Pat. No. 3,161,601 dated Dec. 19, 1964, and assigned to the common assignee, incorporated the radioactive cesium into a glass containing cesium oxide, alumina, phosphate and an additive such as lanthanum or zirconium. Another method was to incorporate the cesium-137 as cesium carbonate with spodumene or with a mixture of silica and kaolin at a temperature of at least 1000.degree. C. to form a synthetic pollucite. Another process mixed an inorganic zeolite containing radioactive cesium with about 20% additives, mainly iron and calcium oxides, which was melted at 1500.degree. C. and control cooled to form an iron-enriched basalt. U.S. Pat. No. 4,537,710, also assigned to the common assignee, describes a cation exchanger which is a modified tobermorite containing aluminum, that is selective for a small number of ions including cesium. The advantage of the modified tobermorite is that it is compatable with matrix materials such as concrete and, therefore, more resistant to leaching. Other, less satisfactory, methods involve incorporating the radioactive waste directly into a matrix material such as asphalt or concrete before emplacement for storage. Almost all of the materials or processes suitable or selective for the separation and recovery of cesium from contaminated water or radioactive waste streams require additional processing to immobilize the cesium, in order to prevent the radioactive ions from being leached or otherwise separated from the storage medium. For example, the cesium ions may be eluted from the ion exchanger and incorporated directly into the matrix material for storage. Preferably, the recovery material (ion exchanger) containing the radioactive ions, is itself incorporated into the storage medium, such as a glass or cement, for storage. Alternatively, the radioactive ions may be mixed with other inorganic materials and by applying heat and/or pressure, formed into a synthetic mineral which is satisfactory for storage. Thus, most processes require several steps to recover and prepare the cesium for storage. This increases the cost of preparing the radioactive ions for storage. Furthermore, some processes require high pressures and/or temperatures which in addition to increasing costs, increases the opportunities for the loss of radioactive material. Finally, many of the storage materials are not wholly suitable for long-term storage because leaching of the radioactive ions can occur. SUMMARY OF THE INVENTION It has been found that a modified phlogopite mica is very ion selective for cesium ions, even from solutions which also contain sodium and calcium ions. Furthermore, it has been found that the modified phlogopite mica will trap the cesium ions in such a manner that the phlogopite containing the cesium is suitable for emplacement for long-term storage, with little or no additional processing. The modified phlogopite mica of the invention is a phlogopite mica which has been hydrated and in which the potassium ions have been replaced by sodium ions. The invention is a process for the separation and recovery of cesium ions from a feed solution containing cesium ions, and which may contain other metal ions, by contacting the solution with the modified phlogopite which is a hydrated sodium phlogopite mica whereby the cesium ions are selectively taken up by the modified phlogopite while the other ions remain in the solution, and separating the modified phologopite containing the cesium ions from the feed solution thereby recovering the cesium ions. The invention is also a process for fixing radioactive cesium for long-term storage by contacting a solution containing radioactive cesium with the modified phlogopite which is a hydrated, sodium mica maintaining the contact until sufficient cesium is taken up by the modified phlogopite to reduce the c-axis spacing an amount sufficient to immobilize the cesium, thereby fixing the radioactive cesium ions for long-term storage. Alternatively, the cesium may be fixed by heating the modified phlogopite containing the cesium to a temperature sufficient and for a period of time sufficient to reduce the c-axis spacing thereby fixing the cesium ions in the modified phlogopite. Since the modified phlogopite of the invention is very selective for cesium ions, it is especially useful for the recovery of radioactive cesium ions which are present in radioactive waste solutions along with other metallic ions including sodium and calcium. The solutions may be either low level, intermediate or high level nuclear wastes. It is also useful for the recovery of cesium-137 from large volumes of water containing low levels of cesium such as nuclear reactor coolant systems which have become contaminated because of fuel element ruptures or from stream or water supplies which have become contaminated due to Cs.sup.137 fallout. Furthermore, because of its inertness and stability in an acidic environment, the modified phlogopite is suitable for ingestion by animals for the recovery and removal of ingested radioactive cesium fron the gastrointestinal tract, result from nuclear mishaps like the Chernobyl incident. It is therefore one object of the invention to provide a process for recovering cesium. It is another object of the invention to provide a process for recovery of cesium from solutions containing cesium together with other ions. It is a further object of the invention to provide a process for fixing cesium for the long-term storage. It is still another object of the invention to provide a one-step process for recovering and fixing cesium ions for long-term storage. Finally, it is the object of the invention to provide a process for recovering and immobilizing radioactive cesium ions for long-term storage which does not require conditions of high temperature or high pressure.
	claims	1. A sample manipulation device, comprisingan observation unit which is used to observe a sample and to select a target position at which a portion to be removed from the sample is located,a specimen stage which receives the sample,a manipulation tool, which is spatially shiftable relative to the observation unit and comprises an exchangeable manipulation tip by which said portion is removed from said sample,a control unit which controls the shifting of the manipulation tool; andan optical position measurement unit which is connected to the control unit,wherein the position measurement unit and the observation unit are each provided as microscopes, the specimen stage being arranged between the observation unit and the optical position measurement unit and the position measurement unit comprising an objective having a depth of focus on the order of magnitude of the manipulation tip; andwherein the position measurement unit is used to determine the actual position of the manipulation tip so that specific shifting of the manipulation tip to the target position can be carried out by the control unit. 2. The sample manipulation device as claimed in claim 1, further comprising an image-recording and image-processing unit provided in the position measurement unit. 3. The sample manipulation device as claimed in claim 1, wherein the specimen stage is shiftable relative to the observation and position measurement units. 4. The sample manipulation device as claimed in claim 1, wherein the position measurement unit is shiftable along its optical axis. 5. The sample manipulation device as claimed in claim 1, wherein the optical axes of the objectives of the observation and position measurement units are parallel to each other. 6. The sample manipulation device as claimed in claim 5, wherein the optical axes coincide. 7. The sample manipulation device as claimed in claim 1, further comprising laser-optical tools coupled into the object plane by the objective of the position measurement unit. 8. The sample manipulation device as claimed in claim 1, wherein a stereo microscope or an incident-light microscope is provided as the observation unit. 9. A sample manipulation device, comprising an observation unit, which is used to observe a sample and to select a portion to be manipulated, a specimen stage which receives the sample, and an optical manipulation unit, by which the selected portion can be manipulated, said specimen stage being arranged between the observation and manipulation units and said manipulation unit being provided as a microscope, wherein a stereo microscope is provided as the observation unit. 10. The sample manipulation device as claimed in claim 9, wherein the specimen stage is shiftable relative to the observation and manipulation units. 11. The sample manipulation device as claimed in claim 9, wherein the manipulation unit is shiftable along its optical axis. 12. The sample manipulation device as claimed in claim 9, wherein the optical axes of the objectives of the observation and manipulation units are parallel to each other. 13. The sample manipulation device as claimed in claim 12, wherein the optical axes coincide. 14. The sample manipulation device as claimed in claim 9, wherein the manipulation unit comprises laser-optical tools. 15. The sample manipulation device as claimed in claim 1, further comprising an image-recording unit and an image-processing unit provided in the observation unit. 16. The sample manipulation device as claimed in claim 1, further comprising laser-optical measurement devices coupled into the object plane by the objective of the position measurement unit. 17. The sample manipulation device as claimed in claim 9, wherein the manipulation unit comprises laser-optical measurement units.
	claims	1. A fuel assembly for a nuclear reactor comprising:a parallel, spaced array of a plurality of elongated nuclear fuel rods having an axial length;a lower nozzle;an upper nozzle,wherein the plurality of elongated nuclear fuel rods is supported between the lower nozzle and the upper nozzle; anda plurality of spaced grids arranged in tandem along the axial length of the fuel rods between the upper nozzle and the lower nozzle, each of the plurality of spaced grids or portions or parts thereof comprising:a plurality of sequential layers sintered together, each of the sequential layers of the plurality of sequential layers comprising:sintered powder, which comprises one or more ternary compounds of the general formula:Mn+1AXn (I)wherein M is a transition metal, A is an element selected from the group A elements in the Chemical Periodic Table, X is selected from the group consisting of carbon and nitrogen, and n is an integer from 1 to 3,wherein the sintered powder for each of the sequential layers of the plurality of sequential layers comprises the same one or more ternary compounds or different one or more ternary compounds, andwherein the sequential layers remain discrete from each other after being sintered together. 2. The fuel assembly of claim 1, wherein M is selected from the group consisting of titanium, zirconium and niobium. 3. The fuel assembly of claim 1, wherein A is selected from the group consisting of aluminum, silicon and tin. 4. The fuel assembly of claim 1, wherein the one or more ternary compounds are selected from the group consisting of Ti2AlC, Ti3AlC2, Ti4AlN3, Ti2SiC, Ti3SiC2, Ti3SnC2, Zr2AlC, Zr2TiC, Zr2SnC, Nb2SnC, Nb3SiC2, (ZrxNb1−x)2AlC wherein x is greater than zero and less than 1. 5. The fuel assembly of claim 1, wherein a molar ratio of M to A to X can be selected from the group consisting of 2:1:1, 3:1:2 and 4:1:3. 6. The fuel assembly of claim 1, wherein the one or more ternary compounds each has a density of greater than 85% of its theoretical density. 7. The fuel assembly of claim 1, wherein the one or more ternary compounds each has a density of greater than 95% of its theoretical density. 8. The fuel assembly of claim 1, wherein the fuel assembly is employed in a water reactor selected from the group consisting of a pressurized water reactor, boiling water reactor and heavy water reactor.
	claims	1. A sample fabricating method including:a first step of irradiating a first ion beam at a surface of a specimen in order to form a first trench relative to the surface of the specimen, the surface of the specimen being set at a predetermined angle that is inclined relative to a right angle between the surface and an optical axis of the first ion beam during the first step of irradiating;a step of rotating the specimen approximately 180 degrees about a rotation axis perpendicular to the surface of the specimen after the first irradiating step while maintaining the specimen at the predetermined angle;a second step of irradiating a second ion beam at the surface of the specimen in order to form a second trench relative to the surface of the specimen, the first and second trenches being formed to converge within the specimen to form a bottom of a sample, the surface of the specimen being set during the second step of irradiating at the same predetermined angle at which the surface of the specimen is set during the first step of irradiating;a step of separating the sample from the specimen;a third step of irradiating a third ion beam at a surface of the specimen in order to form a third trench relative to the surface of the specimen, the surface of the specimen being set at the same predetermined angle before the step of the rotating;a fourth step of irradiating a fourth ion beam at the surface of the specimen in order to from a fourth trench relative to the surface of the specimen, the third and fourth trenches being formed to converge within the specimen to form a bottom of a second sample, the surface of the specimen being set at the same predetermined angle after the step of the rotating; anda step of separating the second sample from the specimen. 2. A sample fabricating method according to claim 1, wherein irradiation of the first ion beam is conducted so as to form the predetermined angle from 30 degrees to 75 degrees between the surface of the specimen and the optical axis of the first ion beam irradiated. 3. A sample fabricating method according to claim 1, wherein the sample separated from the specimen or prepared to be specimen from the specimen is extracted by using a probe. 4. A sample fabricating method according to claim 1, wherein the specimen is a wafer. 5. A sample fabricating method comprising:a step of placing a sample on a specimen stage;a first processing step of processing the sample by irradiating the sample in a first irradiating position with an ion beam deflected relative to a sample placement face of the specimen stage to define an irradiation angle therebetween which is larger than 0° and smaller than 90°;a second processing step of processing the sample by turning the specimen stage by about 180° about an axis perpendicular to the sample placement face without tilting the sample placement face with respect to the axis to bring the sample into a second irradiating position, and irradiating the sample again in the second irradiating position with the deflected ion beam, wherein the irradiation angle is the same for said first and second irradiating positions; anda step of extracting a sample piece obtained by the second processing step,wherein the changing of the optical axis of the ion beam to decrease or increase the irradiation angle is realized by an electric deflecting mechanism. 6. The method of claim 5, wherein the irradiation angle is in a range from 30 degrees to 75 degrees. 7. The method of claim 5, wherein, the sample is a wafer having a diameter of at least 200 mm. 8. The method of claim 5, wherein the method is a method of fabricating a first sample piece in a first region of the sample and a second sample piece in a second region of the sample, a first trench and a third trench are formed in the sample in the first processing step and a second trench and a fourth trench are formed in the sample in the second processing step, a bottom of the first sample piece is formed by the first and third trenches and a bottom of the second sample piece is formed by the second and fourth trenches, and the step of extracting the sample piece is a step of the extracting the first and the second sample pieces. 9. A sample fabricating method including:a first step of irradiating a first ion beam at a surface of a specimen in order to form first and third trenches relative to the surface of the specimen, the surface of the specimen being set at a predetermined angle that is inclined relative to a right angle between the surface and an optical axis of the first ion beam during the first step of irradiating;a step of rotating the specimen approximately 180 degrees about a rotation axis perpendicular to the surface of the specimen after the first irradiating step while maintaining the specimen at the predetermined angle;a second step of irradiating a second ion beam at the surface of the specimen in order to form second and fourth trenches relative to the surface of the specimen, the first and second trenches being formed to converge within a first region of the specimen to form a bottom of a first sample and the third and fourth trenches being formed to converge with a second region of the specimen to form a bottom of a second sample, the surface of the specimen being set during the second step of irradiating at the same predetermined angle at which the surface of the specimen is set during the first step of irradiating; anda step of separating the first and second samples from the specimen. 10. A sample fabricating method according to claim 9, wherein irradiation of the first ion beam is conducted so as to form the predetermined angle from 30 to 75 degrees between the surface of the specimen and the optical axis of the first ion beam irradiated. 11. A sample fabricating method according to claim 9, wherein the first and second samples separated from the specimen is extracted by using a probe. 12. A sample fabricating method according to claim 9, wherein the specimen is a wafer.
	abstract	A radiation photographing apparatus has a radiation image receiving portion for receiving radiation transmitted through an object and obtaining a radiation transmission image, and a grid to be disposed on the object side of the radiation image receiving portion. The grid includes a scattered ray removing member or a radiation detector. The grid is constructed for movement also to the side opposite to the object side of the radiation image receiving portion.
	description	This application claims priority to foreign European Patent Application EP 08425546.2, filed Aug. 7, 2008, the disclosure of which is incorporated herein by reference in its entirety. The present invention relates to a shielding device for optical and/or electronic apparatuses, in particular for space telescopes. In the technical field of space applications, there is an urgent need to selectively shield optical and/or electronic apparatuses from undesired electromagnetic radiation, wherein these apparatuses are carried on board of satellites or generally on board of space vehicles. For example, in the particular field of space telescopes, during observation missions of scientific objectives, more and more often huge telescopes are used, called “Large Telescope”, which are provided on board of satellites. The most recent development of optical telescopes on board of satellites foresees in particular an architecture based on two or more satellites operating within a formation. Telescopes of this type generally comprise a first satellite, called “Mirror Spacecraft”, and a second satellite, called “Detector Spacecraft”. The Mirror Spacecraft houses focusing devices and has to point these devices towards scientific objectives to be observed, whereas the Detector Spacecraft carries the detector, on which images are focused, which have then to be processed, stored and/or transmitted to earthbound stations. In order to ensure adequate operation of such telescopes, it is necessary to provide a shielding device, for shielding the detector of the Detector Spacecraft from incident electromagnetic radiation, having a wavelength which is characteristic of sources at which the optics of Mirror Spacecraft are aimed, but which originate from sources, which are not aligned with the optics. Shielding devices of the known art envisage the use of large cylindrical baffles, called “Shielding Baffle” or “Collimator”, which are arranged on board the Detector Spacecraft, around the detector. However, such shielding devices have some drawbacks. Above said cylindrical baffles are in fact large monolithic structures, for instance including a core made of aluminum alloy and a cover made of carbon fiber composite material, which are characterized by high mass and considerable bulk. This is very important in the case of space applications, where an increase in mass with respect to payload causes a huge increase of costs and where available payload volume at launch generally represents a critical parameter. An object of the present invention is to provide a shielding device for optical and/or electronic apparatuses, which is able to remove above said drawbacks, related to the known art. This and other objects are achieved by using a shielding device for optical and/or electronic apparatuses, which may cooperate with incident electromagnetic radiation, in particular for space telescopes, the shielding device including at least a filter provided for interacting with said incident electromagnetic radiation, for selectively filtering said radiation; and a support structure for the filter; wherein the support structure is an inflatable structure, which is able to achieve an operating stand-by configuration, in which it is substantially folded together, and an active operating configuration, in which it extends along a longitudinal extension axis and is essentially completely unfolded, the filter moreover including a filter body, which is provided in order to be transversely positioned with respect to said longitudinal extension axis, when the support structure reaches its active operating configuration. Another object of the present invention is a space vehicle including a shielding device described in the previous paragraph, and, in a particular embodiment, wherein said vehicle is a satellite including a space telescope. In the drawings, same or similar elements are indicated by same numeral references. With reference to FIG. 1, 10 generally indicates a space telescope, in this example, a telescope for observing X-ray sources, including a pair of space vehicles, which may operate in a formation. More in particular, the pair 10 of space vehicles respectively comprises a Mirror Spacecraft 12 or MSC and a Detector Spacecraft 14 or DSC. The MSC carries on board focusing devices and observation optics, such as special mirrors or other types of optics, which are called “mirrors” in the following, and orients them towards the scientific objectives to be observed, such as a celestial body S1. The DSC carries on board an optoelectronic apparatus 16 or detector (schematically shown in FIG. 6), on which images are focused, which may then be processed, stored and/or transmitted to earthbound stations. More in particular, in the present example, the detector 16 comprises a plurality of optoelectronic sensors (not shown in figures), which may cooperate with incident electromagnetic radiation X1, X2 (FIG. 1), i.e. electromagnetic radiation which is generally within the nominal field of view of detector 16. The field of view of detector generally is the solid angle in the sky visible by detector. In the present example, without shielding devices, the nominal field of view is in particular of hemispherical shape, i.e. it is substantially equal to one half of the sky. With reference to FIG. 1, on board the DSC a shielding device 20 according to a currently preferred embodiment of the present invention is provided. The shielding device 20 has to selectively shield the detector from incident electromagnetic radiation X1, X2, so that detector 16 is almost exclusively struck by incident electromagnetic radiation originating from the source, which is aimed at by the telescope, i.e. desired radiation X1. In this example, this desired radiation is coming from MSC 12. In other words, shielding device 20 is capable of shielding the detector 16 from incident electromagnetic radiation coming from sources, which are different with respect to the one, which is aimed at by the telescope, i.e. the undesired radiation X2. In this particular case, regarding an X-ray telescope, this undesired radiation comprises for example incident electromagnetic radiation coming from X-ray sources, such as celestial bodies S2, which are not the one at which the telescope is aiming. With reference to FIG. 4, shielding device 20 comprises a support structure 22, for supporting at least a shielding element or filter 24, 26, which is capable of cooperating with at least a portion of incident electromagnetic radiation. Advantageously, support structure 22 is an extensible structure, and more in particular, an inflatable structure. The structure 22 is able to achieve a stand-by operating configuration (FIG. 2), in which it is essentially folded together, and an active operating configuration (FIG. 4), in which it is essentially completely unfolded. In other words, the structure 22 is able to achieve a stand-by operating configuration, in which it is essentially deflated and compacted, and an active operating configuration, in which it is essentially completely inflated and extends along the longitudinal axis ZZ. Inflation of support structure is achieved, for example by means of a gas, by using a known pressurizing device. From now on, when not otherwise stated, it is implied that the support structure 20 is in the active operating condition. According to an advantageous embodiment, support structure 22 has a tubular shape, in this example a cylindrical shape, which extends around above said longitudinal extension axis ZZ. This axis, in the present example, is in particular coincident with focal axis ZZ of telescope 10. With reference to FIG. 5, in which a particularly advantageous embodiment of support structure 20 is schematically shown, this structure has a segmented structure, including a plurality of inflatable chambers 28, which are positioned, according to a contiguous arrangement, about the focal axis ZZ. In the embodiment of FIG. 5, inflatable chambers 28 are longitudinally extensible chambers, which preferably extend along the whole length of support structure. Such segmented structure advantageously allows an increase of bending rigidity of inflatable structure. According to an alternate embodiment, structure 22 may however be provided in another way, for example in order to provide only one inflatable chamber. Still referring to FIG. 4, shielding device 20 comprises a plurality of filters 24, 26, which may interact with at least a portion of incident electromagnetic radiation, in order to selectively filter this radiation. In the embodiment shown, filters 24, 26 are capable of filtering incident electromagnetic radiation according to its inclination with respect to focal axis ZZ of telescope, i.e. according to respective incident angles with said axis. According to an embodiment of the invention, filters 24, 26 each comprise a substantially plate-like filter body, to be transversely positioned with respect to focal axis ZZ, when the support structure 20 reaches the active operating configuration. In particular, in the appended figures, filters 24, 26 are rigid plates, which are perpendicular to focal axis of telescope. In the example shown, filters 24, 26 are made of a material which is opaque to X-rays, i.e. it is substantially not transparent with respect to the wavelength of such radiation, and does not emit secondary radiation, when struck by X-rays. Filters 24, 26 may for example be made of aluminum with an external protective layer made of carbon, or they may be honeycomb structures, made of aluminum and carbon or aluminum with a protective multilayer, including for example tantalum, tin, copper, aluminum and carbon. According to an embodiment, each filter body comprises a shielding portion, which in the example shown, is shaped like a shielding annulus 30, and a pass-through portion, which, in the example shown, is a pass-through aperture 32, surrounded by the shielding annulus. In other words, in the present example, wherein the shielding portions are annuluses, pass-through apertures 32 are circular apertures. Pass-through apertures 32 are each capable of letting incident electromagnetic radiation pass through, having incident angles within a respective first incident angle range. Shielding annuluses 30 may each shield the detector 16 (FIG. 6) from incident electromagnetic radiation having incident angles within a respective second incident angle range, which is distinct from first range. More in particular, shielding annuluses may shield from incident electromagnetic radiation having incident angles greater than incident angles of electromagnetic radiation passing through respective pass-through apertures. According to an embodiment, when the support structure is in the active operating configuration, the filters are aligned and mutually separated by predefined distances, along focal axis ZZ of telescope. In particular, in the embodiment shown in figures, filters 24, 26 are centered with respect to focal axis ZZ. According to a particularly advantageous embodiment, pass-through apertures 32 have dimensions transversal to focal axis ZZ, which decrease when distance between filters and detector 16 diminishes. When the size of pass-through apertures 32 is reduced, respective shielding annuluses 30 may shield the detector 16 from incident radiation forming incident angles, which increase with respect to focal axis. Preferably, also external diameter of shielding annuluses 30 decreases with a reduction of distance between filters and detector. According to a particularly advantageous embodiment, the supporting structure may be impregnated with a polymer resin, which may polymerize when exposed to solar radiation, in order to further stiffen the supporting structure. This allows an increase of rigidity of support structure, so that filters 24, 26 may be kept in the correct position even during maneuvers of DSC 14. A further advantage obtained with polymer resin is that it preserves the rigidity of support structure 22 even if the inflatable chambers 28 are damaged, for instance in case of loss of pressure due to perforation by small meteorites. According to an advantageous embodiment, shielding device 20 comprises a fixed baffle 34, which is provided at or near the focal plane of detector (FIG. 6). Such a baffle is provided for shielding the detector from incident radiation having incident angles so high, that it cannot be shielded by shielding annuluses 30. According to an embodiment, the fixed baffle 34 is a tubular rigid baffle, which extends around focal axis ZZ of telescope. Advantageously, the rigid baffle 34 extends inside the support structure. In other words, in this example, wherein the support structure is of cylindrical tubular shape, the fixed baffle is completely housed within the cavity formed by internal walls of cylinder. In the embodiment shown in figures, the fixed baffle 34 advantageously has dimensions, transversal to focal axis ZZ, which are less than those of pass-through apertures 32 of filters 24, 26. In this manner, the fixed baffle may pass through apertures, when support structure 22 goes from stand-by configuration to the active operating configuration, and vice versa. An example of operation of shielding device 20 is described in the following, according to an embodiment of the present invention. Initially, and during the whole launch phase, the support structure is in the stand-by configuration. In particular, this structure is folded together with filters 24, 26, preferably at base of fixed baffle 34 (FIG. 2). Once the space telescope has reached its orbit, the support structure is inflated by the pressurizing device, in order to achieve its active operating configuration. For a better understanding of operation of shielding device 20, reference should be made to FIG. 7. In this figure, a geometric transversal section of shielding device 20 is schematically shown, with the exception of support structure 22, according to a currently preferred embodiment of present invention. In this figure, in particular, focal plane 36 of detector 16, fixed baffle 34, and two filters 24, 26 are shown, wherein the latter are only partially shown for sake of simplicity. With reference to same figure, at a distance from focal plane of detector, which is equal to focal length of telescope 10, in this example approx. 20 (m), the mirror 38, which is provided on board the MSC 12 and an additional shielding element or filter 40, called “mirror skirt”, are schematically shown. The mirror skirt is analogous to filters 24, 26, and is provided on MSC around mirror 38, in order to shield it from a first portion of incident electromagnetic radiation. FIG. 7 also shows a plurality of limit incident angles α, β, γ, δ, which are respectively defined between some reference directions P1, P2, P3, P4 and focal axis ZZ of detector. Above said reference directions are bound to geometry of shielding device and are particularly suitable for defining the shielding action of filters 24, 26, of mirror skirt 40 and fixed baffle 34. Limit angles α, β, γ, δ of FIG. 7 are respectively in an increasing order, and for sake of clarity, are illustrated by reference to an axis YY, which is parallel to focal axis ZZ. Still with reference to FIG. 7, it is to be noted that mirror skirt 40 may shield from a first portion of undesired radiation, and more particularly, radiation having incident angles between angles αand β. In a similar manner, filter 24, including pass-through aperture having a major radius R1, and filter 26, including pass-through aperture having minor radius R2, are capable of shielding from undesired radiation having incident angles between angles β and γ and angles γ and δ, respectively. Finally, incident radiation having incident angles greater than angle δ is blocked by fixed baffle 34. To summarize: mirror 40, filters 24, 26 and fixed baffle are capable of cooperating with incident electromagnetic radiation, so that the whole system substantially operates like a collimator, in order to allow only incident radiation with an incident angle less than angle α to reach focal plane 36 of detector 16. Based on above description, one may understand that a shielding device of above said type is able to fully achieve intended goals, overcoming the drawbacks of devices of the known art. The provision of a shielding device including plate-like filters and an inflatable support structure, provides a significant reduction of mass of shielding device with respect to solutions of known art, which use large monolithic cylinders. Furthermore, the provision of an inflatable support structure allows a considerable reduction of bulk of shielding structure, in particular during launch phase, so that this device is rendered essentially compatible with any launch device. This is particularly important in the case of X-ray space telescopes. These telescopes would in fact require monolithic cylinders, of such a size as to be rendered incompatible even with respect to commonly used launchers, used for putting into orbit space telescopes. According to a particularly advantageous embodiment, the shielding device according to the present invention is capable of shielding the detector also from optical radiation, such as solar radiation, reducing the need for, or even eliminating the need for particular optical filters, used in the known art for shielding detectors operating at X-ray wavelength, from solar radiation. This is particularly advantageous since in the case of low energy X-ray detectors, above said optical filters also attenuate the useful X radiation. It is clear that modifications and/or variations may be introduced in the examples described and illustrated above. According to an embodiment of the invention, both the support structure and the filters may be manufactured in alternate ways, with respect to above embodiments. For instance, the support structure may have a square sectional shape, whereas the filters may comprise square or generally polygonal shaped perforated plates. Filters may also be produced in order to provide pass-through portions, which are different from pass-through apertures. The pass-through portions may for example be made of materials substantially transparent with respect to desired electromagnetic radiation. According to an embodiment, the filters may be provided in such a way to allow a frequency filtering instead of a spatial filtering. In general, the number and size of filters may vary according to specific needs. Anyway, it is to be noted that, in general, a shielding device according to the present invention may be employed also for shielding optical and/or electronic devices, which differ from optics of a space telescopes, for instance those used for protecting telecommunication devices or other instruments provided on board of satellites, space stations or spacecrafts in general. In particular, the shielding device may be advantageously used with optical and/or electronic apparatuses, which operate in the frequency range of X-rays and/or γ-rays. The shielding device may be used for example also in fields, different from the specific space sector, for instance for aeronautical or terrestrial applications. It is useful to note that, in the end, in case of non-optical devices, the focal axis corresponds to another equivalent reference axis, which is characteristic for such devices. For example, for an antenna, the focal axis corresponds to the pointing axis of the antenna.
	claims	1. An automatic reloading and transport system for handling solid targets for a particle accelerator, comprising:a capsule configured to accommodate at least one solid target;a pneumatic tube transport system comprising at least two end stations each configured to dock and undock the capsule;a handling mechanism to open and close the capsule, and to move the target from or to the open capsule;a target positioning device configured to receive the target from the handling mechanism and move the target to a position for receiving a beam of accelerated particles from the particle accelerator, wherein upon irradiation of the target by the beam of accelerated particles, the target positioning device is configured to move the target back to a position for target manipulation by the handling mechanism. 2. The system of claim 1, wherein the particle accelerator is a cyclotron. 3. The system of claim 1, for use with solid targets that have an active face comprising a foiled, electroplated, vacuum evaporated or sputtered active material. 4. The system of claim 1, wherein the pneumatic tube transport system is configured to transport the solid target from a point of irradiation at the particle accelerator to a target processing point and from the target processing point back to the point of irradiation. 5. The system of claim 1, wherein the capsule comprises:a securing mechanism adapted to lock and unlock the capsule;a top piece adapted to be engaged by the securing mechanism; anda bottom piece adapted to hold the target, wherein the bottom piece comprises the securing mechanism. 6. The system of claim 5, wherein the securing mechanism of the capsule further comprises a spring loaded latch system, wherein when the top and bottom piece of the capsule are pushed together the spring loaded latch system is configured engage the top piece of the capsule and lock both pieces together, wherein the spring loaded latch system further comprises an unlocking feature configured to release the spring load of latches of the spring loaded latch system and unlock the capsule in response to the capsule being moved against the end bottom part of the pneumatic tube transport system. 7. The system of claim 5, wherein the handling mechanism includes a plurality of movable jaws configured to engage the top piece and the bottom piece of the capsule, wherein the jaws are configured to move whole capsule to engage unlocking feature of the capsule and to pull the top and bottom piece of the capsule away from each other to open the capsule or push the top and bottom piece of the capsule together to lock the capsule. 8. The system of claim 1, wherein the handling mechanism comprises a target manipulator that includes an arm and at least one suction cup located at the end of the arm, wherein the suction cup is adapted to apply a negative gas pressure to adhere to a surface of the solid target and enable manipulation of the target. 9. The system of claim 8, wherein the target manipulator is adapted to move rotationally about its axis to manipulate the solid target from or to the open capsule, wherein the target manipulator is further configured to move linearly up and down to enable the target removal from target holders. 10. The system of claim 9, wherein the target manipulator is adapted to manipulate the solid target from the open capsule directly to a target port of the solid target processing system and back. 11. The system of claim 8, wherein the target positioning device comprises:an input port having a first sealing system adapted to create a sealed connection between a beam port or a beam selector port of the particle accelerator and the target positioning device;a target holder configured to receive the solid target from the manipulator and pivot about an axis to place an active face of the solid target to be in position to receive the beam of accelerated particles; anda second sealing system configured to create a sealed interface between the active face of the solid target and the input port. 12. The system of claim 11, wherein the target holder is configured to pivot about the axis so that the active face of the solid target is placed perpendicularly to the beam of accelerated particles. 13. The system of claim 11, wherein the target holder is configured to position the active face of the solid target in front of the beam of accelerated particles under an angle, wherein the target holder is further configured to rotate the target when exposed to the beam of accelerated particles. 14. The system of claim 11, wherein the second sealing system comprises a compression seal feature located at the input port of the target positioning device and a mechanism adapted to move the pivoted target holder to engage and disengage the compression seal feature. 15. The system of claim 11, wherein the target holder further comprises a set of fingers configured to be mechanically controlled so that they hold the target within the target holder when the holder is being pivoted or the target positioning mechanism is in the close position.
	summary
053414071	summary	This invention relates to cladding for use in nuclear fuel elements. More particularly, the invention relates to cladding having a substrate, a zirconium barrier layer metallurgically bonded to the inside surface of the substrate, and an inner liner metallurgically bonded to the zirconium barrier. The inner liner has improved resistance to crack initiation and propagation. BACKGROUND OF THE INVENTION Nuclear reactors have their fuel contained in sealed cladding for the isolation of the nuclear fuel from the moderator/coolant system. The term cladding, as used herein, refers to a zirconium based alloy tube. Often the cladding will be composed of various layers including a zirconium alloy substrate and an unalloyed zirconium barrier. The cladding--nominally in the order of 0.030 inches thick--is formed in the shape of a tube with the nuclear fuel contained typically in pellet form therein. These pellets are stacked in contact with one another for almost the entire length of each cladding tube, which cladding tube is in the order of 160 inches in length. Typically, the cladding tube is provided with springs for maintaining the axial position of the fuel pellets and so-called "getters" for absorbing fission gases. Thereafter, the internal portions of the fuel rod are pressurized with helium to help conduct the heat from the fuel material to the cladding. Zirconium and its alloys, under normal circumstances, are excellent for nuclear fuel cladding since they have low neutron absorption cross sections and, at temperatures below about 350.degree. C., are strong, ductile, extremely stable and relatively nonreactive in the presence of demineralized water or steam. "Zircaloys" are a family of corrosion-resistant zirconium alloy cladding materials. They are composed of 98-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are two widely-used zirconium-based alloys for cladding. Zircaloy-2 has on a weight basis about 1.2 to 1.7 percent tin; 0.12 percent iron; 0.09 percent chromium and 0.05 percent nickel. Zircaloy-4 has essentially no nickel and about 0.2% iron but is otherwise substantially similar to Zircaloy-2. Splitting of Zircaloy cladding may occur due to the interactions between the nuclear fuel, the cladding, and the fission products produced during the nuclear reaction. It has been found that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. These localized stresses and strain in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking and liquid metal embrittlement. To combat this problem, some cladding includes barrier layers having low neutron absorption formed on the tubing inner surfaces. Cladding containing barrier layers is sometimes referred to as "composite" cladding. The barrier layer is typically a moderately pure zirconium (such as sponge zirconium) or sometimes highly pure zirconium (such as crystal bar zirconium) sheath metallurgically bonded to the inner surface of the tubing. The pioneering work on barrier layer cladding is described in U.S. Pat. Nos. 4,200,492 and 4,372,817 to Armijo and Coffin, 4,610,842 to Vannesjo, and 4,894,203 to Adamson. Barrier layers have been found to effectively prevent damage to the cladding due to interaction with the pellet. However, if the cladding wall is compromised in some manner (e.g. perforated or split), and water enters the fuel rod interior, the protection afforded by the barrier layer can be reduced. This is because the steam produced by water within the fuel rod can rapidly oxidize the barrier layer. The mechanical initiation of a cladding breach can be attributed to various causes. A breach can start when debris such as wires or metallic shavings or particles find their way into reactor water that flows within the fuel bundles between the fuel rods. The debris may lodge at a fuel rod spacer adjacent the cladding wall. As a result, the debris vibrates or frets against the cladding wall under the influence of the passing steam/water mixture. Such vibration continues until the cladding wall is penetrated. Corrosion also can be the source of crack initiation and propagation. Moreover, manufacturing defects can be the points of crack origin. Still further, crack propagation can start on the inside of the fuel rods in the corrosive high pressure environment present during in service reactor life. To protect the zirconium barrier from oxidation should a cladding breach occur, it has been proposed to use a three layer structure. In addition to the substrate and zirconium barrier, a corrosion resistant inner liner bonded to the fuel side of the barrier is employed. Typically, the inner layer will be made from a Zircaloy. If the cladding is breached and steam forms in the fuel rod interior, the inner liner will protect the barrier from rapid oxidation. Although this three layer design represents a significant advance, certain problems remain. For example, when exposed to fission products, Zircaloy inner liners sometimes serve as a site for crack initiation and propagation. If a crack in the inner liner becomes sufficiently deep (achieving a "critical length" or "critical depth"), it can propagate through the zirconium barrier and possibly through the entire cladding. It should be noted that the terms "critical length" and "critical depth" used herein refer to cracks in the radial direction of the liner wall. Thus, there exists a need for cladding having an inner liner which protects the barrier layer from oxidation and at the same time resists crack initiation and propagation at the cladding fuel side. SUMMARY OF THE INVENTION The present invention provides a cladding having an outer circumferential substrate, a zirconium barrier layer metallurgically bonded to the inside surface of the substrate and an inner circumferential liner metallurgically bonded to the zirconium barrier. The inner circumferential liner is more ductile than conventional Zircaloy and therefore has improved resistance to crack initiation and propagation due to pellet-cladding interaction. Preferably the inner circumferential liner is a zirconium alloy having a reduced tin and/or oxygen content in the liner alloy. In another aspect of the invention, the thickness of the inner liner is maintained below the critical depth for crack propagation. The critical depth is, as noted above, the length at which the crack in the liner can propagate through the zirconium barrier and possibly through the entire cladding. To avoid the possibility of cracks in the liner attaining the critical length, the inner liner thickness is preferably less than about 30 micrometers and more preferably less than about 20 micrometers. Further details and embodiments of the invention are provided in the following discussion and associated drawings.
050251635	abstract	An improved radiographic imaging screen is disclosed in which microscopic high-Z metal particles are coated with or dispersed in luminescent material and formed as a layer on a substrate. The complete surrounding of each particle with the luminescent material assures that secondary electrons emitted from the particles from absorption of incident radiation photons will encounter the luminescent material regardless of the direction of emission of the electrons.
	abstract	An auxiliary illuminating device that has an adjustable color temperature. The color temperature is adjusted by varying the light output at least two independently adjustable light sources. The light source is an array of at least 2 colors. The light source typically uses at least one set of LED""s.
	abstract	A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360° and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.
	summary
	summary
054405980	description	BEST MODE FOR CARRYING OUT THE INVENTION Reference will now be made in detail to a present preferred embodiment of the invention, an example of which is illustrated in the accompanying drawings. Referring now to FIG. 1, there is illustrated a core C of a typical boiling water nuclear reactor, including four side-by-side fuel bundles B disposed in a square array, the four bundles being illustrated at 12, 14, 16 and 18. The fuel bundles are supported in a manner well known to those skilled in this art and description of their support is not believed necessary. As illustrated, each fuel bundle includes a plurality of discrete fuel rods 20 containing nuclear fuel pellets and a water rod 22, whereby the reactor is supplied with fissionable material. Suffice to say that the self-sustaining fission reaction produces fission products, the kinetic energy of which is dissipated as heat in the fuel rods with the heat being removed by coolant water surrounding the rods and bundles boiling into steam from which useful work is extracted. Referring to FIGS. 2a-2c of the prior art, FIG. 2 illustrates a typical fuel bundle design using enriched uranium as fuel. As illustrated in FIG. 2a, each fuel rod 24 in the lattice of rods is provided a number indicating its fuel loading. In FIG. 2b, the enriched uranium and burnable poison, e.g. gadolinium, loadings of each rod are illustrated. For example, the fuel rod denoted 3 in FIG. 2b indicates a fuel rod containing 3.6% enriched uranium, the balance of the rod being formed substantially of uranium. Also, all rods designated 3 in the bundle B of FIG. 2a have these constituents. The rod numbered 7, however, contains 3.95% enriched uranium and 5.00% burnable poison, e.g., gadolinium. As noted in FIG. 2a, the rods numbered 7 having the combined enriched uranium and gadolinium concentration are few in number and are arranged geometrically symmetrically within the interior of the fuel bundle. Thus, only eight of the fuel rods of the sixty rods illustrated in a typical fuel bundle are loaded with the burnable poison gadolinium. In FIG. 2c, there is illustrated a typical burnup reactivity curve showing the reactivity (K-.infin.) as a function of fuel exposure (burnup). Burnup is a measure of the energy produced by the fuel during its useful lifetime. Note that the fuel reactivity rises sharply from a value near 1 at the beginning of its life, to a value of about 1.15 at a predetermined time corresponding to about 8 GWd/MT on the part of the curve designated a. This initial reactivity then declines in a nearly linear fashion to the end of its useful lifetime as indicated on the part of the curve designated b. It is important to maintain this characteristic curve for each fuel bundle as it permits fresh fuel bundles with increasing reactivity to offset the declining reactivity of the older fuel bundles. The reactivity balance enables the reactor to operate for extended periods at a relatively stable steady rate. In FIGS. 3a-3c, there is illustrated a fuel bundle incorporating plutonium as fuel. Similarly as in the preceding drawing figures, enriched uranium, plutonia and gadolinia concentrations are shown for each rod. For example, rod numbered 3 includes 0.2% enriched uranium and 3.00% plutonium as the fissile material, the balance of the rod comprising substantially uranium. Gadolinia is not provided the rods numbered 3 in the lattice of FIG. 3a. Rod numbered 6, as illustrated in FIG. 3b, contains 0.20% enriched uranium, 5.00% plutonium and 3.00% gadolinia. As illustrated in FIG. 3a, the gadolinium-loaded fuel rods numbered 6 are few in number and lie within the fuel rod lattice, eight such rods being shown lying symmetrically within the interior of the fuel rod bundle B. A characteristic reactivity curve for these fuel rods is illustrated in FIG. 3c. Note that the reactivity curve is considerably different in shape, including slope, than the reactivity curve for the enriched uranium fuel rods of FIG. 2a. Turning now to FIGS. 4a-4c, illustrating a fuel bundle design according to the present invention, it will be appreciated that the same number of control rods in an 8.times.8 array as in the prior two fuel bundles is illustrated. FIG. 4b, as in the corresponding figures of the prior art, illustrates the constituents of the fuel rods. For example, fuel rods numbered 5, 6 and 7 have a gadolinia concentration of 1.00% in combination with various percentages of enriched uranium and plutonium. Rods numbered 1, 2, 3 and 4 are void of gadolinia. As illustrated in FIG. 4a, the gadolinia-loaded rods are arranged in an interior array and are wholly surrounded by an exterior array of fuel rods numbered 1, 2, 3 and 4 void of gadolinium. In FIG. 4c, there is illustrated a burnup reactivity curve for the fuel bundle of FIG. 4a having the constituents identified in FIG. 4b. Note the rise in the fuel reactivity from startup to a peak indicated at a' and the nearly linear decline from the peak indicated at b'. In accordance with the present invention, it has been found that the number of fuel rods containing the burnable poison, e.g. gadolinium, when combined with plutonium in an interior array of fuel rods, should be in excess of 20% of the total number of fuel rods in the fuel bundle in order to produce the characteristic reactivity curve illustrated in FIG. 4c. Note the substantial similarity in shape including slope between the reactivity curve of FIG. 4c and the reactivity curve of FIG. 2c, as well as the substantial dissimilarity between the reactivity curve of FIG. 4c and that of FIG. 3c. As indicated previously, it is important that the fuel bundle of the present invention, which results in enhanced use of plutonium as the fuel as in FIG. 4a, has a reactivity curve substantially corresponding in shape including slope to the reactivity curve of the bundles employing enriched uranium and gadolinium, as illustrated in FIGS. 2a-2c. These two reactivity curves of FIGS. 2c and 4c also substantially correspond in reactivity values. Thus, in FIG. 4 a, each rod of the interior array of 32 fuel rods has a combination of enriched uranium, plutonia and gadolinia, hence enhancing plutonium usage, while the exterior array of 28 fuel rods is void of the burnable poison gadolinia and has a combination of only plutonium and enriched uranium. Lesser number of interior fuel rods containing the plutonium, enriched uranium and gadolinium can be used with the reactivity curve remaining substantially as illustrated. Consequently, the number of fuel rods containing the gadolinium lying in the interior array of fuel rods in the bundle should number in a range of 20% to 60% of the total number of rods in the fuel bundle. Also, each rod in the fuel rod bundle preferably has a percentage concentration of plutonium as one of its constituents and in excess of the percentage constituent of any other fissile materials in the rods. In this manner, the reactor fuel contains enhanced quantities of plutonium as compared with the quantities of plutonium previously thought possible as part of the fuel for reactors of this type. By substituting a fuel bundle of the type illustrated in FIGS. 4a and 4b for a conventional fuel bundle of the type illustrated in FIGS. 2a and 2b in a nuclear reactor core, enhanced plutonium usage is obtained. This is made possible because of the substantial correspondence of the reactivity curves of these two different types of fuel bundles as illustrated by a comparison of FIGS. 2c and 4c. By substituting over time the bundles of FIGS. 4a and 4b for those of FIGS. 2a and 2c in the core, the nuclear reactor core may be operated and controlled similarly as if employing fuel bundles of the type illustrated in FIGS. 2a and 2b. While the invention has been described with respect to what is presently regarded as the most practical embodiments thereof, it will be understood by those of ordinary skill in the art that various alterations and modifications may be made which nevertheless remain within the scope of the invention as defined by the claims which follow.
	description	The present invention concerns a method of producing a cladding tube for nuclear fuel for a nuclear boiling water reactor, which method comprises the following steps: forming a tube which comprises an outer cylindrical component mainly containing zirconium and an inner cylindrical component metallurgically bonded to the outer component, wherein also the inner component at least mainly contains zirconium, wherein the material compositions of the inner component and the outer component are selected such that they differ from each other and such that the inner component has a lower recrystallization temperature than the outer component. The invention also concerns a cladding tube, a use of a cladding tube as well as a fuel assembly for a nuclear boiling water reactor comprising such a cladding tube. A method of the kind that is described in the first paragraph above is known from the patent document EP 0 674 800 B1. In this document also the background to the invention described therein is described. When a cladding tube is used in a nuclear reactor it contains nuclear fuel, usually in the form of pellets containing enriched UO2. The cladding tube with its content thus constitutes a fuel rod. Because of the very particular environment in which cladding tubes are used, different requirements must be fulfilled. There are mainly two kinds of modern light water reactors: boiling water reactors (BWR) and pressure water reactors (PWR). In these kinds of reactors different conditions exist, which call for different requirements on the parts that are included in the reactors. In a PWR, the fuel rods are cooled mainly by water that is in a liquid phase under a high pressure. In a BWR, the pressure is lower and the water that cools the fuel rods is evaporated such that the fuel rods are surrounded both by water in a liquid phase and in a steam phase. Furthermore, the fuel assemblies have different construction in a BWR and a PWR. In a certain kind of BWR, the fuel rods in a fuel assembly extend the whole way between a top plate and a bottom plate which keep the fuel assembly together. In a PWR, on the other hand, the fuel rods are usually kept in position with the help of spacers and do not reach all the way to the top plate and to the bottom plate. When a fuel rod is used in a nuclear reactor, it is exposed to neutron radiation. This leads to the fact that the cladding tube tends to grow with time. In certain kinds of BWR, the cladding tube has only a limited possibility to expand in the longitudinal direction. The cladding tube may therefore bend during operation. This can lead to damages. It should therefore be avoided that the cladding tube grows to a larger extent. Modern cladding tubes which are produced in suitable zirconium alloys and which undergo special heat treatments during the production often have a relatively low tendency to grow when they are exposed to neutron radiation. The tendency to grow may be reduced, inter alia, in that the cladding tube during the production undergoes a final recrystallization anneal. Through a suitable choice of the material for the cladding tube and a suitable method of production, the cladding tube can obtain suitable properties concerning for example hardness and ductility. Since the conditions are different in a BWR and a PWR, the cladding tubes are produced with different properties depending on for which kind of reactor they are made. In the environment where the cladding tubes are used they are subject to different corrosive attacks. These attacks may come from the outside or from the inside. The attacks from the inside often have their basis in an influence from the nuclear fuel material that is located there, so-called pellet-cladding interaction (PCI). If a crack is formed through the cladding tube (a so-called primary damage), water may penetrate in through the crack and spread along the inside of the tube. This may lead to new corrosive attacks from the inside of the tube, so-called secondary damages. A cladding tube of zirconium may also react with hydrogen such that hydrides are formed in the cladding tube. These hydrides may be formed from the inside of the tube, particularly if a crack has been formed such that water has penetrated into the tube. These hydrides make the tube more fragile and the probability for the formation of cracks increases. Particularly hydrides that extend in a radial direction through the tube constitute an increased risk for crack formation. Such radial hydrides may therefore speed up possible secondary damages and crack formations. The complicated chemical, mechanical and metallurgical conditions that are the case in a nuclear reactor have lead to the fact that a very large number of suggestions have been proposed for the selection of materials and for the methods of production of cladding tubes. Even small changes in the composition of alloys or production parameters may have a large importance for the properties of the cladding tube. Since different conditions are the case on the inside and on the outside of the cladding tube, cladding tubes are sometimes produced with different compositions in different layers. The above mentioned document EP 0 674 800 B1 thus describes the production of a cladding tube which has an outer component that is made of for example any of the frequently occurring alloys Zircaloy 2 and Zircaloy 4. The cladding tube has an inner component—a so-called liner—which according to an embodiment mainly consists of Zr with the alloying elements 0.25% Sn, 310 ppm Fe and 430 ppm O. The cladding tube is produced according to a particular method with carefully selected heat treatments. The cladding tube undergoes a final anneal at 570° C. during 1.5 h, which means a complete re-crystallization anneal (cRXA). The produced cladding tube has been shown to have a good resistance against corrosion even if water happens to penetrate into the inside of the cladding tube. Another example of a cladding tube is clear from U.S. Pat. No. 4,933,136. This document describes a cladding tube consisting of an outer component of Zircaloy 2 or Zircaloy 4 and an inner component which according to one embodiment mainly consists of Zr with 0.19-0.20 percentage by weight Sn, 0.19 percentage by weight Fe and 615-721 ppm O. The document describes the production of the tube with different rolling steps and heat treatments. As a final anneal three alternatives are described in the document. According to the first alternative, a complete recrystallization (cRXA) occurs in both the outer and the inner component. According to a second alternative, a cRXA occurs in the inner component but only a stress relief anneal (SRA), i.e. no noticeable recrystallization, in the outer component. According to a third alternative, a partial recrystallization (pRXA) occurs in the inner component and an SRA in the outer component. For cladding tubes which are constructed with two layers and which are intended to be used in a BWR, usually a final anneal is carried out which leads to a cRXA in both the layers. Thereby, a good resistance against damages caused by PCI can be achieved at the same time as the cladding tube has a good ductility and also obtains a structure that counteracts growth caused by neutron radiation. An object of the present invention is to accomplish a method of producing a cladding tube for nuclear fuel for a nuclear boiling water reactor, which cladding tube has a good resistance against damages caused by PCI at the same time as the risk for the formation of radial hydrides is low. A further object is to achieve these advantages at the same time as the tendency to growth caused by neutron radiation is kept at a low level. Further objects and advantages of the invention will be clear from the following. These objects are achieved with a method of the kind that has been described in the first paragraph above and which furthermore is characterised in that after that the cladding tube has been formed according to the above, and after possible rolling steps with there between occurring heat treatments, the cladding tube is final annealed at a temperature and during a time such that the inner component substantially completely recrystallizes and such that the outer component partly recrystallizes but to a lower extent than the inner component. Since the inner component is substantially completely recrystallized (cRXA), the tube has a very good resistance against PCI damages. Since the outer component is partially recrystallized (pRXA), this component is relatively ductile at the same time as it does not grow to a too high extent when it is exposed to neutron radiation. By a suitable choice of material it has become clear that the growth of the cladding tube is so low that it is very well suited to be used also in the kind of BWR where the cladding tube only has a limited space for growth. Since the outer component only is pRXA, it has become clear that possible hydrides which are formed tend to extend in essentially a tangential direction while the risk for radial hydrides is low. Thereby an improved resistance against crack formation is obtained. The reason why radial hydrides are avoided is probably that certain tensions which originate from the production of the tube are maintained since the outer layer is not cRXA. These tensions have as a consequence that the tendency for radial hydrides is reduced. By “substantially completely recrystallized” is here meant that the recrystallization is 100% (completely recrystallized) or almost 100% (at least recrystallized to 97% or 98%). An analysis of the cladding tube may thus show that the recrystallization in the inner component is not totally complete. It is preferred that the inner component is completely recrystallized. It should be noted that the final anneal normally is the last heat treatment step in the production method. Possibly a certain after treatment of the cladding tube may be carried out, but such an after treatment should be such that the structure that is obtained through the final anneal is not substantially destroyed. It should also be noted that according to a preferred embodiment, the cladding tube consists only of the outer component and the inner component (the liner). There are thus no further layers. The composition on the outer surface and the inner surface of the tube may however differ from the composition in the inner of the layers, for example due to the substances that the tube has come into contact with. The tube may for example be oxidised through the fact that it has been kept in an environment of air. According to an alternative embodiment, it is however feasible that the tube comprises one or more further layers in addition to the outer component and to the inner component. Finally it is pointed out that when in this document % or ppm are used in connection with contents of different substances it is, if nothing else is said, referred to percentage by weight of the respective substances. According to a preferred manner of carrying out the method according to the invention, the final anneal is carried out such that the degree of recrystallization in the outer component is higher than 50%. Suitably, the degree of recrystallization in the inner component is substantially or completely 100% and the degree of recrystallization in the outer component is suitable between 50% and 96%, particularly suitable 60% to 90%, for example between 70% and 90%. It has become clear that such degrees of recrystallization are particularly suitable for achieving the described advantages when the cladding tube is used in a BWR. Lower degree of recrystallization than 50% is possible, but this tends to lead to the fact that the growth of the cladding tube when it is exposed to neutron radiation is larger. According to another preferred embodiment, the inner component does not contain more than 2000 ppm Fe and preferably not more than 1500 ppm Fe and most preferred less than 1000 ppm Fe. According to another preferred embodiment, the inner component does not contain more than 1000 ppm O. By keeping the contents of Fe and O low, a good resistance-against PCI is obtained. It should be noted that the inner component may be produced in pure Zr (except for possible impurities) and thus not necessarily need to be an alloy. According to a preferred embodiment, the outer component has a composition which is completely or substantially according to Zircaloy 2 or Zircaloy 4. These materials are common in connection with cladding tubes and have been shown to have many good properties. It should however be noted that the outer component does not need to be Zircaloy 2 or Zircaloy 4. Also other alloys may be used. For example different Zr-based alloys which contain Nb. According to preferred embodiment, the inner component contains between 0.1 and 0.7 percentage by weight Sn, preferably between 0.1 to 0.4 percentage by weight Sn, 400 to 1500 ppm Fe, less than 600 ppm O (for example 300 ppm to 500 ppm O) and the rest Zr, except for impurities of a content that does not exceed that which is normally accepted in Zr or Zr-alloys for applications in nuclear reactors. Such an alloy has been shown to have very good properties at the same time as it has a suitable recrystallization temperature in order to be able to obtain substantially cRXA in the inner component at the same time as pRXA is obtained in the outer component. Examples of what is considered as acceptable impurities in this context is described for example in the above mentioned document EP 0 674 800 B 1, column 6. For example, impurities in Zr or Zr-alloys shall be below the limits that normally apply to reactor-grade zirconium, namely, Al 75 ppm, B 0.5 ppm, C 100 ppm, Ca 30 ppm, Cd 0.5 ppm, Cl 20 ppm, Co 20 ppm, Cu 50 ppm, H 25 ppm, Hf 100 ppm, Mg 20 ppm, Mn 50 ppm, Mo 50 ppm, N 65 ppm, Na 20 ppm, Nb 100 ppm, Ni 70 ppm, P 30 ppm, Pb 100 ppm, Si 100 ppm, Ta 200 ppm, Ti 50 ppm, U 3.5 ppm, V 50 ppm, W 100 ppm, and Cr 200 ppm. Suitably, the inner component has a thickness such that it constitutes between 3% and 30%, preferably between 5% and 20% and most preferred 10% of the total thickness of the cladding tube. According to a suitable embodiment, the final anneal is carried out at a temperature of between 485° C. and 550° C. during 1 h to 6 h, preferably during 2 h to 4 h. As has been mentioned initially, the invention also concerns a use. Thereby a cladding tube produced according to the method according to any of the preceding embodiments is used in a fuel assembly for a nuclear boiling water reactor. Thereby, the above described advantages with such a cladding tube are achieved. The invention also concerns a cladding tube as such, suitable to contain nuclear fuel and to be used in a nuclear boiling water reactor. This cladding tube comprises: an outer cylindrical component mainly containing zirconium, an inner cylindrical component which at least mainly contains zirconium and which is metallurgically bonded to the outer component, wherein the material compositions of the inner component and the outer component differ from each other and are such that the inner component has a lower recrystallization temperature than the outer component. The inner component has a substantially completely recrystallized structure and the outer component has a structure such that it is partly recrystallized but not to the same extent as the inner component. Such a cladding tube can be produced according to the above described method. With this cladding tube, the above described advantages are achieved. Advantageous embodiments of this cladding tube are clear from the dependent claims below. With these embodiments, the above described advantages are achieved. Finally, the invention also concerns a fuel assembly for a nuclear boiling water reactor. This fuel assembly comprises: an enclosing tube, and a plurality of cladding tubes according to the invention filled with nuclear fuel suitable for such cladding tubes for a boiling water reactor, wherein said plurality of cladding tubes are arranged inside said enclosing tube. FIG. 1 shows schematically a fuel assembly, known per se, for a BWR. The fuel assembly comprises an enclosing tube 2 (which here only is shown to the right in the figure). Within the enclosing tube 2, a number of fuel rods 3 are arranged. The fuel rods 3 extend from a top plate 5 to a bottom plate 6. The fuel rods 3 consist of cladding tubes which contain pellets with nuclear fuel material. In the figure, a number of pellets 4 are symbolically shown. At the top, the fuel rods 3 are provided with end plugs 8. The fuel rods 3 abut against the lower side of the top plate with the help of coiled springs 9. A plurality of spacers 7 are arranged in order to keep the fuel rods 3 at a distance from each other. When a fuel assembly of for example the described kind is provided with a plurality of cladding tubes according to the present invention, it thus constitutes a fuel assembly according to the invention. It should be noted that there are different kinds of fuel assemblies for BWR. For example, there are fuel assemblies for BWR without a top plate. Furthermore, fuel assemblies for BWR often also comprise so-called part length rods. The present invention is of course applicable to different kinds of fuel assemblies for BWR. FIG. 2 shows schematically a cross-section through a cladding tube according to the invention. The cross section shows the cladding tube strongly enlarged. In reality, the cladding tube has a dimension and a length that are suitable for use in a BWR. The cladding tube comprises an outer cylindrical component 10 and an inner cylindrical component 20. The inner component 20 may be called a liner. Both the outer 10 and the inner 20 components contain mainly Zr. The inner component 20 is metallurgically bonded to the outer component 10. The material compositions of the inner 20 and the outer 10 components differ from each other and are such that the inner component 20 has a lower recrystallization temperature than the outer component 10. The inner component 20 has a completely recrystallized structure or at least a substantially completely recrystallized structure. The outer component 10 has a structure such that it is partly recrystallized but not to the same high extent as the inner component 20. The degree of recrystallization in the outer component 10 is suitably between 50% and 96%, preferably between 70% and 90%. The outer component 10 may consist of Zircaloy 2 or Zircaloy 4 or other suitable alloy based on Zr. The inner component 20 may consist of pure Zr or a Zr-alloy, which thereby suitably is a low-alloy such that it has a lower recrystallization temperature than the outer component 10. The inner component 20 may consist of 0.1 to 0.4 percentage by weight Sn, 400 to 1500 ppm Fe, less than 600 ppm O and the rest Zr, except for impurities of a content that does not exceed that which is normally accepted in Zr or Zr-alloys for applications in nuclear reactors. The thickness of the inner component 20 may for example be 10% of the total thickness of the cladding tube. The invention also concerns a method of producing a cladding tube for nuclear fuel for a nuclear boiling water reactor. The method according to the invention may be carried out in the following manner. A tube is formed that comprises an outer cylindrical component 10, which for example may consist of Zircaloy 2, and an inner cylindrical component 20 which is metallurgically bonded to the outer component 10. The inner component 20 is also based on Zr and has a material composition such that it has a lower recrystallization temperature than the outer component 10. The inner component 20 may contain for example 0.25 percentage by weight Sn, about 500 ppm Fe, less than 600 ppm O and the rest Zr except for impurities of a content that does not exceed that which is normally accepted in Zr or Zr-alloys for applications in nuclear reactors. This tube may be formed in different manners, for example such as have been described in EP 0 674 800 B1. When the two components 10, 20 have been joined together, suitably a number of rolling steps with intermediate heat treatments are carried out. After that the cladding tube has been formed according to the above and after possible rolling steps with there between occurring heat treatments, the cladding tube is final annealed at a temperature and during a time such that inner component 20 substantially completely recrystallizes in such that the outer component 10 partly recrystallizes but to a lower extent than the inner component 20. The final anneal is suitably carried out such that the degree of recrystallization in the outer component 10 is higher than 50% but less than 96%. A suitable time and temperature for the final anneal depend on the composition of the alloys. The temperature and time should therefore be selected such that the desired degrees of recrystallization are achieved in the components. In for example Zircaloy 2, the Sn-content may vary between 1.2 and 1.7 percentage by weight. If the outer component for example contains 1.3 percentage by weight Sn, then a final anneal at a temperature of between 485° C. and 515° C. during 3 h has been shown to lead to a good result. If the Sn content in the outer component is 1.5 percentage by weight, the final anneal is suitably carried out at between 505° C. and 520° C. during 3 h. The inner component 20 may for example have a thickness such that it constitutes about 10% of the total thickness of the cladding tube. A cladding tube produced according to the method may suitably be used in a fuel assembly in a nuclear BWR. The above described cladding tube and method only give examples of suitable materials. As has been pointed out above, other materials may be considered, for example a Zr—Nb alloy for the outer component and possibly pure Zr for the inner component. The invention is not limited to the above described examples but may be varied within the scope of the following claims.
	abstract	A method is provided for creating a plurality of substantially uniform nano-scale features in a substantially parallel manner in which an array of micro-lenses is positioned on a surface of a substrate, where each micro-lens includes a hole such that the bottom of the hole corresponds to a portion of the surface of the substrate. A flux of charged particles, e.g., a beam of positive ions of a selected element, is applied to the micro-lens array. The flux of charged particles is focused at selected focal points on the substrate surface at the bottoms of the holes of the micro-lens array. The substrate is tilted at one or more selected angles to displace the locations of the focal points across the substrate surface. By depositing material or etching the surface of the substrate, several substantially uniform nanometer sized features may be rapidly created in each hole on the surface of the substrate in a substantially parallel manner.
043702961	abstract	A fusion reactor of the toroidal-type having a plasma containing toroidal fusion region producing energy from fusion reactions and comprising a toroidal field generating means for producing a toroidal magnetic field in the fusion region upon passage of current therethrough, said toroidal field generating means positioned proximate the toroidal fusion region, and ohmic heating coils for ohmically heating the plasma wherein the ohmic heating coils are positioned between the toroidal fusion region and the toroidal field generating means.

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