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claims | 1. An electron-beam metrology system comprising:a specimen stage to mount a specimen on which a device pattern is formed;electron optics to radiate the device pattern with an electron-beam;a secondary electron detector to detect a secondary electron generated by the radiation of the electron-beam; andan information processing system to analyze a signal obtained from the secondary electron detector;wherein a standard reference for metrology is held on the specimen stage; andwherein the standard reference comprises:a first grating unit pattern including an array of gratings having pitch sizes which are verified by an optical method;a second grating unit pattern including an array of gratings having pitch sizes which are smaller than the pitch sizes of the first grating unit pattern; anda positioning mark. 2. An electron-beam metrology system according to claim 1, wherein parallelism of a moving direction of the specimen stage is corrected on the basis of the positioning mark. 3. An electron-beam metrology system according to claim 1, wherein the positioning mark is used to search for the first grating pattern. 4. An electron-beam metrology system according to claim 1, wherein a measurement result of the device pattern is calibrated utilizing information obtained from the standard reference. 5. An electron-beam metrology system according to claim 4, further comprising a display on which the calibrated measurement result is displayed. 6. An electron-beam metrology system according to claim 1, further comprising:storage means which stores a size standard value for metrology;wherein the size standard value is an average value of plural pitch sizes of the second grating unit obtained by plural measurements of the second grating unit. 7. An electron-beam metrology system according to claim 6, wherein the information processing system confirms whether the stored size standard value satisfies a predetermined correctness or not by comparing a measurement result of a pitch size of the first grating unit pattern and the verified pitch size by the optical method. 8. An electron-beam metrology system according to claim 7, wherein a reference value for the confirmation is stored in the storage means. 9. An electron-beam metrology system according to claim 7, further comprising a display which displays an alarm when the size standard value does not satisfy the predetermined correctness. 10. An electron-beam metrology system according to claim 1, wherein the information processing system measures a size of the device pattern using the pitch size of the second grating pattern as the size standard, and verifies the measurement result by comparing a measurement result of a pitch size of the first grating unit pattern and the verified pitch size by the optical method. 11. An electron-beam metrology system according to claim 10, further comprising a display which displays an alarm when the measurement result is not verified. 12. An electron-beam metrology system comprising:a specimen stage to mount a specimen on which a device pattern is formed;electron optics to radiate the device pattern with an electron-beam;a secondary electron detector to detect a secondary electron generated by the radiation of the electron-beam;an information processing system to analyze a signal obtained from the secondary electron detector; andstoring means which stores a size standard value for measuring a size of the deice pattern;wherein the size standard value stored in the storing means is obtained by using a standard reference comprising:a first grating unit pattern including an array of gratings having pitch sizes which are verified by an optical method; anda second grating unit pattern including an array of gratings having pitch sizes which are smaller than the pitch sizes of the first grating unit pattern. 13. An electron-beam metrology system according to claim 12, the size standard is an average value of plural pitch sizes of the second grating unit obtained by a plural measurements of the second grating unit. 14. An electron-beam metrology system according to claim 13, the size standard is verified by comparing a measurement result of a pitch size of the first grating unit pattern and the verified pitch size by the optical method. |
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039873064 | summary | BACKGROUND OF THE INVENTION This invention relates to the production of ultra-violet (U.V.) radiation, and relates particularly, but not exclusively, to a method of, and an apparatus for, treatment of a material by irradiation with U.V. radiation. It is known to produce U.V. radiation by the discharge of electrical energy through a gas, and a device operating on such a principle is called a gas discharge lamp. SUMMARY According to one aspect of this invention there is provided a method of producing U.V. radiation comprising the steps of effecting, in a discharge chamber, an electrical discharge between two electrodes, and arranging for at least a part of the electrical discharge between the two electrodes to be surrounded by a preselected liquid. The pressence of the surrounding liquid prevents the hot plasma of the discharge region between the electrodes from expanding freely, in contrast to the operation of the conventional gas discharge lamp, and as a consequence the plasma can attain a high temperature and pressure with the resultant emission of U.V. radiation. It will be appreciated that the preselected liquid must have appropriate electrical characteristics to permit the electrical discharge therethrough, and also that the liquid molecules when ionised by the energy of the electrical discharge must emit radiation of a desired wavelength at which the preselected liquid must be substantially transparent. For convenience this discharge will be referred to as a main discharge. Where one of the electrodes is formed by a preselected conductive liquid immiscible with said preselected liquid, the method may include the step of generating an initiating electrical discharge in the vicinity of the discharge region between said two electrodes just prior to effecting said electrical discharge. This initiating discharge may be generated between a third electrode and either of said two electrodes, and is for the purpose of providing in the discharge region of the main discharge a mixture of the molecules, or even ions, of said preselected liquid and of the liquid electrode. According to another aspect of this invention there is provided a method of treating a material by irradiation with U.V. radiation, comprising the steps of producing U.V. radiation by a method as described above, transmitting U.V. radiation thus produced through a U.V. radiation-transparent portion of the chamber and disposing the material in a treatment region where it is irradiated by the U.V. radiation transmitted from the chamber. The material may be treated in batches, and thus may be stationary in the treatment region, or may be treated in a continuous process by transporting the material continuously through the treatment region and continuously effecting electrical discharge at a rate appropriate to the speed of the material to obtain a required treatment. A required treatment may be each elemental portion of material receiving U.V. radiation from two discharges, or three discharges, and a minimum treatment may be each elemental portion being irradiated once and once only. The material may be solid, e.g. bandage to be sterilized, or liquid. In the latter case the treatment region will be constituted by a chamber through which the liquid flows. The discharge chamber may conveniently be a sealed chamber, and this avoids splashing-out of drops of the preselected liquid ejected from the discharge region. Where it is not required for the chamber to be sealed, it may have an open top portion which constitutes the or a part of the U.V. transparent portion of the chamber. Where the material to be treated is a liquid, the transporting step preferably comprises feeding the liquid to be treated through a treatment chamber which constitutes the treatment region. The treatment chamber may be outside or inside the discharge chamber and have inlet and outlet passages for the flow therethrough of the liquid to be treated. By the term `inside` is meant that the discharge chamber encircles the treatment chamber, and by the term `ouside` is meant that the discharge chamber does not encircle the treatment chamber. In the latter case it will be understood that the treatment chamber may encircle the discharge chamber. It will be appreciated that the thickness or depth of the treatment chamber will depend on the degree of absorption of U.V. radiation by the liquid to be treated and will be such that the extreme regions of this liquid will still receive an adequate amount of U.V. radiation appropriate to the particular treatment. There may be included the step of automatically adjusting the length of the part of the electrical discharge which is surrounded by the preselected liquid such as to maintain predetermined electrical characteristics of the electrical discharge. Where the whole of the electrical discharge is so surrounded, the spacing between the electrodes will be altered, either or both of the electrodes being moved. In one form of this invention the main discharge occurs from the tip of an electrode submerged in said predetermined liquid, through the liquid direct to the surface thereof, and then along the surface to the wall of the chamber which is conductive and forms the other electrode for the discharge. In the operation of such a form, the length of the submerged part of the discharge path can be altered by causing the preselected liquid to have a circulating motion which results in a raising of the level of the peripheral surface regions and a depression of the central surface region. The submerged electrode is preferably arranged directly under the region where greatest depression occurs to obtain maximum sensitivity i.e. the maximum change in submerged length of the discharge path for a given circulatory speed of the preselected liquid. According to a further aspect of this invention there is provided an apparatus for producing U.V. radiation comprising two electrodes spaced apart to form a discharge region therebetween, a discharge chamber containing the discharge region, first means arranged to apply high voltage between the electrodes, and means arranged to supply a preselected liquid to the discharge region such that, on application of the high voltage an electrical discharge occurs between the electrodes, at least a part of this discharge being along a submerged path in the preselected liquid, and at least a portion of the discharge chamber being transparent to the U.V. radiation whereby U.V. radiation produced may be transmitted from the discharge chamber. One of the electrodes may be in the form of a rod mounted in the wall of the chamber, and the other electrode may be in one of several alternative forms. Firstly the other electrode may be in the form of a rod similar to the said one electrode and similarly mounted in the wall of the chamber; secondly the other electrode may be formed by the chamber itself provided that it is formed of a conductive material, or if not, then formed with a conductive coating on its inner surface; thirdly the other electrode may be in the form of a layer of a preselected conductive liquid, preferably mercury, which is immiscible with said preselected liquid and which is disposed at the bottom of the discharge chamber, it will be appreciated that in this last case said one electrode would be disposed in the chamber above the surface of the conductive liquid. In this specification the term rod includes an electrode having a channel for liquid flow. The channel may be in the form of an axial bore or a hollow cylindrical annulus. One preferred form of discharge chamber is formed as a hollow cylinder having two end portions having flat inner surfaces and a hollow cylindrical intermediate portion having a cylindrical inner surface. In this case the two electrodes may be in the form of rod electrodes, each axially mounted in a respective end portion such that the discharge region between the electrodes is approximately at the centre of the chamber. Either an end plate forming an end portion of the chamber, or a tube forming the intermediate portion of the chamber, or both of these, may be formed of a material, preferably quartz, transparent to U.V. radiation. An apparatus for producing U.V. radiation may be used in the treatment of materials by irradiation with U.V. radiation. Where such a material is a liquid, a treatment chamber having inlet and outlet passages for the flow therethrough of a liquid to be treated may be provided either externally or internally of the discharge chamber. A treatment chamber internally of the discharge chamber may be formed by a hollow tube of U.V. transparent material, preferably quartz, passing through the discharge chamber. In this case it will be appreciated that the U.V. transparent wall is common to the treatment chamber and the discharge chamber and constitutes the aforementioned portion transparent to U.V. radiation. Also, the end walls and the intermediate portion of the chamber need not be U.V. transparent unless specifically required. The treatment chamber may be constituted by a plurality of the hollow tubes which may be symmetrically arranged around the axis of discharge chamber. The electrodes may be mounted in the end portions or in the intermediate portion. The inner surfaces of the chamber may be made highly reflective by mechanical treatment, e.g. polishing, to make maximum use of the generated U.V. radiation. If required, the intermediate portion may have an elliptical cross-sectional shape as alternative to the cylindrical shape. The electrodes may be arranged such that the electrical discharge is at one of the foci, and the treatment chamber, e.g. one or more hollow tubes, would then be arranged at the other focus. A form of treatment chamber which can be utilised with a U.V. transparent cylindrical intermediate portion is one which is a cylindrical annulus formed between the outer cylindrical surface of the intermediate portion, and the inner cylindrical surface of a sleeve mounted between the end portions. The mounting of the end portions to the intermediate portion and to the sleeve will be such as to seal the discharge chamber from the treatment chamber. One end portion of the discharge chamber may be made U.V. transparent and means may be provided for producing a thin film flow of liquid to be treated across the outer surface of this end portion. One form of such means for producing a thin film flow may be a plate spaced from the end portion such as to form a thin disc-like chamber. The inlet and outlet passages are arranged such that the thin film flow is substantially radial within the treatment chamber. Preferably the inlet passage is aligned with the axis of the chamber, and the outlet passage communicates with the peripheral regions of the treatment chamber via an annular collecting chamber therearound. In one form of an apparatus the axis of the intermediate portion is arranged vertically and an electrode is mounted axially in the bottom end portion, being insulated therefrom. Liquid inlet and outlet passages are arranged to provide a constant head of liquid in the chamber such that the top of the electrode is submerged in the liquid. The inlet passage is arranged such that a rotational movement is imparted to the liquid thereby to cause the level of liquid above the electrode tip to decrease. The electricl discharge is from the electrode tip to the surface of the liquid, and then along the surface to the inner surface of the intermediate portion which is conductive or has a conductive coating thereon and acts as an electrode. Means may be provided for controlling the rate of flow of liquid into the discharge chamber in dependence upon the electrical parameters of the discharge in order to maintain automatically constant discharge conditions. The inner surface of the discharge chamber may in such an apparatus be shaped to enhance the irradiation of a treatment chamber at the top of the chamber, and may be for example a hollow hemisphere or parabolloid, polished or treated if required to increase the reflectivity of the material. A cheap and convenient form of the preselected liquid through which the electrical energy is discharged is water, either commercial or distilled. The electrodes should be formed of a rigid material resistant to corrosion and electrically conductive. A convenient material is stainless steel, and if required the discharge chamber may also be formed of stainless steel. It is possible to use, say, water, as the liquid to surround the hot plasma of the discharge region and yet to introduce selected molecules of another material into the discharge region for excitation by the electrical discharge in order to obtain an emitted radiation characteristic of this other material. A preferred form of this material is mercury which when excited emits U.V. radiation at 254 nm wavelength. A preferred form of an apparatus for achieving such excitation of the molecules of the material is arranged with the axis of its cylindrical intermediate portion vertical. There is an inlet at the bottom of the discharge chamber for connection to a supply of mercury for establishing a predetermind static height of mercury in the chamber. Further inlet and outlet means permit a swirling flow of water on top of the mercury. A first electrode is mounted in the wall of the discharge chamber and extends through the water to a point just above the surface of the mercury, which forms a second electrode, and on or near the axis of the chamber. Another electrode is similarly mounted and also extends through the water to another point adjacent the first point. Means are provided for applying a high voltage between the other electrode and either the first electrode or the mercury to cause an initiating discharge which results in mercury molecules being interspersed with water molecules in the region between the first electrode and the mercury. A main electrical discharge between the first electrode and the mercury may be synchronised with the initiating discharge. In order to maintain the electrical parameters of the discharge the rate of inflow of the water may be controlled; the swirling water induces a corresponding swirling movement of the mercury by friction. Alternatively, the water need not be swirling and the height of the mercury in the discharge chamber may be adjusted by control of the head of the mercury supply. It will be appreciated that electrodes must be electrically insulated from each other if there is to be a high voltage applied between them. Preferably the electrodes, other than liquid-type electrodes or those formed by the discharge chamber walls, are mounted in an insulating sleeve which is sealingly secured in a wall of the chamber. Instead of introducing water into the chamber via an inlet in the wall, the water may be introduced (and removed if required) via a passage or passages through the electrode, or between the electrode and the inner surface of its sleeve. In order to absorb selectively an unwanted wavelength or range of wavelengths, or reflect selectively a wanted wavelength or range of wavelengths, an appropriate coating may be applied to a surface of the discharge chamber or of the treatment chamber. Selective absorption may also be achieved by the use of an appropriate material as the U.V. transparent portion of the discharge chamber. Conveniently this material may be quartz doped with a metallic ion for example, tungsten. |
abstract | The present invention provides a method of observing a specimen in a field of view of an electron microscope comprising the acts of illuminating the specimen with an electron beam having a first angle and forming a first transmission image of the specimen in the field of view and adjusting the electron beam to a second angle and forming a second transmission image of the specimen in the field of view and calculating a degree of coincidence between the first and second transmission images. |
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063200910 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to a process for making a ceramic composition suitable for immobilizing actinides, in particular plutonium, thorium and uranium. The ceramic composition comprises pyrochlore, brannerite and rutile. The ceramic composition is attractive for immobilization of excess plutonium because of its extremely low leachability, the existence of natural mineral analogues that have demonstrated actinide immobilization over hundreds of millions of years in wet geological environments, and the high solid solubility of actinides in the ceramic thus providing a relatively low overall waste volume. Incorporation of plutonium into ceramic provides a form that is relatively easy to store but renders retrieval of the plutonium to be more difficult than other immobilized forms. The process relates particularly to a cold pressing and sintering process for making said ceramic composition, which allows for impurities in the actinides feedstock. 2. Description of Related Art Because of their extreme durability, ceramic forms have been studied extensively since the late 1970s for the immobilization of high level waste (HLW). The material called Synthetic Rock (SYNROC) is a titanate-based ceramic composed approximately of 30% zirconolite, 30% hollandite, 30% perovskite and 10% rutile. In the HLW application, actinides partition into the zirconolite and perovskite phases. U.S. Pat. Nos. 4,274,976 (Ringwood), 4,329,248 (Ringwood), 4,645,624 (Ramm, Ringwood) and 4,808,337 (Ramm et al) disclose the immobilization of HLW in synthetic rock. A form of synthetic rock to immobilize spent fuel (SYNROC-F) was reported by S. E. Kesson and A. E. Ringwood, "Safe Disposal of Spent Nuclear Fuel", Radioactive Waste Management and the Nuclear Fuel Cycle, Vol. 4(2), pp. 159-174, October 1983. This form of SYNROC consisted of 90 wt % uranium pyrochlore, 5 wt % hollandite and 5 wt % rutile. A variant of SYNROC-F, namely SYNROC-FA was reported by A. G. Solomah, P. G. Richardson and A. K. Mcllwain, "Phase Identification, Microstructural Characterization, Phase Microanalyses and Leaching Performance Evaluation of SYNROC-FA Crystalline Ceramic Waste Form", Journal of Nuclear Materials 148, pp. 157-165, 1987. This form of SYNROC consisted of uranium pyrochlore, perovskite, uraninite and hollandite. A cold press and sinter process is used in the production of mixed oxide (MOX) fuel from uranium and plutonium. The MOX process uses pressing pressures in excess of 20,000 psi and sintering temperatures of 1,700 deg C. in a 4% H.sub.2 atmosphere on a 24 hour cycle. Because the final product is to be used as fuel, impurities in the feedstock cannot be tolerated. SUMMARY OF THE INVENTION An object of this invention is to provide a process for making a ceramic material for immobilization of actinides, including plutonium, uranium, thorium, americium and neptunium, said immobilized actinides then being suitable for storage in an underground repository. The desired characteristics of such a ceramic material include: a) low leachability, b) high solid solubility of actinides in the ceramic, c) ability to incorporate "high-fired" PuO.sub.2 of nominal particle size of 20 microns and less than 50 microns, d) sufficient compositional flexibility to incorporate significant concentrations of Pu and neutron absorbers (such as gadolinium and hafnium) as well as varying impurities in the feed streams, e) thermal stability during high level waste glass pouring in can-in-canister configurations, f) high chemical durability in the geologic repository environments both before and after undergoing radiation damage from alpha decay, g) difficult recoverability of plutonium from the ceramic to impede proliferation. The ceramic composition comprises pyrochlore, brannerite and rutile. A pyrochlore matrix provides the means to incorporate a higher concentration of plutonium than a zirconolite matrix. SYNROC compositions have contained hollandite, which the present composition does not. Other compositions have utilized reactive plutonium (such as dissolved plutonium) whereas the present composition starts with unreactive ("high fired") plutonium. The present composition also tolerates greater than 50 wt % impurities in the actinide feedstocks. The present composition has been found to be stable when subjected to high level waste glass pouring in a storage canister. More specifically the mineralogical composition remains unchanged and the pellets retain their physical integrity. The present invention also relates to a process for making said ceramic composition which involves milling the actinides to a fine powder, blending the actinide powder with ceramic precursors, granulating the blend to facilitate flow into the press and die set, cold pressing the blend, followed by sintering. This is in contrast to the SYNROC process, which involves hot uniaxial or isostatic pressing. In addition, processes producing SYNROC have formed hollandite, which the present process does not. The production of mixed oxide (MOX) fuel also uses a cold press and sinter process, however there are critical differences in process parameters, which lead to differences in the end products. In addition, the MOX process cannot tolerate impurities in the uranium and plutonium feedstocks, whereas the present process does (up to 55.7 wt % in the actinide feedstock). |
039649660 | summary | BACKGROUND OF THE INVENTION This invention relates to a molten core retention system for a liquid-metal-cooled fast breeder reactor. In more detail, the invention relates to an in-vessel safeguard baffle assembly designed to prevent molten fuel from melting through the containment vessel for the reactor. Such a retention system is frequently called a "core catcher" in the literature and for brevity this name will be employed hereinafter. It is postulated that under certain conditions an accident could occur to a fast breeder reactor which could cause all or part of the fuel to melt. One example of a sequence of events that could lead to a release of radioactive material can be hypothesized as follows: A. Objects used during plant construction are forgotten and left in the piping system. PA1 B. Sodium flow during plant operation causes the objects to move downstream to the core inlet region. PA1 C. One section of the core is starved of coolant flow because of the blockage by foreign objects. PA1 D. Fuel pins melt in the region of starved flow. PA1 E. Molten material flows downward by gravity from the core region and collects on the reactor vessel bottom. PA1 F. The reactor vessel bottom melts as a result of the heat generated by the debris collected thereon. PA1 G. Sodium and radioactive debris flow from the reactor vessel into the guard vessel by gravity. PA1 H. The guard vessel bottom melts as a result of the heat generated by the radioactive debris collected thereon. PA1 I. Sodium and radioactive debris flow through the melted hole in the guard vessel to contaminate the biosphere. PA1 1. In-reactor vessel core catchers. PA1 2. In-reactor tank core catchers. PA1 3. Sodium-cooled trays in the reactor cavity. PA1 4. A cooled reactor cavity liner. PA1 5. A sacrificial bed within or outside of the reactor vessel. PA1 a. Space for containment of substantial quantities of radioactive debris material (possibly greater than 50% of the core) must be provided. PA1 b. Criticality of the contained debris must be prevented. PA1 c. The heat generated by the debris must be dissipated without overheating the containment structure. PA1 d. The debris material must be adequately cooled by natural means -- such as natural circulation of sodium. PA1 e. The containment structure must not introduce unacceptable resistance to sodium flow. PA1 f. The containment structure must withstand normal operating loads, anticipated transients and unlikely plant failure events without damage, for extended periods of operation. PA1 g. Manufacturing must be performed using conventional techniques. Clearly such a sequence of events -- or any other sequence of events leading to contamination of the biosphere with radioactivity, even though the probability of occurrence is very low -- cannot be tolerated. Various design approaches have been proposed to solve this potential problem. These include: A number of different attempts have been made to solve the problem with an in-vessel core catcher but these attempts were deterred from finding a practical solution by the inability to satisfy one or more of the following requirements: SUMMARY OF THE INVENTION Damage to the reactor vessel bottom from radioactive debris from an overheated core is prevented by a molten core retention system suspended from the core support plate and consisting of a horizontal baffle plate, a plurality of heat exchange tubes penetrating the plate and having openings in the top, and a cylindrical baffle surrounding the tubes and attached to the plate. The retention system is supported from the core support plate by hanger rods welded to the radial beam which are intermittently welded to the undersurfaces of the horizontal baffle plate. Preferably the upper end of the cylindrical baffle terminates in a lip having slots therein through which the hanger rods extend, permitting a small amount of radial movement of the retention system. |
048184705 | abstract | An apparatus for the remote examination of peripheral shroud hold down bolts on steam separators used in boiling water reactors is disclosed. The testing apparatus is used while the steam separator is in a holding pool. The testing apparatus includes a depending pole having attached at the bottom thereof an aluminum shoe. The shoe has a flat, upwardly exposed bottom and opens to one side at gathering surfaces to receive the lower ends of the shroud bolts. The upwardly exposed bottom of the shoe defines an aperture through which an upwardly exposed piezoelectric device is exposed for direct contact with the bottom of the bolt. An overlying clamp member is provided to clamp the lug onto the bottom of the shoe for testing. The clamp member defines a receiving slot to receive the shaft of the bolt immediately overlying the rectangular lug. The clamp member slides over the shaft at the slot into contact with the top of the lug. The clamp member clamps the lug end of the bolt securely onto the upwardly exposed piezoelectric device on the upwardly exposed bottom of the shoe. A piezoelectric ultrasonic test is then run from the bottom and radioactive portion of the bolt to and towards the top of the bolt. Testing for longitudinal sound waves, refracted sound waves and shear sound waves for inter granular stress corrosion cracking can occur despite the radioactive and remote under water location of the bolts. |
description | This application claims foreign priority under Paris Convention and 35 U.S.C. §119 to Korean Patent Application No. 10-2008-0098982, filed Oct. 9, 2008 with the Korean Intellectual Property Office. 1. Field of the Invention The present invention relates to an apparatus for forming stress corrosion cracks which occur in base metal or welded parts (which are homogeneous or heterogeneous) used in equipment such as reactor heads, steam generators or the like of nuclear power plants, and, more particularly, to an apparatus for forming stress corrosion cracks, in which stress corrosion cracks occurring in the equipment of nuclear power plants or apparatus industries during operation can be directly formed in a piping material actually used in nuclear power plants under environmental conditions similar to those of the nuclear power plants, thus predicting a crack propagation rate, thereby reducing actual risks of nuclear power plants or apparatus industries and effectively assuring nondestructive testing capability. 2. Description of the Related Art As is well known to those skilled in the art, there is a need to construct an experimental apparatus which operates under environmental conditions similar to those existing actual conditions which generate stress corrosion cracks in nuclear power plants. However, in the case of stress corrosion cracks occurring in a primary system, such as a steam generator, a pressure vessel of a reactor or the like, there is a dangerous probability of emitting radioactivity due to the lack of domestic simulation techniques. Thus, in order to ensure the safety of nuclear power plants and the reliability of NDT technologies, it is very important to realize techniques for producing natural cracks similar to actual defects occurring in nuclear power plants during operation. To this end, an experimental apparatus able to simulate actual conditions for generating stress corrosion cracks in nuclear power plant structures must be constructed. If so, the reliability and safety of nuclear power plants may be improved and techniques for precisely diagnosing defects in the power plants occurring during their operation may be advanced. In this way, when techniques for simulating stress corrosion cracks in a piping material actually used in nuclear power plants and for precisely diagnosing defects are ensured, inspection methods during the operation of the power plants may be developed and also data for safety regulation and repair criteria of nuclear power plant structures may be obtained. With the goal of forming stress corrosion cracks in a specimen, the following testing method is mainly illustrated. Specifically, a specimen is produced in the form of a C-ring or a U-band and is then loaded into an autoclave, after which appropriate tensile or compressive stress is applied thereto. In such a state, corrosion environmental conditions are set, and high-temperature high-pressure conditions are applied, thus forming crack. However, the conventional method of forming stress corrosion cracks is disadvantageous because the resulting cracks are obtained not using a piping material actually used for equipment of nuclear power plants or apparatus industries but using a simulation specimen, they are considerably different from actual cracks, thus making it impossible to effectively assure nondestructive testing capability. Accordingly, the present invention has been made keeping in mind the above problems encountered in the related art, and the present invention provides an apparatus for forming stress corrosion cracks, in which not a conventional specimen prepared according to a standard method (ASTM and so on) using an autoclave but a piping material actually used for equipment of nuclear power plants or apparatus industries is directly heated, so that stress corrosion cracks can be directly formed under conditions similar to those observed in nuclear power plants or apparatus industries using steam pressure generated by corrosion conditions in the pipe through direct heating, thus predicting a crack propagation rate, thereby reducing actual risks of nuclear power plants or apparatus industries and effectively assuring nondestructive testing capability. According to a preferred embodiment of the present invention, an apparatus for forming stress corrosion cracks comprises a heating unit, which includes a conductive member provided on an outer surface of one side of a tube specimen in a circumferential direction and a heating coil disposed adjacent to the conductive member to generate steam pressure in the tube specimen, an end holding unit for closing both open ends of the tube specimen so that the steam pressure generated in the tube specimen does not leak, and a control unit for controlling the heating unit and the end holding unit. As a preferred feature of the present invention, the heating unit may comprise an induction heating coil for inducing heating by forming a magnetic field using high-frequency current or a direct heating coil having a heating wire which is heated using power. As another preferred feature of the present invention, the end holding unit may comprise an upper plate and a lower plate for closing both ends of the tube specimen, and a tension bar including a hydraulic or pneumatic cylinder using hydraulic or pneumatic pressure as an operation pressure or an actuator rod using power to adjust a distance between the upper plate and the lower plate. As a further preferred feature of the present invention, the control unit may comprise a steam temperature measuring sensor for measuring a temperature of the steam in the tube specimen, an etchant temperature measuring sensor for measuring a temperature of an etchant in the tube specimen, a pressure measuring sensor for measuring an inner pressure of the tube specimen, and a controller electrically connected to the pressure measuring sensor to receive detection signals and to selectively output control signals to the heating unit and the end holding unit based on the received detection signals. As still a further preferred feature of the present invention, in order to ensure accuracy and safety in the use of the pressure measuring sensor, a cooling unit for lowering the temperature of the steam pressure in the tube specimen may be connected to the lower plate so that steam in the specimen is passed through the cooling unit having at least a predetermined length for phase transformation of steam into water to lower the temperature of the steam and may be simply provided in the form of a coil made of corrosion resistant material. The features and advantages of the present invention will be more clearly understood from the following detailed description, taken in conjunction with the accompanying drawings. The terms and words used in the specification and claims must be regarded as having concepts selected by the inventor as the best method of illustrating the present invention, and must be interpreted as having meanings and concepts adapted to the scope and sprit of the present invention for understanding the technology of the present invention. Hereinafter, a detailed description will be given of an apparatus for forming stress corrosion cracks according to a preferred embodiment of the present invention with reference to the appended drawings. In the drawings, it is noted that the same reference numerals are used throughout the different drawings to designate the same or similar components. In the present invention, a detailed description of functions or structures known in the related art is omitted so as not to obscure the purpose of the present invention. FIG. 1 shows a partial perspective view of the heating unit of the apparatus for forming stress corrosion cracks according to the present invention, and FIG. 2 shows main components of the apparatus for forming stress corrosion cracks according to the present invention. Also, FIG. 3 schematically shows the operation of the apparatus for forming stress corrosion cracks according to the present invention, FIG. 4 schematically shows the heating unit according to the present invention, and FIG. 5 shows the position and pattern of notches formed in a tube specimen according to the present invention. As shown in the drawings, the apparatus 1 for forming stress corrosion cracks according to the preferred embodiment of the present invention includes a heating unit 20 having a direct heating coil or an induction heating coil disposed adjacent to a specimen 10 having a tube shape (hereinafter referred to as a “tube specimen 10”), an end holding unit 30 for maintaining steam pressure generated due to heating, a cooling unit for decreasing steam pressure, and a measurement unit for measuring pressure. The end holding unit 30 comprises an upper plate 31 and a lower plate 32 for closing both ends of the tube specimen 10, and a tension bar 33 including a hydraulic or pneumatic cylinder 34 using hydraulic or pneumatic pressure as an operation pressure or an actuator rod 35 using power to adjust a distance between the upper plate 31 and the lower plate 32. The heating unit 20 functions to apply predetermined heat to the outer surface of one side of the tube specimen 10 to set application environments of temperature and pressure required to form stress corrosion cracks of the tube specimen 10. Examples of the material for the tube specimen 10 include STS 304, STS 316, STS 321, STS 347, STS 308, STS 309, Inconel 600, Inconel 690, Inconel 800, Inconel X750, and Inconel 718, which are actually used as pipes of nuclear power plants or apparatus industries. Also, homogeneous or heterogeneous welded parts for use in nuclear power plants may be used. The heating unit 20 is composed of a conductive member 21 made of magnetic material and attached to the outer surface of one side of the tube specimen 10 in a circumferential direction, and an induction coil or a direct heating coil spaced apart from the conductive member. The heating unit 20 comprises a direct heating coil 22 which is heated using power. To the induction coil is connected an external high-frequency current supplier. When high-frequency current is applied to the induction coil from the high-frequency current supplier, a high-frequency magnetic field is formed in the induction coil, and thus the conductive member adjacent to the induction coil is heated by eddy current loss or hysteresis loss occurring due to the high-frequency magnetic field, thus heating the outer surface of one side of the tube specimen 10 to which the conductive member is attached to a predetermined temperature. The heating process using the induction coil may be conducted through any technique known in the art, and a description thereof is omitted. Alternatively, when alternating current of 220 V is applied to the heating coil, the outer surface of one side of the tube specimen 10 to which the coil is attached may be heated to a predetermined temperature. The heating process using the heating coil may be conducted through any technique known in the art. On the inner surface of the tube specimen 10, a notch n enabling the formation of a stress corrosion crack at a desired position through mechanical or chemical processing such as lathe machining may be formed. The width of the notch n may be set to about 0.5˜50 mm, and the notch n may be provided in various forms, for example, ∪, v, and ␣, as illustrated in (a), (b) and (c) of FIG. 5. In this way, when the notch n is formed on a portion of the inner surface of the tube specimen 10, a position where a stress corrosion crack is generated may be easily controlled and a period of time required to form such a crack may be reduced, compared to when no notch is provided. The end holding unit 30 functions to stably maintain the steam pressure generated in the tube specimen 10 by means of the heating unit 20. If the end holding unit 30 has a structural feature able to stably close both ends of the tube specimen 10 to stably maintain the steam pressure in the tube specimen 10, a member having various shapes may be applied. In the present invention, as illustrated in the drawing, the end holding unit 30 preferably includes an upper plate 31 and a lower plate 32 respectively provided at both ends of the tube specimen 10 and a tension bar 33 for adjustably connecting the upper plate 31 and the lower plate 32. The upper plate 31 and the lower plate 32 may be respectively connected to both ends of the tube specimen 10 using a mechanical process or a hydraulic or pneumatic tool, thereby firmly holding both ends of the tube specimen 10. Such a connection process may be performed through any technique known in the art. The cooling unit 50 functions to lower the temperature of the steam pressure in the tube specimen 10. In order to ensure accuracy and safety in the use of a pressure measuring sensor, the cooling unit is connected to the lower plate 32 so that steam in the specimen is passed through the cooling unit having at least a predetermined length for phase transformation of steam into water to lower the temperature of the steam, and is simply provided in the form of a coil made of corrosion resistant material. The control unit 40 functions to rapidly heat the tube specimen using the heating unit 20 through on/off control and to control the temperature of the steam generated by the increase in the temperature of the etchant in the heated tube specimen through PWM control before reaching desired pressure. When the control unit is used, the temperature of the steam may be controlled to about 10° C. or lower, thereby continuously maintaining the steam pressure of 170˜180 bar. As shown in FIG. 6, the control unit 40 includes a steam temperature measuring sensor 41 provided in the tube specimen 10 to measure the temperature of the steam in the tube specimen 10, an etchant temperature measuring sensor 42 provided in the tube specimen 10 to measure the temperature of an etchant in the tube specimen 10, a pressure measuring sensor 45 for measuring the inner pressure of the tube specimen 10, and a controller 47 such as a micom or a computer electrically connected to the pressure measuring sensor 45 to receive detection signals and to selectively output control signals to the heating unit 20 based on the received detection signals. The respective sensors and the controller of the control unit 40 are known in the art and thus detailed descriptions thereof are omitted. Below, an experimental example using the apparatus for forming stress corrosion cracks according to the present invention is described. A specimen for the present experimental example was made of STS 304 (O.D.: 89 mm, t: 7.7 mm, yield strength: 41.8 kg/mm2) useful as a material for nuclear power plant structures. This specimen had a length l of 500 mm. To control the cracking position in the specimen, a notch having a depth of 0.5 mm was artificially formed at a position in a circumferential direction of the specimen through lathe machining. In order to set conditions for generation of a stress corrosion crack, 2 mol NaOH and 1 mol Na2S were placed in the tube specimen 10, and the temperature of 350° C. corresponding to the actual conditions for the generation of a stress corrosion crack in a nuclear power plant was maintained using a heating unit 20 and a control unit 40. Further, in order to apply at least a yield stress value to the tube specimen 10 at 350° C., a pressure of 170˜180 bar was continued. The experimental apparatus was maintained for 450 hours, after which the cracked portion was cut and then observed using an SEM and EDS. FIG. 7 shows the crack in a longitudinal direction of the tube specimen 10 at the portion of the tube specimen 10 to which stress is intensively applied and with which the etchant is brought into contact, and FIGS. 8 and 9 show EDS and SEM images of the portion of the specimen where the crack is propagated up to the outer surface of the specimen and is thus penetrated therethrough. As is apparent from the results of the analysis of EDS of FIGS. 8 and 9, the atomic % of Na was seen to be decreased toward the lower portion of the specimen from the upper portion thereof. This was considered to be because a crack was first generated at the upper portion of the tube specimen due to the steam generated under stress and corrosion conditions. As described above, the present invention provides an apparatus for forming a stress corrosion crack. According to the present invention, stress corrosion cracks occurring in the equipment of nuclear power plants or apparatus industries during operation can be formed in a tube specimen under conditions similar to those of the actual environment of nuclear power plants using steam pressure. Thereby, accuracy pertaining to the analysis of properties of stress corrosion cracks which are in actuality generated can be increased, thus improving reliability of nuclear power plants or apparatus industries and effectively assuring nondestructive testing capability, resulting in very useful industrial applicability. Although the preferred embodiment of the present invention using the tube specimen has been disclosed for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims. |
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047724480 | claims | 1. A locking nut adapted to be coaxially positioned in threaded engagement with an elongate threaded element and to crimpingly engage same in use, said locking nut having sidewalls and having no external apertures in said sidewalls and comprising: an internally threaded section which threadedly engages said elongate threaded element in use; and a crimpable cylindrical section integrally connected to said internally threaded section and extending from the outermost portion of said internally threaded section, wherein said elongate threaded element comprises an end section having a plurality of recesses provided on the external surface thereof, whereby relative rotation between said locking nut and said elongate threaded element is positively prevented when at least one of said plurality. a support pin having a longitudinal axis, a first pin portion and a second pin portion, said first pin portion being disposed within and passing through said through-bore and having an externally threaded section, and said second pin portion being fixedly secured within said bore and having a solid body section and a split-leaf base section, said solid body section having an outer diameter which is accommodated by said bore by a close clearance fit, said splitleaf base section having a split intermediate section which extends from said solid body section and has an outer diameter which is less than the outer diameter of said solid body section, and a split end section which extends from said split intermediate section and biasingly engages at least a portion of the wall of said bore, whereby said support pin is fixedly secured within said second structural member by a frictional fit and whereby loads applied transversely to said longitudinal axis of said support pin are reacted substantially in pure shear by said second pin portion substantially through said solid body section; and nut means having an internally threaded section which threadedly engages said externally threaded section of said first pin portion and cooperates therewith, whereby said first structural member is retained between said nut means and said solid body section of said support pin, wherein said first pin portion further comprises an end section positioned remotely from said second pin portion, adjacent said externally threaded section, and having a plurality of recesses provided on the external surface thereof; and wherein said nut means is a locking nut having sidewalls and having no external apertures in said sidewalls and further comprises a crimpable cylindrical section, which crimpable cylindrical section is integrally connected to said internally threaded section, extends from the outermost portion of said internally threaded section, and crimpingly engages at least one of said plurality of recesses of said support pin in use, whereby relative rotation between said locking nut and said support pin is positively prevented. wherein said first structural member is further provided with a second counter-bore in the surface thereof proximate to said second structural member; and wherein said first pin portion further comprises an annular shoulder which is adapted to be seated in said second counter-bore, and which merges with said solid body section of said second pin portion, whereby said first structural member is retained between said nut means and said annular shoulder. wherein said first structural member further comprises a first counter-bore in the surface thereof remote from said second structural member; and wherein said nut means is seated within said first counter-bore. wherein said plurality of recesses are longitudinal recesses. PG,31 wherein said first structural member is a nuclear reactor control rod guide tube flange and said second structural member is a nuclear reactor upper core plate. further comprising a locking nut retainer which includes a split cylindrical wall portion which is crimpable and has an axial slot defined therein; and tab means which extends radially from said wall portion and is positioned along the external surface thereof, said locking nut retainer being positioned around at least a portion of said nut; wherein said first structural member is provided with a first countyer-bore in the surface thereof remote from said second structural member, said first counter-bore having an annular recess radially defined in the wall thereof for accommodating at least the portion of said wall portion of said locking nut retainer which includes said tab means, and wherein said tab means is positioned within said annular recess, whereby said locking nut retainer is positively retained in said first structural member; and wherein said locking nut further comprises a crimp receiving section, at least a portion of said wall portion of said locking nut retainer crimpingly engaging said crimp receiving section, whereby said nut is positively retained in position around said support pin and relative rotation therebetween is positively presented. wherein said first counter-bore further receives at least a portion of said locking nut. wherein said plurality of recesses are longitudinal recesses. wherein said nut means is a spline nut and said at least one recess is a plurality of spline grooves. wherein said tab means is a pair of tabs positioned opposite one another and equidistant from said axisl slot of said wall portion, wherein said nut has a plurality of splines and a plurality of spline grooves positioned along the external surface thereof, said splines being alternatingly associated with said spline grooves, and wherein said wall portion of said locking nut retainer crimpingly engages at least two said spline grooves in use. wherein said locking nut retainer crimpingly engages two said spline grooves in use, said engaged spline grooves being spaced 144.degree. apart. a support pin having a longitudinal axis, a first pin portion and a second pin portion, said first pin portion being disposed within and passing through said through-bore and having an externally threaded section, and said second pin portion being fixedly secured within said bore and having a solid body section and a split-leaf base section, said solid body section having an outer diameter which is accommodated by said bore by a close clearance fit, said split-leaf base section having a split intermediate section which extends from said solid body section and has an outer diameter which is less than the outer diameter of said solid body section, and a split end section which extends from said split intermediate section and biasingly engages at least a portion of the wall of said bore, whereby said support pin is fixedly secured within said second structural member by a frictional fit and whereby loads applied transversely to said longitudinal axis of said support pin are reacted substantially in pure shear by said second pin portion substantially through said solid body section; nut means having an internally threaded section which threadedly engages said externally threaded section of said first pin portion and cooperates therewith, whereby said first structural member is retained between said nut means and said solid body section of said support pin; and a locking nut retainer which includes a split cylindrical wall portion which is crimpable and has an axial slot defined therein; and tab means which extends radially from said wall portion and is positioned along the external surface thereof, said locking nut retainer being positioned around at least a portion of said nut means; wherein said first structural member is provided with a first counter-bore in the surface thereof remote from said second structural member, said first counter-bore having an annular recess radially defined in the wall thereof and said first counter-bore accomodating at least said wall portion of said locking nut retainer, and wherein said tab means is positioned within said annular recess, whereby said locking nut retainer is positively retained in said first structural member; and wherein said nut means further comprises a crimp receiving section which has at least one recess provided in the external surface thereof, at least a portion of said wall portion of said locking nut retainer crimpingly engaging said at least one recess, whereby said nut means is positively retained in position around said support pin. wherein said first counter-bore further receives at least a portion of said nut means. wherein said at least one recess is a longitudinal recess or recesses. wherein said nut means is a spline nut and said at least one recess is a plurality of spline grooves. wherein said tab means is a pair of tabs positioned opposite one another and equidistant from said axial slot of said wall portion, wherein said nut has a plurality of splines and a plurality of spline grooves positioned along the external surface thereof, said splines being alternatingly associated with said spline grooves, and wherein said wall portion of said locking nut retainer crimpingly engages at least two said spline grooves in use. wherein said locking nut retainer crimpingly engages two said spline grooves in use, said engaged spline grooves being spaced 144.degree. apart. a split cylindrical wall portion which is crimpable and has an axial slot defined therein whereby said locking nut retainer is rendered substantially resiliently compressible along the radial axis thereof for insertion thereof during use, said wall portion being adapted to be positioned around the external periphery of said nut and to crimpingly engage at least a portion of the periphery of said nut in use; and tab means extending radially from said wall portion along the external surface thereof, wherein said structural member is provided with a counter-bore for accommodating at least said locking nut retainer and said elongate threaded element, said counterbore having an annular recess defined therein for accommodating said tab means, whereby said locking nut retainer is positively retained in said structural member and said nut is positively retained around said elongate threaded element when crimpingly engaged by said locking nut retainer. wherein said tab means is a pair of tabs positioned opposite one another and equidistant from said axial slot of said wall portion. wherein said nut has a plurality of splines and a plurality of spline grooves positioned along the external surface thereof, said splines being alternatingly associated with said spline grooves, and wherein said wall portion of said locking nut retainer crimpingly engages at least two said spline grooves in use. 2. The locking nut according to claim 1, wherein said plurality of recesses are longitudinal recesses. 3. A support pin system for fastening a first structural member having a through-bore defined therethrough to a second structural member having a bore defined therein, said support pin system comprising: 4. In combination with a nuclear reactor, a support pin system as defined in claim 3 wherein said first structural member is a nuclear reactor control rod guide tube flange and said second structural member is a nuclear reactor upper core plate. 5. A support pin system according to claim 3, 6. The support pin system according to claim 3, 7. The support pin system according to claim 3, 8. The support pin system according claim 3, 9. Support pin system according to claim 3, 10. The support pin system according to claim 9, 11. The support pin system according to claim 9, 12. The support pin system according to claim 9, 13. The support pin system according to claim 9, 14. The support pin system according to claim 13, 15. A support pin system for fastening a first structural member having a through-bore defined therethrough to a second structural member having a bore defined therein, said support pin system comprising: 16. The support pin system according to claim 3, 17. The support pin system according to claim 15, 18. The support pin system according to claim 15, 19. The support pin system according to claim 15, 20. The support pin system according to claim 19, 21. A locking nut retainer adapted to be positioned within and accommodated by a structural member having an elongate threaded element disposed therein for positively retaining a nut in position around said elongate threaded element, said locking nut retainer comprising: 22. The locking nut retainer according to claim 21, 23. The locking nut retainer according to claim 21, 24. The locking nut retainer according to claim 23, wherein said locking nut retainer crimpingly engages two said spline grooves in use, said engaged spline grooves being spaced 144.degree. apart. |
summary | ||
047864613 | abstract | Disclosed is a reactor internals hold down spring and upper head region cooling passage. The hold down spring utilizes a plurality of stacks of Belleville spring washers to provide spring load and deflection capability to hold down the reactor vessel internals. Each spring assembly includes a generally leak tight means for passing an adjustable coolant flow to the upper reactor vessel head region in order to maintain that region at inlet coolant temperature. The spring assemblies are angularly spaced about a core barrel support flange to cooperate with coolant flow passage formed therein. |
claims | 1. A method of fabricating metallic fuel from surplus plutonium, the method comprising:combining plutonium oxide powder consisting essentially of PuO2 with uranium oxide powder to obtain a mixed powder, the plutonium oxide powder being present at a quantity of about 20% to 50% by weight relative to a total weight of the mixed powder, the uranium oxide powder having a 235U enrichment of less than 5.0 wt %;electroreducing the mixed powder in a porous basket immersed in a bath of molten salt so as to convert the mixed powder to a first alloy of plutonium and uranium, the first alloy being in a form of a porous metal sponge;pressing the first alloy to remove a majority of the molten salt adhered to the first alloy to form a pressed alloy-salt mixture; andisolating the first alloy from a remainder of the molten salt by melting the pressed alloy-salt mixture via electromagnetic induction in a furnace and discharging the first alloy from a bottom of the furnace. 2. The method of claim 1, wherein the pressing includes removing at least 80% of the molten salt adhered to the first alloy. 3. The method of claim 1, wherein the pressing includes a simultaneous application of heat to facilitate removal of the molten salt adhered to the first alloy. 4. The method of claim 1, further comprising:reintroducing the molten salt back into the bath of molten salt after the pressing and isolating the first alloy. 5. The method of claim 1, wherein the plutonium oxide powder is from a can of a recovered product of reprocessed nuclear fuel or nuclear weapons. 6. The method of claim 1, further comprising:sampling the mixed powder to verify a quantity of the plutonium oxide powder therein prior to the electroreducing. 7. The method of claim 1, wherein the plutonium oxide powder and the uranium oxide powder are not commingled until the combining to obtain the mixed powder. 8. The method of claim 1, further comprising:adding a Group 4 metal and uranium metal to the first alloy after the isolating so as to form a second alloy. 9. The method of claim 8, wherein the adding a Group 4 metal and uranium metal is performed such that the second alloy has about 20% to 50% plutonium by weight. 10. The method of claim 8, wherein the adding a Group 4 metal includes introducing zirconium to the first alloy. 11. The method of claim 8, further comprising:forming a fuel rod from the second alloy. |
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055639224 | claims | 1. A method for indicating a control rod position of a nuclear reactor including the steps of: a. applying a primary AC current to a primary coil of a rod position sensor, said primary AC current induces a secondary AC signal on a secondary coil of said rod position sensor; b. converting a primary AC signal, which is generated by the primary AC current, and the secondary AC signal to a primary DC signal and a secondary DC signal, respectively; c. dividing the secondary DC signal by the primary DC signal to produce a ratiometric signal; d. compensating the ratiometric signal for ambient variations on the rod position sensor producing a compensated ratiometric signal; e. calculating an analog linear correction signal by inputting the compensated ratiometric signal into a linear correction circuit; and f. summing the compensated ratiometric signal with the analog linearity correction signal to obtain a module output signal which is a linear representation of the control rod position. a. a primary AC current source, said source applies an AC current to a primary coil of a rod position sensor which induces a secondary AC signal across a secondary coil of said rod position sensor; b. first means for converting a primary AC signal, which is generated by the primary AC current source, and the secondary AC signal to a primary DC signal and a secondary DC signal, respectively; c. second means for dividing the secondary DC signal by the primary DC signal to produce a ratiometric signal; d. a compensation circuit for compensating the ratiometric signal for ambient variations on the rod position sensor and producing a compensated ratiometric signal; e. third means for calculating an analog linearity correction signal from the compensated ratiometric signal; and f. fourth means for summing the compensated ratiometric signal with the analog linearity correction signal to obtain a module output signal which is a linear representation of the control rod position. 2. The method defined in claim 1 further including the step of galvanically isolating the primary AC current by inducing a primary AC voltage across a transformer, with the value of the primary AC voltage being proportional to the value of the primary AC current. 3. The method defined in claim 1 in which the step of converting the primary and secondary AC signals to primary and secondary DC signals includes filtering the primary AC signal and the secondary AC signal through first low pass filters to eliminate any high harmonic content. 4. The method defined in claim 3 in which the step of converting the primary and secondary AC signals to primary and secondary DC signals includes passing said signals through a rectifying circuit to produce rectified signals. 5. The method defined in claim 4 in which the rectified signals are passed through second low pass filters to smooth the respective signals. 6. The method defined in claim 1 in which the step of compensating the ratiometric signal for ambient variations on the rod position sensor includes separately adjusting zero and span adjustments, said adjustments being decoupled from each other providing for separate calibration of each adjustment. 7. The method defined in claim 1 in which the step of compensating the ratiometric signal for ambient variations on the rod position sensor includes adjusting a temperature compensation adjustment circuit. 8. The method defined in claim 1 further including the step of programming an EPROM with correction data for each control rod. 9. The method defined in claim 1 further including the step of generating a digital address location from the compensated ratiometric signal by passing the signal through an analog-to-digital converter. 10. The method defined in claim 9 further including the step of addressing an EPROM with the digital address location and outputting a digital linearity correction signal. 11. The method defined in claim 10 further including the step of converting the digital linearity correction signal to the analog linear correction signal by passing the digital linearity correction signal through a digital-to-analog converter. 12. The method defined in claim 1 further including the step of detecting a rod bottom position from the module output signal and an adjustable rod bottom threshold set-point. 13. A system for indicating a control rod position of a nuclear reactor including: 14. The system as defined in claim 13 in which a transformer galvanically isolates the primary AC signal by inducing a primary AC voltage across a secondary coil of the transformer, the value of the primary AC voltage being proportional to the value of a primary AC current produced by the primary AC current source. 15. The system as defined in claim 13 in which the first means includes a rectifier connected in series with and between two low pass filters. 16. The system as defined in claim 13 in which the compensation circuit includes decoupled adjustments for zero and span positions of the control rod and for temperature compensation. 17. The system as defined in claim 13 in which the third means includes a linear correction circuit which further includes an analog-to-digital converter which addresses an EPROM, said EPROM outputs a digital linearity correction signal which is converted to an analog linearity correction signal by a digital-to-analog converter. 18. The system as defined in claim 17 in which the analog-to-digital converter includes a clock which is connected to a voltage-to-frequency converter and a timing and control logic circuit, said voltage-to-frequency converter and said timing and control logic circuit being connected to a counter which is connected to the EPROM. 19. The system as defined in claim 13 in which the fourth means is a summation circuit which includes an operational amplifier and a plurality of resistors. 20. The system as defined in claim 19 in which a magnitude comparator is connected to an output of the summation circuit, said magnitude comparator compares the module output signal with an adjustable rod bottom threshold set-point and determines if a rod bottom condition exists. 21. The system as defined in claim 13 in which the second means is an analog divider. |
046541896 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to improvements in or relating to high temperature reactors. More particularly, the present invention is concerned with medium power nuclear reactors in which the primary circuit is enclosed by a prestressed concrete pressure vessel or similar prestressed concrete reactor vessel, and this is supported on foundation walls. A reactor protection building surrounds the prestressed concrete reactor vessel, and is associated with a reactor auxiliary building. The reactor includes a reactor core which is made up of a bed of spherical full elements or pebble bed, which once through the reactor core under the force of gravity, and includes a charging installation for the introduction, as well as means for the withdrawal of operating elements. The pertaining charging conduits, pellet tubes, and exit pipes, as well as functional components for closing, metering, counting, distributing, and collecting of the operating elements are associated with the charging installation. The reactor also includes armored ducts which penetrate the prestressed concrete pressure vessel. 2. Background of the Art Feed apparatus, or similar charging installations, for nuclear power reactors of medium power in pebble bed reactors are known, such apparatus are designed in accordance with modular design principles, and contain functional components, which can be moved by drives, for the introduction, withdrawal, distribution, and sorting of fuel elements. The functional components exhibit bores for the passage of pellets, and they are set in a block, or a plate, which is respectively provided with connecting bores for the pellets. Such feed apparatus are shown in West German Utility Model No. 6,753,677; West German Auslegeschrift No. 1,589,532; and West German Offenlegungsschrift No. 2,357,426. The feed apparatus which is described in West German Offenlegungsschrift No. 2,357,426 is used, for example, in the nuclear power plant THTR-300, in which the feed station for the operating elements, fuel pellets and pure graphite pellets, is arranged in an auxiliary building and at a lower elevation than the high temperature reactor (HTR). A central withdrawal duct is provided in this HTR for the removal of the operating elements, and the duct leads to a burn-up measuring apparatus which is arranged in the charging room beneath the prestressed concrete pressure vessel. Because the fuel pellets are passing several times through the reactor core, until they have reached the desired final degree of burn-up, such a measuring apparatus is a necessity. For the same reason, a feed device is installed in the charging room, for conveying those operating elements which can still be used to the reactor core. The device is also used for fresh operating elements. In order to prevent damage of the operating elements due to impinging on the upper surface of the pebble bed or filling, the upper ends of the feed conduits exhibit a delay or deceleration apparatus. The state of the art further includes German Pat. No. 1,281,046 which also shows a feed apparatus for a so-called pebble bed reactor. The withdrawal and sorting device for the fuel pellets therein is also arranged beneath the reactor core and also exhibits a burn-up measuring apparatus. In accordance with the burn-up readings, the pellets are transferred again into the core, or they are separated from the cycle. SUMMARY OF THE INVENTION It is therefore a general object of the present invention to provide an installation for introduction and withdrawal of operating elements in a high temperature reactor of the type briefly described in the foregoing, which can be built and operated more economically than known installations. In accordance with one aspect of the present invention these objects are attained by an installation for the introduction and for the withdrawal of operating elements which comprises: (a) at least two input stations, with each input station being arranged within said reactor protection building and above said prestressed concrete pressure vessel; and for each input station, at least one distributor block operatively connected thereto; (b) a plurality of charging hatch blocks, with at least two charging hatch blocks being operatively connected to a pertaining distributor block; and each charging hatch block having means for containing functional components for closing, metering, and counting, and said means being adapted to be positioned on said prestressed concrete pressure vessel ceiling; and means operatively connected within a corresponding armored duct, said means including a selector, and charging conduits operatively connected thereto; (c) said charging conduits being shaped as helical conduits, and each charging conduit being adapted to be followed beneath said prestressed concrete pressure vessel ceiling by at least one inlet pipe which is adapted to be disposed in a random or meandering configuration; (d) a plurality of exit conduits, discharge separators, and singularizers, with pertaining exit conduits being operatively connected in said prestressed concrete pressure vessel bottom to pertaining ones of said armored ducts, and wherein in each pertaining exit pipe a pertaining scrap separator is operatively connected to at least one pertaining singularizer; (e) a first exit hatch block arranged outside of said prestressed concrete pressure vessel and within said foundation walls, said first exit hatch block being connected to a pertaining scrap separator and a shielded exit hatch channel, said channel being adapted to pass through said reactor protection building, and said first exit hatch block being adapted to contain functional components for shutoff and counting; (f) a second exit hatch block arranged outside of said foundation walls and inside said reactor protection building, and adapted to be operatively connected to said exit hatch channel and at least one exit station operatively arranged in said reactor auxiliary building, said second exit hatch block being adapted to contain functional components for metering, counting, and shutoff; and (g) wherein pertaining charging hatch blocks and said first exit hatch block are respectively equipped with closing devices, said devices being of at least two different configuration types, whereby for charging and withdrawal operations standard devices are used, and for isolating/closing of the primary cycle, for the control of the non-standard conditions, special devices are used. In accordance with the present invention, several important advantages are gained in such high temperature reactor with charging installation for introduction and withdrawal of operating elements. Thus, a delay or deceleration apparatus for operating elements is obviated at the entry into the reactor core, due to the particular configuration of the various charging and inlet pipes and the conduits. The utilization of selection switches allows the reduction in number of required functional components. The mechanical elements of the selection switches, as well as the elements of the combined singularizers/scrap separators, are integrated in a practical manner in the prestressed concrete pressure vessel because they are interiorly arranged in armored duct passages, in such a way that problems generally do not arise during charging, or similar feeding operations, and during removal operations with respect to the sealing of the primary circuit. The first exit hatch block, for one is arranged below the pertaining level of the prestressed concrete pressure vessel and, accordingly, is easily accessed. It is also located within the space enclosed by thick foundation walls, and these provide a shield for it at the same time. The input stations are also installed outside of the prestressed concrete pressure vessel, such that a simple pipe compensation is possible for the succeeding pebble tubes and similar ducts. The conduits which are subjected to pressure during normal operation are arranged within the reactor protection building. Of particular advantage is the utilization of various shutoff devices at the functional components. For operational charging and withdrawal sequences or procedures, one can use armatures and similar equipment, for example, ball cocks, ball valves, and the like. These close the charging hatch portion at the commencement in the distributor blocks and at the terminus in the charging blocks. In the case of incorrect start-up, or in the case of failure of one of these armatures, gas in the hatch is set free and passes into the reactor protection building because the entire input assembly is arranged in this building. Further standard armatures are adapted to close the exit hatch stretch between the first and the second exit hatch block. In the latter is done the withdrawal of the operating elements from the reactor protection building into the reactor auxiliary building in which the exit station is disposed. When gas is set free into the reactor auxiliary building, due to control errors or due to failure of an armature, only the contents of a hatch is released. All pebble tubes which lead out of the prestressed concrete pressure vessel are adapted to be directly closed by special armatures at the interface locations of pertaining passages through the pertaining vessel walls, for example, double-seat closing armatures and similar shutoff valves. These armatures ensure the safe isolation of the primary circuit even in the case of disruptive situations, they are designed in accordance with disruption-incident requirements and similar fail-safe considerations, and they are arranged within the charging hatch locks and in the first exit hatch block. The configuration of all components which follow in line behind the special armature, beginning at the prestressed concrete pressure vessel, is done in accordance with the operational requirements. The German Patent Application No. P 3,344,527.3 describes a high temperature reactor in which the charging installation for introduction and withdrawal of operating elements can be used. The subject matter of this application is incorporated herein. |
061577014 | summary | BACKGROUND OF THE INVENTION The present invention relates to an x-ray generating apparatus using plasma generated by irradiating a laser beam to a target, and to an x-ray microscope comprising an X-ray generating apparatus using such laser plasma. For example, an x-ray tube or a plasma x-ray source is known as the x-ray source of an x-ray microscope, x-ray laser, an x-ray lithography apparatus, an x-ray photoelectron microscope, an x-ray analyzer or the like. A plasma x-ray source is arranged to use x-rays generated by interaction between electrons and highly ionized ions in plasma. As a method of generating such high-density plasma, there is known a laser excitation method for example. In this specification, plasma generated by a laser excitation method is called laser plasma. Laser plasma is to be generated by condensing a laser beam on the surface of metal, such as Al, Mo, Au or the like, in the form of pulses each having a width of several ns, the laser beam having been stopped down by a lens or a mirror such that its diameter is for example about 10 .mu.m.about.100 .mu.m. As an x-ray microscope comprising an x-ray generating apparatus using such laser plasma, there is known a microscope of the type in which x-rays from the x-ray generating apparatus are irradiated to a sample and the transmitted x-rays are measured. FIG. 21 illustrates a schematic arrangement of a conventional x-ray microscope of the type above-mentioned. This microscope is arranged such that a laser beam 52 is irradiated to a target 51 to generate plasma, that x-rays emitted from the plasma are condensed on a sample 55 in a sample cell 54 by a mirror 53, and that x-rays having passed through the sample 55 are detected by a two-dimensional detector 57 through an enlarging optical system 56. In an x-ray generating apparatus to be used in such an x-ray microscope or for other purpose, the following arrangement is known as a mechanism for irradiating a laser beam to a target to generate x-rays. That is, flat or disk-like solid metal is for example used as the target, a laser beam is condensed on the surface of the solid metal to generate high-density plasma, and x-rays emitted from free-expanding plasma are guided to the outside of the x-ray generating apparatus. FIG. 22 illustrates, as an example, a schematic arrangement of the x-ray generating apparatus having the arrangement above-mentioned. In the arrangement in FIG. 22, when a laser beam 7 is focused on and irradiated to the surface of a target 1 of Al, Mo, Au or the like, laser plasma 6 is generated. The laser plasma 6 not only emits scattering particulates composed of neutral particles, charged particles 8 such as ions, electrons and the like, but also x-rays 9. The x-ray generating apparatus is arranged to use, as the x-ray source, such x-rays 9 emitted from the plasma 6. Usually, the x-rays 9 from the plasma 6 are irradiated to an x-ray supply object 11 through an optical element 10 such as a mirror or the like. In an x-ray analyzer for example, the x-ray supply object 11 is used as a sample to be analyzed, x-rays are irradiated thereto and the x-rays on the sample surface are analyzed. In an x-ray microscope, the x-ray supply object 11 is used as a sample to be observed and a detector is disposed therebehind. In the x-ray generating apparatus having the arrangement above-mentioned, whose pulse shape is controlled by making the laser source the form of multi-pulses or short pulses controls the wavelength of the generated x-rays and the like. However, such an x-ray generating apparatus of prior art is disadvantageous in that the amount of x-rays to be supplied to the x-ray supply object cannot readily be increased. More specifically, to improve the x-rays generating efficiency in the apparatus in FIG. 22, it is required to heat or increase the volume of generated plasma by controlling the pulse shape of the laser source. However, since the plasma generally expands freely at a high speed in a vacuum, it is difficult to control the motion of the plasma itself and the plasma momentarily freely expands and spreads. This results in failure to sufficiently improve the x-ray generating efficiency. In the x-ray generating apparatus in FIG. 22, there is disposed the optical element 10 for introducing the x-rays 9 to the x-ray supply object. In addition to the x-rays 9, scattering particulates composed of charged particles 8 and neutral particles are emitted from the plasma 6, and reach and stick to the surface of the optical element 10, thereby to lower the reflection efficiency of the x-rays 9. This contributes to a reduction in the amount of x-rays to be supplied to the x-ray supply object 11. Thus, the following countermeasure are taken. To prevent scattering particles from sticking to the optical element 10, there are disposed slits 12 and a scattering particulate preventing means 13 in the direction in which the scattering particles advance from the plasma 6 toward the optical element 10. In the scattering particulate preventing means 13, there may be for example used a method in which there is used a high-speed mechanical shutter arranged such that using a difference in speed between the x-rays and the scattering particulates, the shutter is closed to intercept the passage of the scattering particulates after the high-speed x-rays 9 have passed therethrough. Also, there may be used a method in which a gas inflow device is disposed to let gas to flow from the outside into the path of scattering particulates, causing the gas to come into collision with the scattering particulates to change the tracks thereof. However, such countermeasures cannot securely prevent the scattering particulates from sticking to the optical element 10. On the other hand, the following apparatus is conventionally known as an x-ray generating apparatus improved in x-ray generating efficiency as compared with the x-ray generating apparatus in FIG. 22. That is, the apparatus is arranged such that an x-ray transmitting film is disposed at one side of the target such that there is formed, between the target and the x-ray transmitting film, a space in which plasma is to be confined. FIG. 23 shows, as an example, a schematic arrangement of such an x-ray generating apparatus. In the arrangement in FIG. 23, an x-ray transmitting film 72 is so disposed as to form a space 73 adjacent to a tape-like target 71, and plasma generated by irradiating a laser beam 74 to the target 71 is confined in the space 73. In the x-ray generating apparatus in FIG. 23, x-rays are generated in the order as shown in FIG. 24. When the laser beam 74 is irradiated to the target 71 as shown in FIG. 24(a), a target at the irradiation position is evaporated to generate plasma, and x-rays 75 emitted from the plasma pass through the x-ray transmitting film 72 and are then released to the outside, as shown in FIG. 24(b). At this time, a hole is formed in the target 71 by the laser beam 74. Further, particulates are emitted together with the x-rays 75 from the plasma thus generated, but these particulates are reduced in speed by the x-ray transmitting film 72. As shown in FIG. 24(c), a hole is formed in the x-ray transmitting film 72 by the collision of particulates therewith or by the plasma pressure. Accordingly, the particulates are emitted together with the x-rays 75 through this hole. When the irradiation of a laser beam is finished and the next irradiation is to be conducted, the target 71 and the x-ray transmitting film 72 are moved at their bored portions such that unbored portions of the target 71 and the x-ray transmitting film 72 are located as facing the laser beam irradiation position as shown in FIG. 24(d). In the x-ray generating apparatus having the arrangement above-mentioned, since the x-ray transmitting film 72 is disposed, the plasma is confined in the space 73 to improve the x-ray generating efficiency. However, the scattering particulates are released through the bored portion of the x-ray transmitting film 72 as above-mentioned. This requires a device for eliminating such scattering particulates as done in the apparatus in FIG. 22. It is therefore required to dispose a scattering particulate preventing means such as a high-speed mechanical shutter 76 or the like as shown in FIG. 23. The scattering particulate preventing means such as the high-speed mechanical shutter 76 or the like is disposed between the target and a sample (x-ray supply object). Due to the presence of such scattering particulate preventing means, the target-sample distance in the order of cm is required. The x-rays emitted from the target can substantially be regarded as those from a point light source. Accordingly, when the target-sample distance is great, the amount of x-rays irradiated to the sample is disadvantageously reduced. Further, the requirement for such a scattering particulate preventing means causes the following trouble when such an x-ray generating apparatus is applied to an x-ray microscope shown in FIG. 21. In the x-ray microscope shown in FIG. 21, the condensing mirror 53 between the target 51 and the sample cell 54 is required because the distance between the sample 55 and the x-ray source is long due to the disposition of a scattering particulate preventing means. Due to the provision of the condensing mirror 53, the wavelength characteristics of x-rays reflected by the condensing mirror 53 should accord with the x-ray wavelength characteristics of the enlarging optical system 56 for enlarging and guiding the transmitted x-rays to the detector side. If the x-ray wavelength characteristics of these two optical systems are not identical with each other, it is not possible to condense the x-rays from the x-ray generating apparatus or enlarge the transmitted x-rays. This fails to produce a good x-ray image to disadvantageously lower the quality of x-ray analysis of the sample. |
050874101 | description | |
claims | 1. A spacecraft comprising:a propulsion device comprising:neutron beam regulator system comprising:a magnetic coil configured around a neutron beam that is produced by a neutron beam source and emitted by a neutron beam generator;wherein the neutron beam is emitted from the spacecraft to propel the spacecraft in space. 2. The spacecraft of claim 1, wherein the neutron beam source is a generated neutron beam source and wherein the propulsion device comprises a power control system comprising:a neutron beam source power supply output. 3. The spacecraft of claim 1, wherein the magnetic coil is a power generated magnetic coil that produces a magnetic field through a supply of electrical power to the magnetic coils. 4. The spacecraft of claim 3, wherein the neutron beam regulator system comprises:a magnetic coil power supply output;a neutron beam source power supply output;magnetic coil power sensor;a power safety feature configured to prevent power supply to said neutron beam source power supply output when said magnetic coil power supply sensor detects that a power level below a threshold power level is being drawn from the magnetic coil power supply output;whereby the neutron beam generator will not receive power from the power control system unless the magnetic coil is drawing said threshold power level and producing a confining magnetic field. 5. The spacecraft of claim 1, wherein the neutron beam source is a self-contained neutron beam source. 6. The spacecraft of claim 5, wherein the magnetic coil is a power generated magnetic coil that produces a magnetic field through a supply of electrical power to the magnetic coils. 7. The spacecraft of claim 6, wherein the neutron beam regulator system comprises:a magnetic coil power supply output;magnetic coil power sensor;a power safety feature configured to prevent generation of the neutron beam from the neutron beam source when said magnetic coil power supply sensor detects that a power level below a threshold power level is being drawn from the magnetic coil power supply output;whereby the neutron beam generator will generate the neutron beam unless the magnetic coil is drawing said threshold power level and producing a confining magnetic field. 8. The spacecraft of claim 1, wherein the neutron beam source is a self-contained neutron beam source. 9. The spacecraft of claim 8, wherein the self-contained neutron beam source comprises Californium 252. 10. The spacecraft of claim 8, wherein the self-contained neutron beam source comprises Cesium-137. 11. The spacecraft of claim 8, wherein the self-contained neutron beam source comprises both Beryllium and Polonium. 12. The spacecraft of claim 8, wherein the magnetic coil is a self-contained magnetic coil that produces a magnetic field without a supply of electrical power to the magnetic coils. 13. The spacecraft of claim 9, wherein the self-contained magnetic coil comprises neobdium. 14. The spacecraft of claim 9, wherein the spacecraft is propelled without utilizing any electrical power. 15. The spacecraft of claim 9, wherein the spacecraft is self-contained propelled spacecraft that is propelled by the propulsion device without utilizing any electrical power. 16. The spacecraft of claim 1, wherein the magnetic coil is a substantially continuous coil that extends a length of the neutron beam. 17. The spacecraft of claim 16, wherein the magnetic coil extends substantially the length of the neutron beam, at least 80% of the length from a beam outlet to where the neutron beam exits the spacecraft. 18. The spacecraft of claim 1, comprising a plurality of magnetic coils configured to extend around a neutron beam. 19. The spacecraft of claim 18, wherein the plurality of magnetic coils are discrete coils having discrete coil power inputs. 20. The spacecraft of claim 1, wherein the magnetic coil extends upstream of a neutron beam output. |
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summary | ||
description | Preferred embodiments of the present invention will hereinafter be described in detail with reference to the drawings. Note that in FIGS. 1 to 16, the thickness of each component, the width of each slot, the number of radiation-absorbing plates, the ratio of the dimensions of each component, etc., do not always agree with reality. Referring to FIG. 1, there is shown an x-ray scatter reducing grid (hereinafter referred simply to as a grid) 1 in accordance with a first embodiment of the present invention. The grid 1 has support members (first support members) 2, 2 consisting of radiation-transmitting material (radiation non-absorbing material) such as wood, aluminum and the like. The support members 2 are formed thick and connected at the opposite ends of each member to two connecting members (second support members) 6. That is, the support members 2 and the connecting members 6 as a whole constitute a rectangular frame 8,thereby giving rigidity to the grid 1. The connecting members 6 and the support members 2 may be coupled by means of adhesion, or they may be formed integrally with one another. While this first embodiment is provided with the connecting members 6, structure without the connecting members 6 is also possible. Similarly, in other embodiments to be described later, structure without the connecting members 6 is possible. The grid 1 further has radiation-absorbing plates 4. Each radiation-absorbing plate 4 consists of a plate containing a substance, which absorbs radiation relatively well, such as lead, tantalum, tungsten and the like. Note that in other embodiments to be described later, radiation-absorbing plates also consist of the same material. In the support members 2 of the first embodiment, from the upper edge 2a thereof toward the lower edge 2b a plurality of plate-receiving means (in this embodiment, slots 14) are formed in parallel at predetermined intervals at approximately half (xc2xd h) of the height h of the support member 2, as shown in FIG. 1B. The slots 14 extend in a direction going substantially toward the side of a radiation source (not shown), i.e., in a direction perpendicular to the paper surface of FIG. 1A. On the other hand, the radiation-absorbing plate 4 is formed with two parallel slots 16 (which extend in the direction opposite from the slots 14 of the support member 2), at positions corresponding to the two opposite support members 2, i.e., positions crossing the opposite support members 2 perpendicularly. That is, each slot 16 of the radiation-absorbing plate 4 is formed from the lower edge 4b thereof toward the upper edge 4a at approximately half (xc2xd h) of the height h of the radiation-absorbing plate 4. If the slots 16 of the radiation-absorbing plates 4 are positioned with respect to the slots 14 of the support members 2 and engage with the slots 14, a linear grid, i.e., a grid with the radiation-absorbing plates 4 disposed in parallel at predetermined intervals, is constructed as shown in FIG. 1A. In this construction, the radiation-absorbing plates 4 are disposed in parallel to one another and form a parallel grid and are also disposed at right angles to the support members 2. In this way, the support members 2 are capable of supporting and holding the radiation-absorbing plates 4 at predetermined positions. Since the slots 14 and 16 each have a dimension of half the height h of the respective members, the upper edge 4a of the radiation-absorbing plate 4 becomes substantially coplanar with the upper edge 2a of the support member 2 after fabrication. The height dimension h of the radiation-absorbing plate 4 is, for example, 1 to 3 cm, while the thickness is 0.1 mm. In addition, the spacing between adjacent slots 14 of the support member 2, i.e., the intervals at which the radiation-absorbing plates 4 are disposed, is approximately 1 mm. In fabricating the radiation-absorbing plates 4 and the support members 2, the radiation-absorbing plates 4 are inserted in the support members 2 through the respective lower edges 16 and upper edges 14. In this case, the height h of the support member 2 is short compared with the longitudinal direction thereof, and consequently, the resistance during the insertion becomes low. Furthermore, the insertion up to half of the height h is very easy because the resistance between the slot 16 of the radiation-absorbing plate 4 and the slot 14 of the support member 2 is much lower. The same may be said of the following embodiments in which the slot length is approximately half of the height h. Of course, the same is also true of the case where the length of one slot is one-third of h and the other slot length is two-thirds of h. After fabrication, the radiation-absorbing plates 4 and the support members 2 support one another without having solid matter as a member intervening between adjacent radiation-absorbing plates 4, and consequently, the radiation-absorbing plates 4 and the support members 2, as they are, can hold the fabricated form and result in a so-called self-supporting grid. The fixation between the radiation-absorbing plates 4 and the support members 2 may remain inserted, or the fixation may be reinforced by an adhesive agent, fusing, etc. Reinforcing the structure by an adhesive agent, fusing or the like is likewise possible for other embodiments that are to be described later. FIGS. 2 and 3 show a grid 20 similar to the grid 1 of the first embodiment, constructed according to a second embodiment of the present invention. Notice that in FIG. 3, the thickness of each component and the connecting members 26 shown in FIG. 2 are omitted for a clear understanding of the present invention. As illustrated in FIGS. 2 and 3, the essential difference between the grid 20 of the second embodiment and the grid 1 of the first embodiment is that a radiation-absorbing plate 24 has no slot and the slots 34 of a support member 22 extend from its upper edge 22a to the vicinity of its lower edge 22b. The manufacture of the radiation-absorbing plate 24 is easy because it has no slot. When fabricating the grid 20, all that is required is to insert the radiation-absorbing plates 24 into the slots 34 of the support members 22. As the slots 34 of the two support members 22 are aligned with one another and formed in parallel, the radiation-absorbing plates 24 are disposed in parallel and constitute a parallel grid, as with the first embodiment. In FIGS. 2 and 3, although the number of radiation-absorbing plates 24 is omitted for convenience, a large number of radiation-absorbing plates 24 are actually disposed in the slots 34 of the support members 22. It is preferable that the radiation-absorbing plates 24 be bonded to the slots 34 of the support members 22 so that the plates 24 do not to move within the slots 34. Alternatively, protrusions (FIG. 3) may be formed on the radiation-absorbing plate 24 to interpose the support member 22 therebetween in order to prevent positional misalignment. In this case, the fixation between the protrusions 25 and the support member 22 can also be reinforced by adhesion. FIGS. 4 and 5 show a grid 40 constructed according to a third embodiment of the present invention. In the third embodiment, the plate-receiving means for receiving and supporting radiation-absorbing plates is constructed by grooves 54 formed in support members 42. Note that in FIG. 5, the connecting members 46 shown in FIG. 4 are omitted for a clear understanding of the present invention. As illustrated in FIGS. 4 and 5, the grid 40 of the third embodiment, as with the aforementioned two embodiments, is a linear grid, but differs in that the plate-receiving means is constructed by the grooves 54 of the support members 42. Radiation-absorbing plates 44 have no slot, as in the second embodiment. In the inner surfaces of the opposite support members 42, a plurality of grooves 54 are formed in parallel from the upper edge 42a of the support member 42 to the lower edge 42b. Therefore, the opposite edges 44c of each radiation-absorbing plate 44 are inserted and supported in the corresponding grooves 54 of the support members 42 through the upper edges 42a of the support members 42, and the parallel grid 40 is formed. The width of the groove 54 of the support member 42 is of such a dimension that the edge 44c of the radiation-absorbing plate 44 is press-fitted and supported. However, since the insertion is performed over a short distance, the frictional resistance at the time of insertion is low even if the groove 54 is not formed wide, and there is only a slight possibility that the radiation-absorbing plate 44 will bend. Because the structure of the radiation-absorbing plate 44 in the third embodiment is also simple, it can be easily manufactured and is inexpensive. In addition, as the groove 54 is formed over the overall length from the upper edge 42a of the support member 42 to the lower edge 42b, the two support members 42 can be made the same. In the third embodiment, the support member 42 is very strong because the groove 54 is not an opening penetrating the plate thickness of the support member 42. Therefore, the rigidity of the grid 40 is significantly increased and positioning accuracy of the radiation-absorbing plate 44 is enhanced. FIGS. 6 and 7 show a grid 60 constructed according to a fourth embodiment of the present invention. This fourth embodiment, as with the aforementioned embodiments, is a linear grid, but is different in that a focusing grid in which radiation-absorbing plates 64 incline toward a radiation source X (FIG. 7) is located at a predetermined position. As illustrated in FIGS. 6 and 7A, the plate-receiving means in the fourth embodiment is constructed by a plurality of slots 74, which extend by approximately half of the height h of a support member 62 in the directions that focus toward the radiation source X. Note that some of the slots 74 shown in FIGS. 6 and 7 are omitted in order to make understanding of the present invention easy, but there are actually a large number of slots 74. Since the radiation source X is usually positioned above the central portion of the grid 60, the opposite slots 74d of the support member 62 incline most so that they are directed toward the radiation source X. As shown in FIG. 7A, the slots 74 inside the opposite slots 74d gradually sequentially approach a right angle with respect to the upper edge 62a of the support member 62, and only the central slot 74c crosses the upper edge 62a at a right angle. The radiation-absorbing plate 64 has two slots 76 similar to those of the radiation-absorbing plate 4 of the first embodiment shown in FIG. 1. If the support members 62 and the radiation-absorbing plates 64 are assembled, the grid 60 is obtained as shown in FIG. 6. Since the radiation-absorbing plates 64 are disposed in the directions that focus at the radiation source X, some of the rays, transmitted through a subject (not shown) positioned between the radiation source X and the grid 60, are linearly incident on the grid 60 without being intercepted by the radiation-absorbing plates 64. These rays then reach a radiation detector (not shown) positioned under the grid 60, and form a transmitted image. As a result, so-called cutoff, which is normally caused by interception of the transmitted radiation performed by the radiation-absorbing plates 64, will not occur, and a variation in the transmittance is eliminated and an image of high image quality is obtained. As with the aforementioned embodiments, the two support members 62 can be made the same. FIG. 8 shows a cross grid 80 constructed according to a fifth embodiment of the present invention The difference between the grid 80 of the fifth embodiment and the linear grids 1, 20, 40 and 60 of the aforementioned four embodiments is that radiation-absorbing plates 84 are each provided with a plurality of slots 96 disposed in parallel at predetermined intervals. Also, a plurality of thin support members (plates) 82, which are composed of the same material as the radiation-absorbing plate 84, i.e., a radiation-absorbing substance such as lead, tantalum and the like, are disposed in parallel in the slots 96 of the radiation-absorbing plates 84. With this disposition, the radiation-absorbing support members 82 and the radiation-absorbing plates 84 as a whole constitute the cross grid 80. The opposite ends of each radiation-absorbing support member 82 are connected to the opposite connecting members 86 through the opposite slots 96 of the radiation-absorbing support member 84. In addition, since the radiation-absorbing support members 82 and the radiation-absorbing plates 84 engage with one another, the self-supporting grid 80 with great structural strength is obtained. In cooperation with the radiation-absorbing plates 84, the radiation-absorbing support members 82 in the cross grid 80 absorb more scattered radiation than the linear grid, and consequently, the cross grid 80 achieves high image quality. However, cutoff will occur in the circumferential portion of the grid 80, because the radiation-absorbing support members 82 and the radiation-absorbing plates 84 in the fifth embodiment of FIG. 8 do not incline in the directions that focus at the radiation source X (FIG. 7). For this reason, radiation, transmitted through the subject and traveling linearly, is absorbed to some degree in the circumferential portion of the grid 80, so there is a possibility that the image quality will degrade. A grid 100 of a sixth embodiment improving the above disadvantage is shown in FIGS. 9 and 10. FIG. 10 shows a support member 102 and a radiation-absorbing plate 104 used in the grid 100. In the grid 100 of the sixth embodiment, slots 114 and 116, inclining in the directions that focus at a radiation source X (FIG. 10), are formed in the support member 102 and the radiation-absorbing plate 104, respectively. The slot 116 of the radiation-absorbing plate 104 is formed from one edge 104b of the radiation-absorbing plate 104 toward the other edge 104a by approximately half of the height h of the radiation-absorbing plate 104. With this construction, the support members 102 and the radiation-absorbing plates 104 engage with one another, whereby the cross grid 100 is formed as shown in FIG. 9. As with the fifth embodiment, it is desirable that the support members 102 intervening between the opposite support members 102 be thin. The height of the slot 114 of the support member 102 is approximately half of the height h of the support member 102, as in FIG. 7A. Since the intervening support members 102, as with the fifth embodiment, consist of a radiation-absorbing substance, rays scattered at the subject (not shown) are absorbed by the cross grid 100. In addition, the rays, transmitted through the subject and traveling linearly, arrive at a detector (not shown) without being intercepted by the cross grid 100, i.e., without giving rise to cutoff. Therefore, in the cross grid 100 of this sixth embodiment, the transmittance is enhanced and the scattered radiation are effectively reduced. Thus, a high quality transmitted image is obtained over the entire surface of the grid 100. FIG. 11 shows a grid 120 of a seventh embodiment of the present invention. The seventh embodiment differs from the aforementioned embodiments in that the plate-receiving means provided in the support members 122 are constructed by elongated holes 134. The support members 122 are connected at the opposite ends to the connecting members 126 and are formed into the shape of a frame as a whole, as with the first embodiment. In each support member 122, a plurality of vertical elongated holes 134 (i.e., plate-receiving means) are formed at predetermined intervals along the longitudinal direction of the support member 122. Rectangular radiation-absorbing plates 124 are inserted into these elongated holes 134, and the end portions 125 of each radiation-absorbing plate 124 penetrate the elongated holes 134 and project from the holes 134. After the radiation-absorbing plates 124 have been inserted into the elongated holes 134, movement of the radiation-absorbing plates 124 in the vertical direction perpendicular to the longitudinal direction is regulated and therefore there is no possibility that the radiation-absorbing plates 124 will slide in the vertical direction. In this way, the radiation-absorbing plates 124 are supported in parallel by the support members 122, whereby the grid 120 is constructed. In this condition the radiation-absorbing plates 124 may be fixed to the support members 122 by adhesion or the like. However, in the case where there is deformation, such as deflection, wrinkles and the like, in the radiation-absorbing plates 124, there is a need to correct the plate deformation before fixation and make the radiation-absorbing plates 124 flat. The method of correcting plate deformation will be described with reference to FIG. 12. As shown in FIG. 12A, the end portions of two metal wires 131 are passed through holes 126 formed in the end portions 125 of a radiation-absorbing plate 124a and are tied in loop form. Then, the radiation-absorbing plate 124a is pulled in the opposite directions by the two metal wires 131, whereby deformation, such as wrinkles and the like, is corrected. This correcting operation is performed after the radiation-absorbing plates 124a have been inserted into the support members 122, and the same applies to radiation-absorbing plates 124b, 124c to be described later. A frame-shaped jig 133 (only the part of which is shown in FIG. 12A) is disposed to surround the circumference of the grid 120, and the other end of the metal wire 131 which stretches each radiation-absorbing plate 124a is wound and fixed to this jig 133. Next, the radiation-absorbing plates 124a thus stretched are fixed to the support members 122 by adhesion or the like. In addition, instead of the metal wire 131, a rod (not shown) may be inserted into the hole 125 and the other end of this rod fixed to the jig 133 by an appropriate method. In the case of the radiation-absorbing plate 124b shown in FIG. 12B, cutouts 128 are formed in the opposite end portions 125 of the radiation-absorbing plate 124b, respectively. The end portions of the aforementioned wires 131 are wound around these cutouts 128 and tied in the form of a loop. The operation thereafter is the same as the case of FIG. 12A. In the case where the metal wires 131 are not used, irregularities 130 on the surfaces of both end portions 125 of the radiation-absorbing plate 124c may be clamped by a tool 135 such as cutting pliers and pulled in the opposite directions, as shown in FIG. 12C. The irregularities 130 are formed by embossing and prevent the tool 135 from slipping when clamped by the tool 135. When the tool 135 is not used, the aforementioned jig 133 is not used. In addition, the irregularities 130 may be formed by notching. Note that while the method of correcting plate deformation has been described in the case of the elongated holes 134, plate deformation can also be corrected for the slots 14, 34 (FIGS. 1 and 2) and the grooves 54 (FIG. 4) in the same manner. For instance, for the slots 14 shown in FIG. 1, the radiation-absorbing plates 4 are inserted into the support members 2, as in the elongated holes 134. After insertion, the end portions of each radiation-absorbing plate 4 protruding from 64 the slots 14 are pulled, and after deformation in each radiation-absorbing plate 4 has been corrected, the radiation-absorbing plates 4 are glued to the support members 2. This method can also be used in the cross grid 80 (FIG. 8) in which the radiation-absorbing support members 82 and the radiation-absorbing plates 84 are disposed in the form of a lattice. In this case, deformation in all the radiation-absorbing support members 82 and radiation-absorbing plates 84 can be corrected by pulling them vertically and horizontally, i.e., in 4 directions. Thereafter, they may likewise be fixed by adhesion. In the grooves 54 shown in FIG. 4, each radiation-absorbing plate 44 is pulled to a length equal to the space between the support members 42 plus two groove depths, and then the radiation-absorbing plates 44 are connected to the grooves 54 by adhesion. When the radiation-absorbing plate 44 is longer than the aforementioned length, it may be cut to coincide with that length. Thereafter, the radiation-absorbing plates 44 are likewise glued to the support members 42. FIG. 13 shows a grid 140 that is capable of keeping radiation-absorbing plates 124 stretched, after the grid has been constructed. Note that a description is made by applying the same reference numerals to the same components. As illustrated in FIG. 13A, two compression coil springs (hereinafter referred to simply as springs) (elastic bodies) 144 are interposed between both end portions of two support members 142 supporting a large number of radiation-absorbing plates 124 in parallel. As the springs 144 pull support members 142 in the opposite directions, the radiation-absorbing plates 124 fixed to the support members 142 are stretched and their flatness is ensured. The springs 144 are inserted onto shafts (not shown) or into a cylindrical member (not shown), whereby the shape is maintained. Instead of the springs 144, other elastic bodies, for example, synthetic resin material with elasticity, such as polyurethane, may be employed. In a grid 160 shown in FIG. 13B, springs 164 for urging support members 162 are provided on both sides of a pair of fixed or unmovable portions 166. The fixed portions 166 are disposed at the opposite end portions of the support members 162 and are coupled with a base 168, which is part of the grid 160, or are formed integrally with the base 168. The fixed portions 166 are disposed approximately midway between the two support members 162. This can make the length of the springs 164 shorter and prevent the springs 164 from being deflected horizontally. FIG. 14 shows a grid 180 that is an eighth embodiment of the present invention, in which stretched radiation-absorbing plates 184 are fixed by use of surface plates consisting of carbon, i.e., a ceiling plate 186 and a bottom plate 188. First, the radiation-absorbing plates 184 are fixed to the support members 182 by an adhesive agent 185, or protrusions 187, etc. Then, the ceiling plate 186 and the bottom plate 188 are disposed to interpose the radiation-absorbing plates 184 therebetween and are glued to the radiation-absorbing plates 184 by adhesion or the like. The ceiling plate 186 and the bottom plate 188 are slightly smaller in outside dimensions than a frame 192, constructed by the support members 182 and connecting members 190. The ceiling plate 186 and the bottom plate 188, therefore, can easily be inserted into the frame 192 and glued to the radiation-absorbing plates 184. In this way, fixing of the radiation-absorbing plates 184 can be performed even more reliably and therefore the rigidity of the entire grid and the structural strength of the frame 192 are enhanced. In this case, the support members 182 with slots are removable, since the ceiling plate 186, the bottom plate 188, and the radiation-absorbing plates 184 are fixed. In addition, in the case where the ceiling plate 186 and the bottom plate 188 are glued and fixed to the circumferential edges 194 of the frame 192 instead of being inserted into the frame 192, the radiation-absorbing plates 184 are not glued to the ceiling plate 186 and the bottom plate 188, but can maintain the entire rigidity. Furthermore, the radiation-absorbing plates 184 can be held in position, as they are protected from external influence. In the case of using the ceiling plate 186 and the bottom plate 188 in this manner, the radiation-absorbing plates 184 can be fixed by various methods. For instance, another embodiment of the grid 180 is illustrated in FIG. 15. In the case of this grid 180, the bottom plate 188 is glued to the radiation-absorbing plates 184, while the ceiling plate 186 is glued to the support members 182, i.e., the upper edge of the frame 192. In this case, the bottom plate 188 can also be glued to the frame 192, because it is located inside the frame 192. With this construction, straightness in the radiation-absorbing plates 184 is ensured and the rigidity of the frame 192 can be maintained. Conversely, the ceiling plate 186 may be inserted into the frame 192 and glued to the radiation-absorbing plates 184, and the bottom plate 188 may be glued to the lower edge 194 of the frame 192, away from the radiation-absorbing plates 184. Similarly, the same effect is obtainable. In the former case, i.e., in the case where the ceiling plate 186 and the bottom plate 188 are glued to the radiation-absorbing plates 184, grooves may be formed at positions on the inner surfaces of the ceiling plate and bottom plate 186 and 188 which correspond to the radiation-absorbing plates 184. In this case, adhesion and positioning of the radiation-absorbing plates 184 can be performed reliably by inserting the radiation-absorbing plates 184 into the grooves. In addition, in the latter case, i.e., in the case where the ceiling plate 186 and the bottom plate 188 are not glued to the radiation-absorbing plates 184, grooves or stepped portions may likewise be formed at positions on the ceiling plate and bottom plate 186 and 188 which correspond to the support members 182 and the connecting members 190. In this case, positioning of the frame 192 can be formed reliably and these components become difficult to deform. Illustrated in FIG. 16 is a grid 180a of still another embodiment. Although the radiation-absorbing plates used in this embodiment are the same as the aforementioned radiation-absorbing plates 184, they are mounted on the bottom plate 188 so that they incline toward a source of radiation (not shown). For example, the radiation-absorbing plates 184 are inclined by use of support members 112a in which the elongated holes 134 shown in FIG. 11 are arranged to incline toward the radiation source. Then, the inclined radiation-absorbing plates 184 are glued and fixed to the bottom plate 188. Notice that in FIG. 16, only one of the two support members 122a is shown. Thereafter, if the support members 122a are removed, the grid 180a is obtained as shown. In this case, the radiation-absorbing plates 184 are kept inclined by the bottom plate 188 alone, because they are not glued to the ceiling plate 186. conversely, as another variation, the radiation-absorbing plates 184 may be glued and fixed to the ceiling plate 184, and the bottom plate 188 and the support members 122a may be removed. In the case where the support members 122a are finally made unnecessary in this manner, the grid 180a can be reduced in size and becomes easy to handle. When the radiation-absorbing plates 184 are great in width, i.e., height, the effect of removing the support member 122a becomes much greater because the support members 122a becomes greater in height and weight. While the present invention has been described with reference to the preferred embodiments thereof, the invention is not limited to the details given herein, but may be modified within the scope of the appended claims. |
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claims | 1. A debris trap for catching debris falling through a fuel bundle orifice in a nuclear reactor, the debris trap comprising:a shaft including a pivot actuator at one end thereof;a debris capture tray pivotally attached to an opposite end of the shaft;a pivot shaft coupled to the pivot actuator and the debris capture tray and displaceable by the pivot actuator relative to the shaft; anda pivot mechanism connected between the pivot shaft and the debris capture tray, the pivot mechanism being actuated by the pivot actuator,wherein the debris capture tray includes a tray cavity, wherein the pivot actuator is operable to displace the pivot shaft relative to the shaft to pivot the debris capture tray about a pivot joint between an insertion position and a capture position, wherein in the insertion position, a lengthwise axis of the debris capture tray is substantially in alignment with the shaft, and wherein in the insertion position, the debris capture tray in its entirety extends beyond the pivot joint in substantial alignment with the shaft. 2. A debris trap according to claim 1, wherein a length of the shaft is sufficient to extend from a refueling bridge of a nuclear reactor to a bottom of the fuel bundle. 3. A debris trap according to claim 1, wherein the tray cavity of the debris capture tray is conical. 4. A debris trap for catching debris falling through a fuel bundle orifice in a nuclear reactor, the debris trap comprising:a shaft including a pivot actuator at one end thereof;a debris capture tray pivotally attached to an opposite end of the shaft;a pivot mechanism connected between the debris capture tray and the shaft, the pivot mechanism being actuated by the pivot actuator;a pivot shaft slidably coupled with the shaft and connected to the pivot actuator, wherein the pivot mechanism comprises a tray bracket secured to the debris capture tray, and a link connected between the tray bracket and the pivot shaft,wherein the debris capture tray includes a tray cavity, and wherein the pivot actuator is operable to pivot the debris capture tray between an insertion position and a capture position; anda shaft connector connected between the shaft and the debris capture tray, the shaft connector being pivotally connected to the debris capture tray via a pivot joint,wherein relative movement of the pivot shaft to the shaft causes the debris capture tray to pivot between the insertion position and the capture position, wherein in the insertion position, a lengthwise axis of the debris capture tray is oriented substantially in alignment with the shaft, and wherein in the insertion position, the debris capture tray in its entirety extends beyond the pivot joint in substantial alignment with the shaft. 5. A debris trap according to claim 1, further comprising a vacuum port in the debris capture tray and attachable to a vacuum source. 6. A debris trap for catching debris falling through a fuel bundle orifice in a nuclear reactor, the debris trap comprising:a shaft;a debris capture tray attached to the shaft, wherein the debris capture tray includes a conical tray cavity;a pivot actuator at one end of the shaft, wherein the debris capture tray is pivotally attached to an opposite end of the shaft about a pivot joint;a pivot mechanism connected between the debris capture tray and the shaft, the pivot mechanism being actuated by the pivot actuator, wherein the pivot actuator is operable to pivot the debris capture tray via the pivot mechanism between an insertion position and a capture position; anda vacuum port in the debris capture tray and attachable to a vacuum source,wherein in the insertion position, a lengthwise axis of the debris capture tray is substantially in alignment with the shaft, and wherein in the insertion position, the debris capture tray in its entirety extends beyond the pivot joint in substantial alignment with the shaft. |
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abstract | Pressurizer for a pressurized water nuclear power station, comprising an outer casing which delimits an inner space; a duct (11) which extends beneath the casing and which is capable of being tapped from the coolant system of the nuclear power station; a tap (18) which places the inner space of the casing in communication with the duct (11), this tap (18) being welded to the duct (11) by means of a weld seam (32); a sleeve (42) for protecting the weld seam (32), which sleeve is arranged inside the tap (18) and which has a lower peripheral edge (46) which is engaged in the duct (11), the sleeve (42) defining with the tap (18) and the duct (11) an annular space (74) which is capable of being filled with the primary liquid;wherein the annular space (74) is open along at least a portion of the lower peripheral edge (46) of the sleeve (42) and opens inside the duct (11). |
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summary | ||
055919839 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows a part of a radiation therapy unit 2 of conventional design, in which a multileaf collimator 4 constructed in accordance with the principles of the invention is used. The radiation therapy unit 2 comprises a gantry 6 which can be swiveled around a horizontal axis of rotation 8 in the course of a therapeutic treatment. Collimator 4 is fastened to a projection 5 of gantry 6. Projection 5 includes a linear accelerator (not shown) for generating a radiation beam 10 which is emitted from a central axis of the therapy unit 2 which is coincident with a central axis of collimator 4. Either electron radiation or photon radiation (gamma radiation) can be used for therapy. During treatment, beam 10 is trained on a zone 12 of a patient 13 to be treated and which lies in the isocenter of the gantry rotation. The rotational axis 8 of the gantry 6, the rotational axis 14 of a treatment table 16 and the beam axis 10 all intersect at the isocenter. As illustrated in FIGS. 2 and 3, the multiple layer collimator 4 of the present invention comprises two identical layers, an upper layer 20 and a lower layer 22, of pairs of opposed multiple leaves. As shown in the side view of FIG. 3, the top layer 20 comprises a middle section having a plurality of relatively narrow leaves 24 positioned in a side-by-side relationship, which is flanked on its left side by a relatively wide trimmer leaf 26 and on its right side by a relatively wide end leaf 28. The construction of bottom layer 22 is a mirror image of layer 20 and therefor common reference numbers are used for leaves 26 and 28, however, since in the middle section of layers 20 and 22 the narrow leaves are physically overlapping, the narrow leaves of layer 22 are referred by reference number 30. As shown in FIG. 2, wherein top layer 20 is shown in solid lines, bottom layer 22 is shown in dashed lines and a support frame 32 is shown in thick solid lines, frame 32 supports each of leaves 24 and 28 of top layer 20 and each of leaves 30 and 28 of lower layer 22 in a paired opposed relationship, so that they are independently movable in their longitudinal dimension into and out of beam axis 10 (the Y direction shown in FIG. 2), thereby allowing a user to create an infinite number of irregular and rectangular treatment fields 34 through which the radiation beam will pass. The size of the treatment field is limited in the X direction to the distance W between the inner edges of the opposed pairs of leaves 28 of layers 20 and 22 when fully closed, and in the Y direction to the distance L between the narrow leaves 24 or 30 when fully open. Thus, in general the maximum size field is a rectangle of dimension W.times.L. Note, as shown in FIG. 2, and as will be described in detail later, trimmer leaf 26 extends across the full width L of support frame 32, and is movable in the X direction for defining the ends of rectangular treatment fields. Also note that in FIG. 2 the leftmost leaves 24 and 30 are shown fully closed, and the second leftmost leaf 24 is shown partially open. The operation of leaves 24 and 28 of layer 20 (or leaves 30 and 28 of layer 22) for creating a treatment field is as is conventional in prior art single layer multileaf collimator arrangements. However, as more clearly shown in FIG. 3, for the multiple layer multileaf arrangement constructed in accordance with the principles of the invention, the leaves of each layer are not required to be manufactured to tolerances as close as those in the prior art so as to maintain a radiation blocking fit therebetween. Multileaf collimator arrangements constructed in accordance with the present invention are permitted a relatively wide interleaf spacing "b", which allows a relatively large manufacturing tolerance for the leaves and the assembly of each layer of the collimator. This significantly reduces the manufacturing cost of the leaves, as well as their completed assembly cost. Additionally, the increased interleaf spacing reduces the interleaf friction, thereby solving the previously noted positioning/repositioning problem. In order to prevent radiation leakage from between the leaves from reaching the patient, in accordance with the principles of the invention, lower layer 22 is positioned laterally offset from the axis of beam 10 as compared with the position of layer 20, so that underneath each space "b" between adjacent leaves 24 of layer 20 their is a leaf 30 of lower layer 22. Of course, the thickness (in the beam direction) of each leaf is designed to have sufficient density to safely block the treatment beam. An additional advantage of the present invention is that due to the offsetting relationship between the upper and lower leaves, the spatial resolution available for creating the edges of the treatment fields is increased from a (the width of a narrow leaf) to (a+b)/2. In accordance with another aspect of the invention, although leaves 24, 30, and 28 move in their longitudinal direction and only extend across one-half of treatment field 34, as previously noted, trimmer leaves 26 extend across the full width of support frame 32 (as more clearly shown at the left side of FIG. 2 for leaf 26 of layer 20), and are movable in their lateral, as opposed to their longitudinal, dimension. As such, they are positionable into and out of the axis of beam 10 in the X direction shown in FIG. 2. Thus, trimmer leaves 26 of layers 20 and 22 can be used to create left and right edges, respectively, of a rectangular treatment field 34 in the event that it is desired to create a rectangular treatment field having a width greater than the combined lateral dimension of all the adjacent narrow leaves 24 (or 30) and the end leaves 28. When using a trimmer leaf 26 to define one end of a rectangular treatment field, since any leaves adjacent leaf 26 will prevent its movement in the X direction, only those leaves 24,28 or 30,28 needed to define the field end opposing trimmer leaf 26 are fully extended, and all of the remaining leaves in that layer are fully retracted. The opposed leaves in the other of the layers (which are either above or below) can be used to define the length L of the treatment field. For creating a small rectangular field, a necessary number of adjacent narrow leaves 24 and 30 between the wide leaves 28 would be retracted, and the remainder of the leaves 24, 30 and 28 would be fully extended. For a creating a larger rectangular field, the narrow and wide leaves 8, 30, and 28 would be fully retracted, and the trimmer leaves 26 can be moved outward, thereby enlarging the X dimension to as much as W'=W+2a' (where a' is the width of a trimmer leaf 26). FIG. 4 illustrates in greater detail a portion of the top view of FIG. 2. As functionally show therein, frame 32 includes a plurality of motors 40 mounted thereon which are used in a conventional manner to individually position the leaves 24, 30 and 28 of the collimator into and out of the radiation beam for controllably defining the treatment field. One example of drive means (not shown) suitable for this is an individually driven worm gear for individually engaging a toothed track or floating nut mounted on each leaf. A similar arrangement can be used for driving the trimmer leaves 26. Details of one such prior art leaf driving means are provided in U.S. Pat. No. 5,160,847, issued Nov. 3, 1992 to Leavitt, et al. It is noted that due to the modular nature of a multi-leaf collimator constructed in accordance with the principles of the present invention, the manufacturing and assembly cost of the present design is significantly reduced as compared with prior art designs. That is, collimators in accordance with the present design are constructed from four identical leaf array jaws (and two end trimmers); i.e., two sets of wide leaves 28 (one set in each of layers 20 and 6) and two sets of narrow leaves (one set of leaves 24 in layer 20 and one set of leaves 30 in layer 6). The jaws of each of these sets can be manufactured and tested independently and later assembled together, thereby providing the above-noted cost reductions. Additionally, each part may be serviced and replaced, if necessary, separately, thereby reducing potential maintenance costs. Thus, what has been shown and described herein is a novel radiation treatment system that both overcomes problems inherent in the prior art and improves the functionality of multileaf collimators. Changes, modifications, variations and other uses and applications of the subject invention will become apparent to those skilled in the art after considering this specification and its accompanying drawings, which disclose preferred embodiments thereof. For example, in an alternative embodiment the trimmer leaves could be positioned in a separate, third leaf layer, for example one nearer to the source of the radiation beam. In this alternative embodiment, trimmer leaves 26 in layer 20 and 22 could be replaced with leaves similar in structure and operation to leaves 28. Additionally, various design changes can be made to the shape of the individual leaves, etc. All such changes, modifications, variations and other uses and applications which do not depart from the spirit and scope of the invention are deemed to be covered by this patent, which is limited only by the claims which follow. |
claims | 1. A method of early sensing of water leakage, through chemical concentration monitoring, in a nuclear reactor system using a liquid metal and molten salt, the method, implemented by a computer, comprising:measuring an electrical conductivity of a first channel and a second channel of a steam generator or a heat exchanger, the first channel including an outlet 1 and an outlet 2 and the second channel including an inlet;performing at least one of a dilution correction and a temperature compensation with respect to the measured electrical conductivity;calculating a first difference value between a corrected electrical conductivity of the outlet 1 and a corrected electrical conductivity of the inlet based on a method using a Wheatstone Bridge circuit;calculating a second difference value between a corrected electrical conductivity of the outlet 2 and the electrical conductivity of the inlet based on the method using the Wheatstone Bridge circuit;calculating an identification value associated with the water leakage in the steam generator or the heat exchanger by summing a first comparison value and a second comparison value, the first comparison value being outputted by comparing a predetermined threshold value with the calculated first difference value and the second comparison value being outputted by comparing the predetermined threshold value with the calculated second difference value;attempting to sense water leakage by comparing the calculated identification value and a reference value; anddetermining that the water leakage is sensed when the identification value is greater than the reference value, in which case a leakage alarm is raised; anddetermining that no water leakage is sensed when the identification value is less than or equal to the reference value,wherein the first comparison value is one of 0, 1 and the first difference value, and the second comparison value is one of 0, 1 and the second difference value. 2. The method of claim 1, wherein:the first channel comprises an m (m being a natural number) number of outlets,the second channel comprises an n (n being a natural number) number of inlets, andthe measuring of the electrical conductivity comprises measuring an electrical conductivity of each of the m number of outlets and the n number of inlets. 3. The method of claim 2, wherein the calculating of the difference value comprises:computing a difference value between an electrical conductivity of a single random outlet of the m number of outlets and an electrical conductivity of a single random inlet of the n number of inlet. 4. The method of claim 1:wherein the performing of the dilution correction or the temperature compensation comprises:correcting the measured electrical conductivity to be consistent with a standard temperature or a standard dilution which are used for determining when water has leaked. 5. The method of claim 1, wherein the steam generator or the heat exchanger included in the nuclear reactor system using the liquid metal and the molten salt which is any one of a liquid metal reactor, a sodium cooled fast reactor, a nuclear transmutation, a pressurized water reactor, and a heavy water reactor. |
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description | The present application is a Continuation of U.S. patent application Ser. No. 14/981,606, entitled “Fission Reaction Control in a Molten Salt Reactor”, issued on Oct. 8, 2019 as U.S. Pat. No. 10,438,705. U.S. patent application Ser. No. 14/981,606 claims priority to U.S. Provisional Patent Application No. 62/098,984, entitled “Molten Salt Nuclear Reactor and Method of Controlling the Same” and filed on Dec. 31, 2014, and U.S. Provisional Patent Application No. 62/234,889, entitled “Molten Chloride Fast Reactor and Fuel” and filed on Sep. 30, 2015, both of which are specifically incorporated herein for all that they disclose and teach. U.S. patent application Ser. No. 14/981,606 also claims priority to U.S. Provisional Patent Application No. 62/097,235, entitled “Targetry Coupled Separations” and filed on Dec. 29, 2014, which is specifically incorporated herein for all that it discloses and teaches. U.S. patent application Ser. No. 14/981,606 is also related to U.S. patent application Ser. No. 14/981,512, entitled “Molten Nuclear Fuel Salts and Related Systems and Methods” and filed on Dec. 28, 2015, which is specifically incorporated herein for all that it discloses and teaches. Molten salt reactors (MSRs) identify a class of nuclear fission reactors in which the fuel and coolant are in the form of a molten salt mixture containing solid or dissolved nuclear fuel, such as uranium or other fissionable elements. One class of MSR is a molten chloride fast reactor (MCFR), which uses a chloride-based fuel salt mixture that offers a high uranium/transuranic solubility to allow a more compact system design than other classes of MSRs. The design and operating parameters (e.g., compact design, low pressures, high temperatures, high power density) of an MCFR offer the potential for a cost-effective, globally-scalable solution to zero carbon energy. The described technology provides a molten salt reactor including a nuclear reactor core configured to contain a nuclear fission reaction fueled by a molten fuel salt. A molten fuel salt control system coupled to the nuclear reactor core is configured to remove a selected volume of the molten fuel salt from the nuclear reactor core to maintain a parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity. In one implementation, a molten salt reactor including a nuclear reactor core configured to sustain a nuclear fission reaction fueled by a molten fuel salt. The molten fuel salt control system includes a molten fuel salt exchange system that fluidically couples to the nuclear reactor core and is configured to exchange a selected volume of the molten fuel salt with a selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt. In another implementation, the molten fuel salt control system includes a volumetric displacement control system having one or more volumetric displacement bodies insertable into the nuclear reactor core. Each volumetric displacement body is configured to volumetrically displace a selected volume of molten fuel salt from the nuclear reactor core when inserted into the nuclear reactor core. In one implementation, the volumetric displacement body removes the selected volume of molten fuel salt from the nuclear reactor core, such as via a spill-over system. This Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. This Summary is not intended to identify key features or essential features of the claimed subject matter, nor is it intended to be used to limit the scope of the claimed subject matter. Other implementations are also described and recited herein. A molten salt fast reactor system employs a molten fuel salt in a fast neutron spectrum fission reactor. One type of molten salt reactor includes a fluoride salt as the carrier salt for the fissile fuel. Another type of molten salt reactor is a molten chloride fast reactor (MCFR) with a chloride salt as the carrier salt for the fissile fuel. Although the below description is written with respect to a molten salt chloride reactor, it is to be appreciated that the description, components, and methods described herein may be applicable to any molten fuel salt reactor. In an MCFR system, the fast neutron spectrum provided by chloride salts enables good breed-and-burn performance using the uranium-plutonium fuel cycle. The fast neutron spectrum also mitigates fission product poisoning to provide exceptional performance without online reprocessing and the attendant proliferation risks. During operation of an MCFR system, a molten fuel salt control system allows maintenance of fuel reactivity and/or fuel composition within desired operational bounds. In one implementation, the molten fuel salt control system includes a molten fuel salt exchange system that removes molten fuel salt from the nuclear reactor core, such as to maintain a parameter indicative of reactivity within a selected range of a nominal reactivity. In an additional or alternative implementation, a molten fuel salt control system includes a volumetric displacement control assembly to remove molten fuel salt from a nuclear reactor to control the fission reaction in the MCFR system (e.g., to maintain a parameter indicative of reactivity within a selected range of a nominal reactivity). The volumetric displacement control assembly may contain or be formed of non-neutron absorbing materials, neutron absorbing materials, and/or moderators. FIG. 1 schematically illustrates an example molten chloride fast reactor (MCFR) fuel cycle 100 with a MCFR parent reactor 102 and a MCFR daughter reactor 104. A particular classification of fast nuclear reactor, referred to as a “breed-and-burn” fast reactor, is a nuclear reactor capable of generating more fissile nuclear fuel than it consumes. For example, the neutron economy is high enough to breed more fissile nuclear fuel (e.g., plutonium-239) from fertile nuclear reactor fuel (e.g., uranium-238) than it burns. The “burning” is referred to as “burn-up” or “fuel utilization” and represents a measure of how much energy is extracted from the nuclear fuel. Higher burn-up typically reduces the amount of nuclear waste remaining after the nuclear fission reaction terminates. The example MCFR fuel cycle 100 is designed to use molten salt as a carrier for the fissile fuel in the reactor(s). In one example, this carrier salt may include one or more of a sodium salt, a chloride salt, a fluoride salt, or any other appropriate molten fluid to carry the fissile fuel through the reactor core. In one example, the molten chloride salt includes a ternary chloride fuel salt, although other chloride salts may be employed alternative to or in addition to the ternary chloride salt, including without limitation binary, ternary and quaternary chloride fuel salts of uranium and various fissionable materials. Various compositions have been explored through modelling and testing with a focus on high actinide concentrations and a resulting compact reactor size. For example, bred plutonium can exist as PuCl3 within the MCFR fuel cycle 100, and reduction-oxidation control can be maintained by adjusting the ratio of the oxidation states of the chloride salt used as fertile feed material. The example MCFR fuel cycle 100 enables an open breed-and-burn fuel cycle (e.g., exhibiting equilibrium, quasi-equilibrium, and/or non-equilibrium breed-and-burn behavior) employing a uranium-plutonium fuel cycle and resulting in significantly lower volumes of waste than a conventional open fuel cycle. Various implementations of the described technology provide for a molten fuel salt having a uranium tetrachloride (UCl4) content level above 5% by molar fraction, which aids in establishing high heavy metal content in the molten fuel salt (e.g., above 61% by weight). Uranium tetrachloride implementations may be accomplished through a mixture of UCl4 and uranium trichloride (UCL3) and/or an additional metal chloride (e.g., NaCl), such that desirable heavy metal content levels and melting temperatures (e.g., 330°-800° C.) are achieved. In one implementation, the MCFR parent reactor 102 includes a reactor vessel designed to hold the molten fuel salt as a reactor core section, one or more heat exchangers, control systems, etc. In one implementation, the reactor vessel may have a circular cross-section when cut along a vertical or Z-axis (i.e., yielding a circular cross-section in the XY plane), although other cross-sectional shapes are contemplated including without limitation ellipsoidal cross-sections and polygonal cross-sections. The MCFR parent reactor 102 is started with a loading into the reactor vessel an enriched fuel charge of initial molten fuel 106, such as using uranium-235 as a startup fuel, such as in the form of UCl4 and/or UCl3, along with a carrier salt (e.g., NaCl). In one example, the initial molten fuel 106 mixture contains enriched uranium at 12.5 w %, although other compositions may be employed. The initial molten fuel 106 circulates through a reactor core section in the reactor vessel of the MCFR parent reactor 102. In one implementation of the MCFR parent reactor 102, the molten fuel salt flows in an upward direction as it is heated by the fission reaction in the internal central reactor core section and downward around the internal periphery of the reactor vessel as it cools. It is to be appreciated that other additional or alternative molten fuel flows may also be employed (such as the primary coolant loop 313 of FIG. 3) that are designed to use the convention flows of a heated fluid and gravity, and/or assisted fluid flows through values, pumps, and the like. The constituent components of the molten fuel are well-mixed by the fast fuel circulation flow (e.g., one full circulation loop per second). In one implementation, one or more heat exchangers are positioned at the internal periphery of the reactor vessel to extract heat from the molten fuel flow, further cooling the downward flow, although heat exchangers may additionally or alternatively be positioned outside the reactor vessel. After initial startup, the MCFR parent reactor 102 reaches criticality in nuclear fission and the initial fissile fuel (e.g., enriched uranium) converts the fertile fuel to fissile fuel (breeds up). In the example of initial fissile fuel including enriched uranium, this fissile enriched uranium can breed depleted and/or natural uranium up to another fissile fuel, e.g., plutonium. This breed-and-burn cycle can breed enough plutonium-239 fissile nuclear fuel (e.g., in the form of PuCl3) to not only operate for decades but to also supply fuel for the MCFR daughter reactor 104 and other daughter and granddaughter reactors. Although other daughter and/or granddaughter reactors are not shown, it is to be appreciated that multiple reactors may be fed by the removed used fuel from the parent reactor 102 to one or more daughter reactors, which may then feed start up material to one or more granddaughter reactors, and on and on. In one implementation, the MCFR parent reactor 102 operates at 1000 MWt, which corresponds to a natural fuel circulation point design, although other operating outputs are achievable under different operating conditions, including forced fuel circulation to achieve higher thermal power levels. Other fertile fuels may include without limitation used nuclear fuel or thorium. As previously suggested, during normal operations, the MCFR parent reactor 102 breeds with sufficient efficiency to support a gradually increasing reactivity. The MCFR parent reactor 102 can be maintained at critical (e.g., barely critical) by removing molten fuel salt 108 (which may contain fissile fuel, fertile fuel, carrier salt, and or fission products) from the MCFR parent reactor 102 and replacing the removed molten fuel salt 108 with fertile fuel salt at a slow rate. In this manner, reactivity can be controlled by periodic removal of a volume of fully mixed molten fuel salt that circulates within the reactor vessel, depicted as removed molten fuel 108, and periodic replacement of the removed molten fuel 108 with depleted uranium chloride salt, depicted as fertile molten fuel feed 110. Other fertile fuels may include without limitation natural uranium, used nuclear fuel or thorium. In one implementation, the removed molten fuel 108 can be prepared for disposal as waste or it can be stored until sufficient material is available to start a new MCFR plant (e.g., the MCFR daughter reactor 104). In some cases, the removed molten fuel 108 can be used to start or initiate the MCFR daughter plant without reprocessing the removed molten fuel 108. In the latter scenario, it may be possible for nearly all actinides to move to the next MCFR plant for additional burn-up, thus avoiding proliferation risks associated with nuclear waste. Furthermore, the molten fuel salt exhibits a large negative temperature coefficient, very low excess reactivity, and passive decay heat removal, which combine to stabilize the fission reaction. The MCFR parent reactor 102 outputs certain waste components, illustrated as waste 112. In one implementation, the waste 112 does not contain actinides. Instead, the waste 112 includes gaseous and possibly volatile chloride fission products 114 and solid fission products 116, such as noble metals. The waste 112 can be captured through mechanical filtering and/or light gas sparging or any other appropriate technique to filter waste 112 from the molten fuel salt while the MCFR parent reactor 102 is in operation or the removed molten fuel 108 may be separated, treated, and re-introduced to the reactor. The mechanical filtering captures the solid fission products 116 and other particulates that are less soluble in the molten fuel salt. Similarly, noble fission product gasses are captured and allowed to decay in holding tanks. The filters containing the insoluble and longer lived solid fission products 116 form a portion of the waste stream. In one implementation, the waste 112 also reduces or eliminates criticality concerns as the waste 112 does not contain fissile isotopes separated from the fuel salt. The waste 112 components may include any one or more of transmutation products of the nuclear fission or any one of its decay products, chemical reaction products of the fuel salt with other fission products, corrosion products, etc. The elemental components of the waste 112 (also generally called fission products herein) are based upon the elemental components of the fuel salt, carrier salt, components and coatings, etc. For a molten chloride salt, fission products may include any one or more of noble gases and/or other gases including Iodine, Cesium, Strontium, halogens, tritium, noble and semi-noble metals in aerosol form, and the like. Solid waste fission products may include noble metals, semi-noble metals, alkali elements, alkali earth elements, rare earth elements, etc. and molecular combinations and thereof. FIG. 2 illustrates example MCFR reactivity control resulting from periodic molten fuel removal of molten fuel salt and replacement with a fertile molten fuel feed material, referred to as molten fuel salt exchange. Molten fuel salt exchange systems represent a type of molten fuel salt control system. The X-axis 200 represents time in effective full power years, and the Y-axis represents reactivity in terms of modeled k-effective 202. The parameter, k-effective, represents the multiplication factor, which indicates the total number of fission events during successive cycles of the fission chain reaction. Each drop in k-effective, such as drops 204, 206, and 208, represents a molten fuel salt exchange event. By replacing bred up or fissile molten fuel salt within the reactor with a fertile molten fuel feed, the MCFR can be maintained within a threshold level of a nominal reactivity. In some cases, the nominal reactivity is at an average near-zero excess reactivity operating condition with an upper threshold defining a maximum reactivity of that fuel cycle to trigger a molten fuel exchange, and the lower threshold defining the minimum reactivity to be achieved after the molten fuel exchange. The nominal, upper threshold, and/or lower threshold reactivity levels may stay the same or change over the lifetime of the MCFR based upon design, operation, and/or safety parameters. These parameters, which are indicative of reactivity, may include, without limitation, thermal energy desired to be generated by the reactor, safety levels, component design and lifetime constraints, maintenance requirements, etc. It should be understood that other reactivity control techniques may be employed in combination with molten fuel salt exchange, including without limitation use of a volumetric displacement assembly, neutron-absorbing control assemblies, etc. Furthermore, other molten salt reactors may employ a similar molten fuel exchange feature. As illustrated in FIG. 2, the periodic replacement of molten fuel salt with the fertile molten fuel feed may be used to limit reactivity and maintain ongoing breed-and-burn behavior within the reactor. Chronologically, the initial enriched fuel charge of molten fuel salt and fertile molten fuel salt can breed up, thereby increasing the reactivity within the reactor. After the reactor breads up, the periodic removal of fissile material acts to periodically (whether with uniform or non-uniform periods over time) reduce or control the reactivity of the reactor, returning the reactivity of the molten fuel salt back to an acceptable and pre-selected threshold level which may be a critical condition 210 (e.g., a barely critical condition) at each molten fuel salt exchange operation to approximate an average near-zero excess reactivity operating condition. This exchange operation can be repeated over time, resulting in the “saw tooth” reactivity curve, such as that shown in the MCFR reactivity control graph of FIG. 2. In some implementations, periodic exchange operations can allow the reactor to operate indefinitely without adding supplemental enriched fuel material. While molten fuel salt exchange is described as periodic, it should be understood that such exchange may be performed in a batch-wise, continuous, semi-continuous (e.g., drip) manner, etc. It is to be appreciated that increasing the frequency (which may be paired with smaller volumes of removed bred up fuel) can tighten the control or thresholds around the nominal reactivity to which the MCFR is controlled. FIG. 3 illustrates an example MCFR system 300 equipped with a molten fuel salt exchange assembly 301. In one implementation, the MCFR system 300 includes a reactor core section 302. The reactor core section 302 (which may also be referred to as a “reactor vessel”) includes a molten fuel salt input 304 and a molten fuel salt output 306. The molten fuel salt input 304 and the molten fuel salt output 306 are arranged such that, during operation, a flow of molten fuel salt 308 may form or include conical sections acting as converging and diverging nozzles, respectively. In this regard, the molten fuel salt 308 is fluidically transported through the volume of the reactor core section 302 from the molten fuel salt input 304 to the molten fuel salt output 306. The reactor core section 302 may take on any shape suitable for establishing criticality within the molten fuel salt 308 within the reactor core section 302. As shown in FIG. 3, the reactor core section 302 may be in the form of an elongated core section and may having a circular cross-section when cut along a vertical or Z-axis (i.e., a circular cross-section in the XY plane), although other cross-sectional shapes are contemplated including without limitation ellipsoidal cross-sections and polygonal cross-sections. The dimensions of the reactor core section 302 are selected such that criticality is achieved within the molten fuel salt 308 when flowing through the reactor core section 302. Criticality refers to a state of operation in which the nuclear fuel sustains a fission chain reaction, i.e., the rate of production of neutrons in the fuel is at least equal to rate at which neutrons are consumed (or lost). For example, in the case of an elongated core section, the length and cross-sectional area of the elongated core section may be selected in order to establish criticality within the reactor core section 302. It is noted that the specific dimensions necessary to establish criticality are at least a function of the type of fissile material, fertile material and/or carrier salt contained within the example MCFR system 300. As part of the reactor startup operation, the example MCFR system 300 is loaded with an initial enriched fuel charge of molten fuel salt. The reactor startup operation initiates a fission reaction with a breed-and-burn fuel cycle. The reactivity of the fission reaction of the example MCFR system 300 increases over time (see FIG. 2.). When reactivity fails to satisfy an acceptable reactivity condition (e.g., k-effective meets or exceeds a threshold, such as an upper threshold of 1.005, as indicated in the example shown in FIG. 2), also referred to as an “exchange condition” or a “control condition,” a selected volume of molten fuel salt 308 is removed from the reactor core section 302 and a selected volume and composition of fertile molten fuel feed 310 (e.g., a salt loaded with fertile material, such as depleted and/or natural uranium, used nuclear fuel or thorium.) is loaded into the reactor core section 302 in place of the removed molten fuel salt 308. The removed molten fuel salt 308 may include without limitation one or more of the following: lanthanides, other fission products, fissile material, fertile material and/or carrier salt. It is noted that a non-specific removal of lanthanides reduces the fission product inventory reactor core section 302 and the associated poisoning but also removes some of the fissile material from the reactor core section 302. In FIG. 3, the molten fuel salt exchange assembly 301 is operably coupled to the reactor core section 302 (or another portion of the example MCFR system 300) and is configured to periodically replace a selected volume of the molten fuel salt 308 with a selected volume and composition of the feed material 310. In this regard, the molten fuel salt exchange assembly 301 can control the reactivity and/or composition of the molten fuel salt 308 within the example MCFR system 300. The composition of the molten fuel salt 308 influences the oxidation states of the molten fuel salt 308. In one implementation, it is noted that the molten fuel salt 308 removed from the reactor core section 302 (shown as removed molten fuel 312) includes at least some fissile material, while the feed material 310 includes at least some fertile material. In another implementation, the removed molten fuel 312 includes one or more fission products. For example, the removed molten fuel 312 may include without limitation one or more lanthanides generated via fission within the molten fuel salt 308. In yet another implementation, the removed molten fuel 312 may include without limitation a mixture of fissionable material (e.g., UCl4), one or more fission products (e.g., one or more lanthanides and/or a carrier salt (e.g., NaCl). While molten fuel salt exchange is described as periodic, it should be understood that such exchange may be performed in a batch-wise, continuous, semi-continuous (e.g., drip) manner, etc. As the molten fuel salt 308 within the reactor core section 302 breeds up, converting fertile material to fissile material, the molten fuel salt exchange assembly 301 removes some of the molten fuel salt 308 as the removed molten fuel 312, which contains some volume of fissile material, and replaces the removed molten fuel 312 with the feed material 310, which includes at least some fertile material. In another implementation, the removed molten fuel 312 includes one or more fission products. Accordingly, the molten fuel salt exchange assembly 301 may act as a control mechanism on the reactivity within the example MCFR system 300 and may serve to return the reactivity of the molten fuel salt 308 to a critical condition (e.g., a barely critical condition). Thus, in one implementation, the molten fuel salt exchange assembly 301 of the example MCFR system 300 can allow operation of the example MCFR system 300 indefinitely without adding further enrichment. The molten fuel salt of the feed material 310 may include without limitation one or more fertile fuel salts, such as a salt containing at least one of depleted uranium, natural uranium, thorium, or used nuclear fuel. For example, in the case of a chloride-based fuel, one or more fertile fuel salts may include a chloride salt containing at least one of depleted uranium, natural uranium, thorium, or a used nuclear fuel. In some cases, the feed material 310 may contain fissile fuel, such as enriched uranium, which can be fed into the example MCFR system 300 at a rate or molecular volume less than the initial volume (e.g., 12.5%). This inclusion of fissile fuel in the feed fuel may be used throughout the lifetime of the example MCFR system 300, or alternatively, may be occasionally used to speed up or enrich the molten fuel salt within the example MCFR system 300 to enhance later removed fuel in future molten fuel salt exchanges for placement in daughter reactors. Furthermore, the molten fuel salt of the feed material 310 may include without limitation one or more fissile and/or fertile fuel salts mixed with a carrier salt, such as NaCl, although other carrier salts may be employed. The reactor core section 302 may be formed from any material suitable for use in molten salt nuclear reactors. For example, the bulk portion of the reactor core section 302 may be formed from one or more molybdenum alloys, one or more zirconium alloys (e.g., Zircaloy), one or more niobium alloys, one or more nickel alloys (e.g., Hastelloy N), ceramics, high temperature steel and/or other appropriate materials. The internal surface of the reactor core section 302 may be coated, plated or lined with one or more additional material in order to provide resistance to corrosion and/or radiation damage. In one example, the reactor core section 302 may be constructed wholly or substantially from a corrosion and/or radiation resistant material. In one implementation, the reactor core section 302 includes a primary coolant system 311, which may include one or more primary coolant loops 313 formed from piping 315. The primary coolant system 311 may include any primary coolant system suitable for implementation in a molten fuel salt context. In the illustrated implementation, the primary coolant system 311 circulates molten fuel salt 308 through one or more pipes 315 and/or fluid transfer assemblies of the one or more of the primary coolant loops 313 in order to transfer heat generated by the reactor core section 302 via one or more heat exchangers 354 to downstream thermally driven electrical generation devices and system or other heat storage and/or uses. It should be understood that an implementation of the example MCFR system 300 may include multiple parallel primary coolant loops (e.g., 2-5 parallel loops), each carrying a selected volume of the molten fuel salt inventory through the primary coolant system 311. In the implementation illustrated in FIG. 3, the molten fuel salt 308 is used as the primary coolant. Cooling is achieved by flowing molten fuel salt 308 heated by the ongoing chain reaction from the reactor core section 302, and flowing cooler molten fuel salt 308 into the reactor core section 302, at the rate maintaining the temperature of the reactor core section 302 within its operational range. In this implementation, the primary coolant system 311 is adapted to maintain the molten fuel salt 308 in a subcritical condition when outside of the reactor core section 302. It is further noted that, while not depicted in FIG. 3, the example MCFR system 300 may include any number of additional or intermediate heating/cooling systems and/or heat transfer circuits. Such additional heating/cooling systems may be provided for various purposes in addition to maintaining the reactor core section 302 within its operational temperature range. For example, a tertiary heating system may be provided for the reactor core section 302 and primary coolant system 311 to allow a cold reactor containing solidified fuel salt to be heated to an operational temperature in which the salt is molten and flowable. Other ancillary components may also be utilized in the primary coolant loop 313. Such ancillary components may be include one or more filters or drop out boxes for removing particulates that precipitate from the primary coolant during operation. To remove unwanted liquids from the primary coolant, the ancillary components may include any suitable liquid-liquid extraction system such as one or more co-current or counter-current mixer/settler stages, an ion exchange technology, or a gas absorption system. For gas removal, the ancillary components may include any suitable gas-liquid extraction technology such as a flash vaporization chamber, distillation system, or a gas stripper. Some additional implementations of ancillary components are discussed in greater detail below. It is noted herein that the utilization of various metal salts, such as metal chloride salts, in example MCFR system 300 may cause corrosion and/or radiation degradation over time. A variety of measures may be taken in order to mitigate the impact of corrosion and/or radiation degradation on the integrity of the various salt-facing components (e.g., reactor core section 302, primary coolant piping 315, heat exchanger 354 and the like) of the example MCFR system 300 that come into direct or indirect contact with the fuel salt or its radiation. In one implementation, the velocity of fuel flow through one or more components of the example MCFR system 300 is limited to a selected fuel salt velocity. For example, the one or more pumps 350 may drive the molten fuel salt 308 through the primary coolant loop 313 of the example MCFR system 300 at a selected fuel salt velocity. It is noted that in some instances a flow velocity below a certain level may have a detrimental impact on reactor performance, including the breeding process and reactor control. By way of non-limiting example, the total fuel salt inventory in the primary loop 313 (and other portions of the primary coolant system 311) may exceed desirable levels in the case of lower velocity limits since the cross-sectional area of the corresponding piping of the primary loop 313 scales upward as flow velocity is reduced in order to maintain adequate volumetric flow through the primary loop 313. As such, very low velocity limits (e.g., 1 m/s) result in large out-of-core volumes of fuel salt and can negatively impact the breeding process of the example MCFR system 300 and reactor control. In addition, a flow velocity above a certain level may detrimentally impact reactor performance and longevity due to erosion and/or corrosion of the internal surfaces of the primary loop 313 and/or reactor core section 302. As such, suitable operational fuel salt velocities may provide a balance between velocity limits required to minimize erosion/corrosion and velocity limits required to manage out-of-core fuel salt inventory. For example, in the case of a molten chloride fuel salt, the fuel salt velocity may be controlled from 2-20 m/s, such as, but not limited to, 7 m/s. In the example implementation illustrated in FIG. 3, the molten fuel salt exchange assembly 301 (a “molten fuel salt exchange system”) includes a used-fuel transfer unit 316 and a feed-fuel supply unit 314. In one implementation, the used-fuel transfer unit 316 includes a reservoir 318 for receiving and storing used-fuel 312 (e.g., burned fuel) from one or more portions of the MCFR system 300. As previously noted, the used-fuel 312 transferred to and stored in reservoir 318 represents a portion of the molten fuel salt mixture 308 previously used fission reaction within the MCFR system 300 and may include initial fissile material, bred up fissile material, fertile material and/or fission products, such as lanthanides. In another implementation, the used-fuel transfer unit 316 includes one or more fluid transfer elements for transferring molten fuel salt 308 from one or more portions of the MCFR system 300 to the reservoir 318. The used-fuel transfer unit 316 may include any fluid transfer element or device suitable for molten salt transfer. By way of non-limiting example, the used-fuel transfer unit 316 may include one or more pipes 320, one or more valves 322, one or more pumps (not shown) and the like. In another implementation, the used-fuel transfer unit 316 may transfer molten fuel salt 308 from any portion of the MCFR system 300 fluidically coupled to the reactor core section 302. By way of non-limiting example, the used-fuel transfer unit 316 may transfer molten fuel salt 308 from any portion of the primary circuit, such as, but not limited to, the reactor core section 302, the primary coolant system 311 (e.g., primary coolant loop 313) and the like, to the reservoir 318. In one implementation, the feed-fuel supply unit 314 includes a feed material source 317 for storing feed material 310 (e.g., mixture of fertile material and carrier salt). In one implementation, the feed material 310 may include a mixture of a selected fertile material (e.g., depleted uranium, natural uranium, used nuclear fuel, thorium and the like) and a carrier salt (e.g., NaCl) mixed such that the concentration of the molten feed material has a concentration of fertile material compatible with the molten fuel salt 308 remaining in the primary circuit of the MCFR system 300. In another implementation, the fertile material may include a fertile salt, such as uranium chloride, thorium chloride and the like. In this regard, the particular components of the feed material may be selected so as to at least approximately maintain or adjust the stoichiometry and/or chemistry (e.g., the chemical composition and/or reactivity) present in the molten fuel salt 308 contained within the MCFR system 300. In one implementation, the molten fuel salt exchange assembly 301 is capable of transferring the used fuel 312 out of the one or more portions of the MCFR system 300 while concurrently or sequentially transferring the feed material (e.g., which can include a mixture of a selected fertile material and a carrier salt) into the one or more portions of the MCFR system 300. In another implementation, the transfers may be performed synchronously or asynchronously. In another implementation, the feed-fuel supply unit 314 includes one or more fluid transfer elements for transferring feed material 310 from the feed material source 317 to one or more portions of the MCFR system 300. The feed-fuel supply unit 314 may include any fluid transfer element or device. By way of non-limiting example, the feed-fuel transfer unit 314 may include one or more pipes 324, one or more valves 326, one or more pumps (not shown) and the like. In another implementation, the feed-fuel supply unit 314 may transfer feed material 310 from the feed material source 317 to any portion of the MCFR system 300 fluidically coupled to the reactor core section 302. By way of non-limiting example, the feed-fuel supply unit 314 may transfer feed material 310 from the feed material source 317 to any portion of the primary circuit, such as, but not limited to, the reactor core section 302, the primary coolant system 311 (e.g., primary coolant loop 315) and the like. In one implementation, the feed material 310 is continuously transferred by the feed-fuel supply unit 314 to the reactor core section 302. By way of non-limiting example, the feed material 310 is continuously transferred at a selected flow rate by the feed-fuel supply unit 314 to the reactor core section 302. It is to be appreciated that the method of molten fuel salt removal may be continuous, semi-continuous, or in batches, and may be the same as or different from the method or timing of the fuel replacement. In another implementation, the feed material 310 is transferred batch-wise (i.e., in discrete volume units) by the feed-fuel supply unit 314 to the reactor core section 302. By way of example, the feed material 310 is transferred to the reactor core section 302 at a selected frequency (or at non-regular time intervals), a selected volume transfer size, and a selected composition for each batch transfer. The selected frequency, volume transfer size, and composition can vary over time. In another implementation, the feed material 310 is transferred by the feed-supply unit 314 to the reactor core section 302 in a semi-continuous matter. By way of non-limiting example, the feed material 310 is transferred to the reactor core section 302 via drip delivery. Such a semi-continuous feed of material (and simultaneous removal of utilized fuel from the reactor core section 302) may allow for limiting reactivity swings to less than 100 pcm (per cent mille or change in keff of less than 0.01). In another implementation, the feed-fuel supply unit 314 may include multiple feed material sources and associated fluid transfer elements (e.g., valves and piping) to allow an exchange of multiple variations of feed materials, so as to maintain the oxidation state of the reactor core section 302. For example, individual feed material sources, each containing one of UCl3, UCl4, or NaCl, may be used to selectively adjust the chemical composition of the molten fuel salt 308. See FIG. 8 for an explanation of the ternary phase diagram for UCl3—UCl4—NaCl (in mole %), wherein the oxidation states and stoichiometry of the molten fuel salt 308 may be controlled by adding selected volumes of UCl3, UCl4, or NaCl. In one implementation, the reservoir 318 includes one or more storage reservoirs suitable for receiving and storing the molten fuel salt from the reactor core section 302. The reservoir 318 may be sized and or designed to limit reactivity of the used fuel salt 312 to reduce or limit reactivity below criticality. The reservoir 318 may include any one or more of neutron absorbers, moderating materials, heat transfer devices, etc. to ensure any ongoing nuclear fission reactions within the used fuel salt 312 do not exceed some specified threshold of design and/or safety. In another implementation, the reservoir 318 may include a second generation (“daughter”) fast spectrum molten salt reactor. It should be understood that used-fuel removal and feed material supply are coordinated to maintain the reactivity and/or composition of the molten fuel salt 308 within the reactor core section 302. Accordingly, in one implementation, the molten fuel salt exchange assembly 301 includes an exchange controller 328. In one implementation, the exchange controller 328 may control one or more active fluid control elements in order to control the flow of feed material 310 from the feed material source 317 and the flow of used fuel salt 312 from the reactor core section 302 to the reservoir 318. In one implementation, the valves 322 and 326 are active valves controllable via electronic signal from the exchange controller 328. By way of non-limiting example, the valves 322 and 326 may include, but are not limited to, electronically-controlled two-way valves. In this regard, the exchange controller 328 may transmit a control signal to one of or both of the valves 322 and 326 (or other active flow control mechanisms) to control the flow of feed material 310 from the feed material source 317 and the flow of used fuel salt 312 from the reactor core section 302 to the reservoir 318. It is noted herein that the present implementation is not limited to the electronically controlled valves, as depicted in FIG. 3, which are provide merely for illustrative purposes. It is recognized herein that there are a number of flow control devices and configurations applicable to molten salt transfer that may be implemented to control the flow of feed material 310 from the feed material source 317 and the flow of used fuel salt 312 from the reactor core section 302 to the reservoir 318. In one implementation, the molten fuel salt exchange assembly 301 includes one or more reactivity parameter sensors 330, as discussed above. As previously noted, the one or more reactivity parameter sensors 330 may include any one or more sensors for measuring or monitoring one or more parameters indicative of reactivity or a change in reactivity of the fuel salt 308 of the reactor core section 302. The reactivity parameter sensor 330 may include, but is not limited to, any one or more capable of sensing and/or monitoring one or more of neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotropic concentration, burn-up and/or neutron spectrum. By way of non-limiting example, as discussed above, the one or more reactivity parameter sensors 330 may include, but are not limited to, a fission detector (e.g., micro-pocket fission detector), a neutron flux monitor (e.g., a fission chamber or an ion chamber), a neutron fluence sensor (e.g., an integrating diamond sensor), a fission product sensor (e.g., a gas detector, a β detector or a γ detector) or a fission product detector configured to measure a ratio of isotope types in a fission product gas. By way of another non-limiting example, as discussed above, the one or more reactivity parameter sensors 330 may include, but are not limited to, a temperature sensor, a pressure sensor or a power sensor (e.g., power range nuclear instrument). In another implementation, the reactivity is determined with one or more of the measured reactivity parameters (discussed above). In one implementation, the reactivity of the reactor core section 302 is determined by the controller 328 using a look-up table. In another implementation, the reactivity of the reactor core section 302 is determined by the controller 328 using one or more models. In another implementation, the reactivity parameter may be determined by an operator and entered directly into the controller 328 via an operator interface. It is noted herein that, while the reactivity parameter sensor 330 is depicted as being located within the fuel salt 308 in the reactor core section 302 of the MCFR system 300, this configuration is not a limitation on the present implementation, as noted previously herein. In one implementation, the determined reactivity parameter (whether measured or modeled), or a parameter indicative of reactivity, is compared with a predetermined reactivity threshold. If the determined reactivity parameter, or a parameter indicative of reactivity, satisfies a control condition (e.g., exceeds a high threshold or falls below a low threshold), a control system (e.g., a molten fuel salt exchange system, a volumetric displacement system, and/or other control systems) may be actuated to adjust the reactivity of the reactor core section 302 back into a nominal reactivity range. In another implementation, the one or more reactivity parameter sensors 330 are communicatively coupled to exchange controller 328. The one or more reactivity parameter sensors 330 are communicatively coupled to the exchange controller 328. For example, the one or more reactivity parameter sensors 330 may be communicatively coupled to the exchange controller 328 via a wireline connection (e.g., electrical cable or optical fiber) or wireless connection (e.g., RF transmission or optical transmission). In one implementation, the exchange controller 328 includes one or more processors and memory. In one implementation, the memory maintains one or more sets of program instructions configured to carry out one or more operational steps of the molten fuel salt exchange assembly 301. In one implementation, the one or more program instructions of the exchange controller 328, in response to the determined reactivity parameter exceeding the upper reactivity threshold, may cause the exchange controller 328 to direct the molten fuel salt exchange assembly 301 to replace a selected and determined volume of the molten fuel salt 308 of the MCFR system 300 with a selected and determined volume and composition of feed material 310 in order to control the reactivity and/or composition of the molten fuel salt 308 within the reactor core section 302. In another implementation, the one or more program instructions are configured to correlate a determined reactivity of the molten fuel salt 308 of the reactor core section 302 with a selected replacement volume and composition to compensate for the measured excess reactivity of the reactor core section 302, as well as other molten fuel salt compositional considerations. By way of non-limiting example, the reactivity parameter sensor 330 may acquire a reactivity parameter associated with the molten fuel salt 308 within the reactivity core section 302 (or another portion of the MCFR system 300). In settings where the reactivity parameter is indicative of a reactivity larger than a selected upper threshold, the exchange controller 328 may determine the replacement volume and composition to compensate for the elevated reactivity and direct the molten fuel salt exchange assembly 301 to remove the determined volume of molten fuel salt 308 from the reactor core section 302 (e.g., removed by used-fuel transfer unit 316) and replace the removed fuel salt with a substantially equal volume of feed material 310 (e.g., replaced by the feed-fuel supply unit 314). The amount of used-fuel 312 to be removed from the reactor core section 302 may be determined based upon the determined reactivity (measured or modeled) of the reactor core section 302, the determined amount of fissile and/or fertile fuel (measured or modeled), the waste (including fission products and other possible neutron absorbers) in the molten fuel salt 308, etc. The determined core reactivity, exceeding the upper threshold, may be compared to a lower threshold to determine an amount of change in reactivity needed to maintain the core reactivity within the bounds of the selected nominal reactivity. This amount of required change in reactivity can then be used with the existing fuel to determine the amount of used-fuel 312 to be removed to maintain core reactivity within the bounds of the upper and lower thresholds of reactivity. For example, the worth of a determined volume of removed used-fuel 312 may be determined (based upon the burn up of fissile fuel, the available fissile fuel, the remaining fertile fuel, and other components, e.g., fission products and carrier salts) of the existing fuel composition, and compared if sufficient to reduce reactivity of the reactor core to the lower threshold. Based upon the determined core reactivity after fuel removal, the worth, volume and components of the feed fuel may be determined to maintain reactivity for continued breeding of fuel, fuel volume requirements for the system, and maintain or adjust stoichiometry of the fuel overall. These determinations can be based upon computational models of reactivity and reactions, look up tables based on empirical and/or modeled data, etc. As noted above, any one or more (or combination of) the nominal reactivity level, the upper threshold reactivity level, and/or the lower reactivity threshold may dynamically change over the lifetime of the reactor for various operational and/or safety reasons. In another implementation, in settings where the frequency, volume, and composition of the replacement of molten fuel salt 308 with feed material 310 is predetermined, the exchange controller 328 may carry out a pre-determined scheduled exchange process via the control of active elements (e.g., valves 322 and 326, pumps and the like) of the molten fuel salt exchange assembly 301, based on time since last exchange cycle and/or determined reactivity of the reactor core section 302, as discussed herein. In alternative implementations, exchange may be performed at dynamically determined frequencies and/or volumes, based on results from reactivity parameter sensors 330 and other sensors, monitoring techniques, and computations. In one implementation, the selected volume and/or composition of feed-material added to the reactor core section 302 has a predetermined “worth” that can be adjusted up or down in volume and/or composition to match a target reactivity removal from a selected volume of used fuel removed from the reactor core section 302. In another implementation, the exchange controller 328 may direct the molten fuel salt exchange assembly 301 to perform a continuous exchange of molten fuel salt 308 with feed material 310, with feed material 310 being continuously fed to the reactor core section 302 and used-fuel 312 being continuously removed from the reactor core section 302 at a selected rate (e.g., 0.1-10 liters/day). In another implementation, the exchange controller 328 may direct the molten fuel salt exchange assembly 301 to perform semi-continuous exchange (e.g., drip) of molten fuel salt 308 with feed material 310. By way of example, the exchange controller 328 may direct the molten fuel salt exchange assembly 301 to perform drip exchange of molten fuel salt 308 with feed material 310, with feed material 310 being drip fed to the reactor core section 302 and discrete amounts of used-fuel 312 being simultaneously removed from the reactor core section 302. In another implementation, the exchange controller 328 may direct the molten fuel salt exchange assembly 301 to perform a batch-wise exchange of molten fuel salt 308 with feed material 310. By way of example, the exchange controller 328 may direct the molten fuel salt exchange assembly 301 to perform a series of discrete, or batch-wise, exchanges of molten fuel salt 308 with feed material 310, with discrete amounts of feed material 310 being fed to the reactor core section 302 and discrete amounts (equal in volume to the feed material) of used-fuel 310 being concurrently or sequentially removed from the reactor core section 302 at selected time intervals. By way of another non-limiting example, the exchange controller 328 may direct the molten fuel salt exchange assembly 301 to perform a single discrete, or batch-wise, exchange of molten fuel salt 308 with feed material 310, with a discrete amount of feed material 310 being fed to the reactor core section 302 and an equal amount of used-fuel 312 being concurrently or sequentially removed from the reactor core section 302 at the selected time. In another implementation, the MCFR system 300 includes one or more gas sparging units. The one or more gas sparging units are operably coupled to the reactor core section 302 and configured to continuously remove one or more waste gases (such as gaseous fission products like noble gases) from the molten fuel salt 308 of the reactor core section 302. By way of non-limiting example, the one or more gas sparging units include a helium and/or hydrogen gas sparging unit. It is noted that the noble gases include He, Ne, Ar, Kr and Xe. It is further noted that the gaseous waste absorbed in the molten fuel salt 308 may diffuse out of the molten fuel salt 308 of the reactor core section 302, allowing for them to be pumped out of the reactor via an associated gas pump. In another implementation, the reactor includes one or more filtering units. The one or more filtering units are operably coupled to the reactor core section 302 and configured to continuously remove one or more solid waste components, e.g., solid fission products such as noble and/or semi-noble metals or other particulate waste. By way of non-limiting example, the one or more filtering units may include one or more filters located in a bypass flow of the reactor core section 302 arranged to collect the one or more components of the solid waste, which precipitate and/or plate (depending on the design geometry) out of the molten fuel salt 308. It is noted that the noble and semi-noble metals include Nb, Mo, Tc, Ru, Rh, Pd, Ag, Sb and Te. In another implementation, the primary coolant system 311 includes one or more pumps 350. For example, one or more pumps 350 may be fluidically coupled to the primary coolant system 311 such that the one or more pumps 350 drive the molten fuel salt 308 through the primary coolant/reactor core section circuit. The one or more pumps 350 may include any coolant/fuel pump applicable to molten fuel salt 308. For example, the one or more fluid pumps 350 may include, but are not limited to, one or more mechanical pumps fluidically coupled to the primary coolant loop 313. By way of another example, the one or more fluid pumps 350 may include, but are not limited to, one or more electromagnetic (EM) and/or mechanical pumps fluidically coupled to the primary coolant loop 313. In another implementation, the MCFR system 300 includes a secondary coolant system 352 thermally coupled to the primary coolant system 311 via one or more heat exchangers 354. The secondary coolant system 352 may include one or more secondary coolant loops 356 formed from pipes 358. The secondary coolant system 352 may include any secondary coolant system arrangement suitable for implementation in a molten fuel salt context. The secondary coolant system 352 may circulate a secondary coolant through one or more pipes 358 and/or fluid transfer assemblies of the one or more secondary coolant loops 356 in order to transfer heat generated by the reactor core section 302 and received via the primary heat exchanger 354 to downstream thermally driven electrical generation devices and systems. For purposes of simplicity, a single secondary coolant loop 360 is depicted in FIG. 3. It is recognized herein, however, that the secondary coolant system 352 may include multiple parallel secondary coolant loops (e.g., 2-5 parallel loops), each carrying a selected portion of the secondary coolant through the secondary coolant circuit. It is noted that the secondary coolant may include any second coolant suitable for implementation in a molten fuel salt context. By way of example, the secondary coolant may include, but is not limited to, liquid sodium. It is further noted that, while not depicted in FIG. 3, the MCFR system 300 may include any number of additional or intermediate coolant systems and/or heat transfer circuits. It is noted herein that the utilization of various metal salts, such as metal chloride salts, in MCFR system 300 may cause corrosion and/or radiation degradation over time. A variety of measures may be taken in order to mitigate the impact of corrosion and/or radiation degradation on the integrity of the various salt-facing components (e.g., reactor core section 302, primary coolant piping 315, heat exchanger 354 and the like) of the MCFR system 300. In one implementation, using a noble metal as a cladding for various salt-facing components can mitigate the impact of corrosion of such components. In one implementation, the use of molybdenum cladding on the sodium-exposed surfaces can mitigate the impact of corrosion on such surfaces. In another implementation, the molten fuel salt may be maintained (e.g., via molten fuel salt exchange) in a redox (chemical reduction oxidation) state that is less corrosive. Certain additives may also be employed to mitigate the corrosive impact of the molten fuel salt on such components. FIG. 4 illustrates a graph 400 of modeled keff values (curve 402) of a reactor core and the total percentage of burn up of heavy metal (HM) fuel (curve 404) over time for a molten salt reactor controlled by the periodic exchange of molten fuel salt of the reactor with a fertile fuel salt. As also noted with regard to FIG. 2, the periodic exchange of molten fuel salt of the reactor with a fertile fuel salt may be used to limit reactivity and maintain ongoing breed-and-burn behavior within the molten salt reactor. In another implementation, the molten fuel salt exchange assembly may feed the molten salt reactor with salt loaded with fertile material (e.g., depleted uranium) at a rate that matches the rate at which fissile material is burned by the molten salt reactor, as discussed with regard to FIG. 5. Alternatively, the fertile material may be added at a different rate and/or time than the fissile fuel is removed. FIG. 5 illustrates a graph 500 of keff (curve 502) versus time for a modeled molten salt reactor with a depleted uranium feed provided at a rate that matches the reactor burn rate. It is noted that, in this implementation, the exchange assembly does not or need not specifically target lanthanides for removal from the molten salt reactor but rather removes them via bulk volume removal of the molten fuel salt within the molten salt reactor. The removed material may include without limitation one or more of the following: lanthanides, other fission products, fissile material, fertile material and/or the carrier salt. As shown in FIG. 5, the molten salt reactor breeds up and reaches a peak in keff of approximately 1.03 at around 10-15 years. The molten salt reactor thereafter experiences a loss in reactivity as the actinide inventory, including fissile material, falls while the fission product inventories increase. It is noted that such a configuration may operate for over 20 years and burn greater than 36% of the heavy metal fuel initially loaded into the reactor and later fed to the molten salt reactor during the molten salt reactor's lifetime. Example keff ranges that may be employed can include without limitation 1.0 as a low threshold and 1.035 as a high threshold, defining an example nominal reactivity range. Another example of keff can include without limitation 1.001 as a low threshold and 1.005 as a high threshold, defining another example nominal reactivity range. Yet another example nominal reactivity range may extend from just over 1.0 to about 1.01. Other nominal ranges and thresholds may be employed. Furthermore, other control systems may be employed, including without limitation control rods or control drums, moderators, etc. FIG. 6 illustrates a graph 600 depicting keff as function of time for a molten salt reactor with no addition of feed material and no removal of lanthanides. Curve 602 depicts keff for the case where waste fission products, such as noble gases and noble/semi-noble metals, are removed from the reactor core section 302. In such a scenario, calculations indicate that 30% burn-up may be achieved, with a lifetime of approximately 9 years. Curve 604 depicts keff as a function of time for the cases where nothing is removed from the reactor core section 302. In such a scenario, calculations indicate that a 10% burn-up may be achieved, with a lifetime of approximately 3 years. FIG. 7 illustrates an alternative example MCFR system 700 equipped with a molten fuel salt exchange assembly 701. The primary coolant system is configured such that a primary coolant 740 includes the molten fuel salt that circulates within the reactor vessel 742 of the reactor core section 702 (e.g., main vessel core). In this regard, the molten fuel salt does not flow out of the reactor core section 702 as part of the primary coolant circuit but rather the molten fuel salt is flowed as the primary coolant through the reactor core section 702. It is noted that in this implementation, the MCFR system 700 may include one or more heat exchangers 746 in the primary coolant circuit for the reactor core section 702, such that the molten fuel salt flows as the primary coolant 740 through the one or more heat exchangers 746, through the reactor core section 702, does not flow out of the reactor core section 702, and back through the one or more heat exchangers 746, as part of the primary coolant circuit. As such, heat from the reactor core section 702 is transferred from the molten fuel salt via one or more heat exchangers 746 to a secondary coolant system (not shown). In FIG. 7, the molten fuel salt exchange assembly 701 is operably coupled to the reactor core section 702 (or another portion of the example MCFR system 700) and is configured to periodically replace a selected volume of the molten fuel salt 708 with a selected volume and composition of the feed material 710. In this regard, the molten fuel salt exchange assembly 701 can control the reactivity and/or composition of the molten fuel salt 708 within the example MCFR system 700. In one implementation, it is noted that the molten fuel salt 708 removed from the reactor core section 702 (shown as removed molten fuel 712 in a reservoir 718) includes at least some fissile material, while the feed material 710 includes at least some fertile material. In another implementation, the removed molten fuel 712 includes waste that can include one or more fission products. For example, the removed molten fuel 712 may include without limitation one or more lanthanides generated via fission within the molten fuel salt 708. In yet another implementation, the removed molten fuel 712 may include without limitation a mixture of fissionable material (e.g., UCl4), one or more fission products (e.g., one or more lanthanides and/or a carrier salt (e.g., NaCl). While molten fuel salt exchange is described as periodic, it should be understood that such exchange may be performed in a batch-wise, continuous, or semi-continuous (e.g., drip) manner and may be periodic, sporadic or vary in timing from one fuel exchange to the next. In the example implementation illustrated in FIG. 7, the molten fuel salt exchange assembly 701 (a “molten fuel salt exchange system”) includes a used-fuel transfer unit 716 and a feed-fuel supply unit 714. The molten fuel salt exchange assembly 701 may include the same or similar elements and operate the same or in a similar manner as the molten fuel salt exchange assembly 301 of FIG. 3, although alternative structures and operations may also be employed. As shown in FIG. 7, an exchange controller 728 may control one or more active fluid control elements in order to control the flow of feed material 710 from the feed material source 717 and the flow of used fuel salt 712 from the reactor core section 702 to the reservoir 718. As the molten fuel salt 708 within the reactor core section 702 breeds up, converting fertile material to fissile material, the molten fuel salt exchange assembly 701 removes some of the molten fuel salt 708 as the removed molten fuel 712 in a feed material source 717, and replaces the removed molten fuel 712 with the feed material 710, which includes at least some fertile material. In another implementation, the removed molten fuel 712 includes one or more fission products. Accordingly, the molten fuel salt exchange assembly 701, removing not only fissile fuel but also lanthanides and other neutron absorbers, may act as a control mechanism on the reactivity and lifetime extender of the molten fuel salt 708 within the example MCFR system 700. The control advantage of the fuel exchange may serve to return the reactivity of the molten fuel salt 708 (monitored by a reactivity sensor 730 as discussed above with reference to reactivity sensor 330 of FIG. 3) to a critical condition (e.g., a barely critical condition) and may also increase the effectiveness of the reactor by removing neutron absorbers and/or modifiers. Thus, in one implementation, the molten fuel salt exchange assembly 701 of the example MCFR system 700 can allow operation of the example MCFR system 700 indefinitely without adding further enrichment. It should be understood that molten fuel salt exchange may occur during operation of the nuclear reactor and/or during maintenance shut-down periods. The molten fuel salt of the feed material 710 may include without limitation one or more fertile fuel salts, such as a salt containing at least one of depleted uranium, natural uranium, thorium, or used nuclear fuel. For example, in the case of a chloride-based fuel, one or more fertile fuel salts may include a chloride salt containing at least one of depleted uranium, natural uranium, thorium, or a used nuclear fuel. Furthermore, the molten fuel salt of the feed material 710 may include without limitation one or more fertile fuel salts mixed with a carrier salt, such as NaCl, although other carrier salts may be employed. FIG. 8 illustrates an example ternary phase diagram 800 for UCl3—UCl4—NaCl (in mole %). In one implementation, an MCFR system, as modelled, uses a salt mixture composed of various sodium chloride and uranium chloride components. One example of such compositions may include one more components of NaCl, UCl3, and/or UCl4, as shown in the ternary phase diagram 800 of FIG. 8. The shaded region 802 shows the extent of a 500° C. melting point envelope. Multiple fuel salt compositions have been considered and have been shown to be capable of net breed and burn behavior. Selection of the final composition depends on a variety of factors including oxidation state/corrosion, solubility, viscosity and reactor size. Modelling has investigated different specific salts in the ternary diagram 800 with melting points suitable for use in the MCFR implementations, including without limitation 82UCl4-18UCl3, 17UCl3-71UCl4-12NaCl, and 50UCl4-50NaCl. Results of the modelling indicate that such fuel salt implementations will sustain breed and burn behavior and could be used in reactor implementations described herein. As mentioned, the ternary phase diagram 800 shows the expected melting temperature for any mixture of UCl3—UCl4—NaCl. Of particular interest are mixtures having a melting point less than about 500° C., which mixtures are illustrated in the shaded region 802 of the ternary phase diagram 800. The eutectic point 804 has a melt temperature of 338° C. and a composition of 17UCl3-40.5UCl4-42.5NaCl (i.e., 17 mol % UCL3, 40.5 mol % UCL4 and 42.5 mol % NaCl). The shaded region 802 indicates a melting point envelope of 500° C. Moving to the far-right of this shaded region 802 provides an example implementation 806, 17UCl3-71UCl4-12NaCl, but it should be understood that many possible compositions exist within the melting point envelope of the shaded area 802 as various fuel salt mixtures having a melting point below 500° C. Furthermore, if the melting temperature limit is slightly extended to 508° C., a composition of 34UCl3-66NaCl provides an option that is free of UCl4. Likewise, the ternary diagram 800 allows a range of specific UCl3-UCl4-NaCl fuel salt implementations to be identified for any given melting point limit between about 800° C. and 338° C. For example, ternary salts with melting points between 300-550° C., 338-500° C., and 338-450° C. may be easily identified. Example methods of detecting composition changes may include without limitation: 1) measurements of redox (chemical reduction oxidation) 2) online glow discharge mass spectrometry of a sample 3) reactivity changes in the core 4) offline sample analysis including GDMS (glass discharge mass spectroscopy) 5) gamma spectroscopy The specific composition of the mixture may include any formulation including two or more of UCl4, UCl3 or NaCl, such that the resulting uranium content level and melting temperature achieve desired levels. By way of non-limiting example, the specific composition may be selected so that the corresponding melting temperature falls between 330 and 800° C. By way of another non-limiting example, the specific composition may be selected so that the overall uranium content level is at or above 61% by weight. In addition to selecting the overall uranium content level the fuel composition may also be determined to meet a selected amount of fissile uranium (as opposed to fertile). For example, the specific composition of the molten fuel salt may be selected such that the U-235 content of the molten fuel salt is below 20%. The following discussion will identify particular implementations of interest, however the following discussion does not limit the scope of the invention as claimed to only the implementations described below, but rather, that any implementations identifiable from FIG. 8 are contemplated, as well as any implementations having different metal chlorides other than NaCl. Examples of additional, non-fissile metal chlorides include NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 and/or NdCl3. Liquid fuels have an inherent advantage over solid fuels in that the heat is “born” within the fuel coolant. A solid fuel may (1) conduct heat to the outer surface of the fuel element, (2) conduct heat through the cladding (including past a physical gap or through a bond material), (3) convect the heat from the cladding surface to the primary coolant, and (4) advect the heat out of the core. By comparison, a liquid fuel provides acceptable thermal transfer with step (4) and transport the fuel salt/primary coolant out of the core and to the primary heat exchanger. Additionally, the liquid salts under consideration have volumetric heat capacities that are nearly twice that of liquid sodium at similar temperatures. Another key advantage provided by a molten fuel salt is the strong negative temperature coefficient—hot salt is less reactive than cold salt. As a result, transients that result in overheating (e.g., loss of heat sink) are limited in severity by the expansion of the fuel salt. For example, in a molten chloride fast reactor (MCFR), as the selected chloride salt composition is heated from 600 to 800° C., its density drops by more than 12%, providing a negative reactivity feedback that is approximately 50× stronger than that provided by the Doppler effect. Fuel salts with similar ratios of the number of mono-chlorides, tri-chlorides, and tetra-chlorides behave similarly. The oxidation state within reactor core section of a molten chloride fast reactor (MCFR), for example, may be defined as the ratio of the molecules grouped by the number of attached chlorine molecules. The oxidation state of the reactor core section can be controlled by exchanging a selected amount of fuel salt in the reactor core section with a similar amount of makeup salt or feed material, where the composition of the feed material is designed to bring the oxidation state of the reactor core section toward a target oxidation state. In one implementation, the feed material contains a mixture of a selected fertile material and a carrier salt. In one implementation, the fuel salt in the reactor core section is initially at an oxidation state that is mostly composed of mono-chlorides, tri-chlorides, and tetra-chlorides. This initial fuel salt composition (prior to removal a selected volume of the fuel salt and addition of feed material) is represented by the initial fuel salt vector (f), where the subscript x represents the number of chloride ions present in each molecule of the fuel salt. Molecules with 2, 5 and 6 chloride atoms can exist within the reactor core section in very small quantities, so they can be ignored—the bulk properties of the molten chloride fuel are dominated by the mono-chlorides, tri-chlorides, and tetra-chlorides (see Equation (1), which indicates a simplified fuel salt vector in which the molten chloride fuel is dominated by mono-chlorides (f1), tri-chlorides (f3), and tetra-chlorides (f4)). As such, if the target salt mixture is PbCl2—UCl3—UCl4 (or PuCl2—UCl3—UCl4), one would control on di-chlorides, tri-chlorides, and tetra-chlorides. Note: the fuel salt vector may be generalized to other chloride salts and fluoride salts. Accordingly, a similar control approach may be applied to fluoride salts, where the subscript x represents the number of fluoride ions in each molecule of the fuel salt. ( f 1 f 2 f 3 f 4 f 5 f 6 ) ∼ ( f 1 f 3 f 4 ) = ( f ) ( 1 ) As such, the initial fuel salt vector (f) may be represented by the simplified fuel salt vector given in Equation (1). Removal of a selected volume (r) of the initial fuel salt over a period of time (either as a large batch, a set or sequence of smaller batches, or a continuous or partially continuous stream) normalized to the amount of initial fuel salt present in the reactor at the start of that period of time (e.g., about 1% per year for a specific MCFR system) yields an adjusted fuel salt vector (f′), which is shown by Equation (2), representing the fuel salt remaining in the reactor after removal of a selected volume of the initial fuel salt. ( f 1 f 3 f 4 ) ∼ C * ( f 1 f 3 f 4 ) → ( f ′ 1 f ′ 3 f ′ 4 ) = ( f ′ ) ( 2 ) A target fuel salt composition within the reactor, represented by a target fuel salt vector (t), may be set to achieve a particular oxidation state and/or stoichiometry from the adjusted fuel salt composition (adjusted fuel salt composition (f′) by adding a selected volume and composition of feed material, which is represented by a feed fuel salt vector (m). This relationship is represented by Equations (3) and (4), where (r)˜C*(f).(f)−(r)=(f′) (3)(f′)+(m)=(t) (4) In an alternative notation, this relationship is represented by Equations (5) and (6). ( f 1 f 3 f 4 ) - ( r 1 r 3 r 4 ) = ( f ′ 1 f ′ 3 f ′ 4 ) ( 5 ) ( f ′ 1 f ′ 3 f ′ 4 ) + ( m 1 m 3 m 4 ) = ( t 1 t 3 t 4 ) ( 6 ) Given Equations (3)-(6), the volume and composition of the feed material to be added to the reactor to achieved the target oxidation state and/or stoichiometry may be determined (e.g., (m)). For each molecule type, the makeup fuel salt vector (mx) may be represented by Equation (7), where the subscript x represents the number of fluoride ions in each molecule of the fuel salt and C represents the normalized amount removed in a given period of time.(mx)=(tx)−(1−C)*(fx) (7) Nuclear fission reactors operate at zero or approximately zero excess reactivity to operate at a constant power. In addition to controlling the oxidation state of the molten fuel salt in the reactor, the reactivity of the described molten salt reactor implementations can be adjusted in situ by swapping fuel salt for a feed material. In a burner molten salt reactor, fissile material is burned so reactivity tends to decrease with time. As such, the feed material is designed to contain a significant quantity of high reactivity fuel salt rich in fissile material, such as enriched uranium or reprocessed transuranics. In a breeder molten salt reactor, fissile material is produced faster than it is consumed by the fission reaction, so the reactivity tends to increase with time. As such, the feed material is designed to contain low reactivity fuel salt that is rich in fertile material, such as natural uranium, depleted uranium, used nuclear fuel, or thorium. The rate at which feed material is introduced to the reactor core is selected to maintain the reactivity within certain design limits, such as nominal reactivity (e.g. keff equaling 1 or slightly greater than 1, an upper reactivity threshold, and/or a lower reactivity threshold). FIG. 9 illustrates example operations 900 for a molten fuel salt exchange process. A system provisioning operation 902 provides a molten chloride fast reactor (which is an example molten salt reactor) with a molten fuel salt exchange system. A monitoring operation 904 monitors for an exchange condition for the molten fuel salt. For example, one or more reactivity parameter sensors may monitor the reactivity within the molten chloride fast reactor, and/or chemical composition sensors, such as Raman spectroscopy may monitor the composition of the molten fuel salt within the molten chloride fast reactor. In an implementation, the monitoring may be performed in real-time using Raman spectroscopy. Raman spectroscopy provides information about molecular vibrations that can be used for sample identification and quantitation. The technique involves shining a monochromatic light source (i.e. laser) on a sample and detecting the scattered light. Some amount of fuel may be removed from the reactor core, such as in a side stream, and passed through a monitoring cell that includes a ‘window’ through with the spectroscopy can be performed. Examples of Raman windows materials are fused quartz, fused silica, sapphire, diamond, and some glasses. Any material may be used as long as it can meet the operational parameters of the reactor and monitoring system. An exchange condition may be set for monitored reactivity, composition, or some other operating parameter to trigger a molten fuel salt exchange event. If the exchange condition has not been satisfied, then a decision operation 906 returns processing to the monitoring operation 904. If the exchange condition has been satisfied, then the decision operation 906 progresses processing to a removal operation 908, which removes a selected volume of molten fuel salt from the molten chloride fast reactor. A replacement operation 910 replaces the removed volume of the molten fuel salt with a selected volume and/or composition of feed material into the molten chloride fast reactor. Processing returns to the monitoring operation 904. FIG. 10 illustrates a molten salt reactor 1000 equipped with a volumetric displacement element assembly 1002. Volumetric displacement systems represent a type of molten fuel salt control system. In one implementation, the volumetric displacement assembly 1002 is operably coupled to the reactor core section 1004 containing a molten fuel salt 1006. The volumetric displacement assembly 1002 is arranged so as to selectively displace a volume of the molten fuel salt 1006. In this regard, the volumetric displacement assembly 1002 may displace a volume of the fuel salt 108 in order to control reactivity within the molten fuel salt 1006. The volumetric displacement element assembly 1002 may control reactivity of the molten salt reactor 1000 by controlling the volume of molten fuel salt 1006, and thus the fissile material, displaced in the reactor core section 1004 (e.g., center region of the core section). By way of a non-limiting example, in settings where the reactor core section 1004 possesses excess reactivity, a sufficient volume (e.g., 0.1 to 10.0 m3) of molten fuel salt 1006 may be displaced by the volumetric displacement assembly 1002 such that the reactivity decreases to a lower reactivity threshold, such as critical or sub-critical levels. It should be appreciated that multiple volumetric displacement assemblies may be used in various configurations within the molten salt reactor 1000. In one implementation, the volumetric displacement assembly 1002 includes a volumetric displacement element 1010, an actuator 1012 and an actuator controller 206. In one implementation, the volumetric displacement element 1010 is formed from a non-neutron-absorbing material. In this regard, the volumetric displacement element 1010 controls reactivity in the molten salt reactor 1000 via the volumetric fluid displacement of the molten fuel salt 1006 (and fissile material) and not through a neutron absorption process. It is noted that the utilization of a non-neutron-absorbing material is particularly advantageous in the molten salt reactor 1000 as it avoids large impacts on reactivity, which may occur with the introduction of neutron-absorbing materials into the reactor core section 1004. A non-neutron-absorbing volumetric displacement element, which operates based on volumetric fluid displacement of the molten salt, may provide subtler reactivity control than neutron-absorbing control elements. It should be understood, however, that the volumetric displacement element 1010 (e.g., displacement rod) may be formed from any non-neutron absorbing material, although neutron absorbing and/or moderating materials may additionally or alternatively be employed in such elements. As such, the volumetric displacement element 1010 may alternatively include a neutron transparent material or a neutron reflector material. For example, the volumetric displacement element 1010 may be formed, but is not required to be formed, from zirconium, steel, iron, graphite, beryllium, molybdenum, lead, tungsten, boron, cadmium, one or more molybdenum alloys (e.g., TZM alloy), one or more tungsten alloys (e.g., tungsten carbide), one or more tantalum alloys, one or more niobium alloys, one or more rhenium alloys, one or more nickel alloys, silicon carbide and the like. In such implementations, the volumetric displacement element 1010 may limit reactivity through the volumetric fluid displacement of fuel and through the absorption of neutrons. In one implementation, the volumetric displacement element 1010 includes a rod 1016, as shown in FIG. 10. For example, the volumetric displacement element 1010 includes a solid rod or a hollow rod. It is noted herein that the displacement rod 1016 may take on any type of rod shape. For example, a displacement rod of the volumetric displacement assembly 1010 may take on a cylindrical shape, a square or rectangular prism shape, a triangular prism shape, a polygonal prism shape and the like. In another implementation, the volumetric displacement element 1010 may include a set of rods (not shown). For example, the set of rods may be arranged in an array or spoke pattern. In one implementation, the actuator 1012 is operably coupled to the volumetric displacement element 1010, such that the actuator 1012 may selectively translate the volumetric displacement element 1012. The actuator 1012 may include any actuation device. For example, the actuator 1012 may include, but is not limited to, a displacement rod drive mechanism. In one implementation, the actuator 1012 is configured to drive the volumetric displacement element 1010 bidirectionally. In this regard, the actuator 1012 may drive the volumetric displacement element 1010 into and/or out of the reactor core section 1004 as desired. In another implementation, the actuator 1012 is configured to stop driving the volumetric displacement element 1010 at one or more intermediate positions between a first stop position and a second stop position. In this regard, the actuator 1012 may translate the volumetric displacement element 1010 along a selected direction (e.g., axial direction) so as to insert a selected amount of the volumetric displacement element 1010 into the molten fuel salt 1006 of the reactor core section 1004. For example, in the case of a rod-shaped volumetric displacement element 1010, the actuator 1012 may insert a selected volume of the volumetric displacement element 1010 by controlling the length L of the rod-shaped volumetric displacement element 1010 inserted into the molten fuel salt 1006. It is noted that the volumetric displacement assembly 1002 may displace any amount of volume of the molten fuel salt 1006 within the reactor core section 1004 necessary to reduce the reactivity of the molten fuel salt 1006 within the reactor core section 1004 as desired. By way of non-limiting example, the volume of molten fuel salt 1006 within the reactor core section 1004 may range from 10 to 100 m3, depending on the particular fuel formulation and operation context of the molten salt reactor 1000. In this setting, a displacement volume of only a fraction of a cubic meter may supply sufficient volumetric salt displacement to significantly reduce reactivity within the reactor core section 1004 and, in some cases, shutdown the reactor. For example, in marginal control or non-shutdown operations, the displacement volume imparted by the volumetric displacement element 1010 may include, but is not limited to, a displacement volume between 0.1 to 10 m3. In one implementation, as shown in FIG. 10, the volumetric displacement assembly 1010 may insert the volumetric displacement element 1010 into a central region of the reactor core section 1004. In this regard, the actuator 1012 may translate the volumetric displacement element 1010 along the axial direction of the reactor core section 1004, as shown in FIG. 10. It is noted that given a rotationally symmetric core section, as that depicted in FIG. 10, the greatest reactivity worth associated with the volumetric displacement element 1010 may be realized by positioning the volumetric displacement element 1010 at the cross-sectional center of the reactor core section 1004. It is noted that a centered volumetric displacement element 1010 is not a limitation on the molten salt reactor 1000 of the present disclosure and is provided merely for illustrative purposes. Moreover, although displacement element 1010 is shown in FIG. 10 as a single element, it is to be appreciated that the displacement element may include a plurality of insertable elements, which may move into and out of the reactor core in tandem or may be moved and controlled individually to manage reactivity, fuel flow, local temperature, etc. In another implementation, the actuator controller 1010 is configured to selectively direct the actuator 1012 to insert a selected volume of the volumetric displacement element 1010 a selected distance into a volume of the molten fuel salt 1006 contained within the reactor core section 1004. For example, the actuator controller 1014 may direct the actuator 1012 to translate the volumetric displacement element 1010 such that the volumetric displacement element 1010 partially or entirely submerses in the molten fuel salt 1006. The actuator controller 1014 is communicatively coupled to the actuator 1012. For example, the actuator controller 1014 may be communicatively coupled to the actuator 1012 via a wireline connection (e.g., electrical cable or optical fiber) or wireless connection (e.g., RF transmission or optical transmission). In one implementation, the actuator controller 1012 includes an operator interface configured to receive volumetric displacement actuation instructions from an operator. In this regard, an operator may selectively direct the control the actuation state of the volumetric displacement element 1010. In another implementation, the actuation controller 1014 may automatically direct the actuation of the volumetric displacement element 1010 in response to one or more sensed or monitored parameters of the molten salt reactor 1000, as discussed below. In another implementation, the molten salt reactor 1000 includes a reactivity parameter sensor 1030. The reactivity parameter sensor 1030 includes any one or more sensors capable of measuring or monitoring one or more parameters indicative of reactivity or a change in reactivity of the molten fuel salt 1006 of the molten salt reactor 1000. For example, the reactivity parameter sensor 1030 may include, but is not limited to, any one or more sensors capable of sensing and/or monitoring one or more of neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotropic concentration, burn-up and/or neutron spectrum. In one implementation, the reactivity parameter sensor 1030 includes a fission detector. For example, the reactivity parameter sensor 1030 may include, but is not limited to, a micro-pocket fission detector. In another implementation, the reactivity parameter sensor 1030 includes a neutron flux monitor. For example, the reactivity parameter sensor 1030 may include, but is not limited to, a fission chamber or an ion chamber. In another implementation, the reactivity parameter sensor 1030 includes a neutron fluence sensor. For example, the reactivity parameter sensor 1030 may include, but is not limited to, an integrating diamond sensor. In another implementation, the reactivity parameter sensor 1030 includes a fission product sensor. For example, the reactivity parameter sensor 1030 may include, but is not limited to, a gas detector, a β detector or a γ detector. In another implementation, the reactivity parameter sensor 1030 includes a fission product detector configured to measure a ratio of isotope types in a fission product gas. In another implementation, the reactivity parameter sensor 1030 includes a temperature sensor. In another implementation, the reactivity parameter sensor 1030 includes a pressure sensor. In another example, the reactivity parameter sensor 1030 includes a power sensor. For example, the reactivity parameter sensor 1030 may include, but is not limited to, a power range nuclear instrument. In another implementation, the reactivity is determined with one or more of the measured reactivity parameters (discussed above). In one implementation, the reactivity of the reactor core section 1004 is determined by the actuator controller 1012 using a look-up table. For example, measured values for temperature, pressure, power level and the like may be used in conjunction with one or more look up tables to determine the reactivity of the reactor core section 1004. In another implementation, the reactivity of the reactor core section 1004 is determined by the actuator controller 1014 using one or more models. For example, the one or more models may include, but are not limited to, a neutronics modeling software package executed by the one or more processors of the actuator controller 1014. For instance, a suitable neutronics software package may include, but is not limited to, MCNP, CINDER, REBUS and the like. In another implementation, the reactivity parameter may be determined by an operator and entered directly into the actuator controller 1014 via an operator interface. It is noted herein that, while the reactivity parameter sensor 1030 is depicted as being located within the molten fuel salt 1006 in the reactor core section 1004 of the molten salt reactor 1000, this configuration is not a limitation on the present implementation and is provided merely for illustrative purposes. Rather, it is noted that one or more reactivity parameter sensors 1030 may be located at various positions of the molten salt reactor 1000 including, but not limited to, at a position within the reactor core section, at a position external to the reactor core section 1004 (e.g., at external surface of reactor core section 1004), in or along one or more pipes of a primary coolant system, in or near a primary heat exchanger, in or along one or more pipes of a secondary coolant system and the like. In another implementation, the one or more reactivity parameter sensors 1030 are communicatively coupled to actuator controller 1014. The one or more reactivity parameter sensors 1030 are communicatively coupled to the actuator controller 1014. For example, the one or more reactivity parameter sensors 1030 may be communicatively coupled to the actuator controller 1014 via a wireline connection (e.g., electrical cable or optical fiber) or wireless connection (e.g., RF transmission or optical transmission). In one implementation, the actuation controller 1014 may direct the actuator 1012 to adjust the position of the volumetric displacement element 1010 (and, thus, the reactivity of the molten fuel salt 1006) based on the measured reactivity parameter. In one implementation, the actuation controller 1014 includes one or more processing units and memory. In one implementation, the memory maintains one or more sets of program instructions configured to carry out one or more operational steps of the volumetric displacement assembly 1010. In one implementation, the one or more program instructions of the actuation controller 1014 may cause the actuator controller 1014 to direct the actuator 1012 to drive the volumetric displacement assembly 1010 into the reactor core section 1004 to displace a selected volume of the molten fuel salt 1006 within the reactor core section 1004. In another implementation, the one or more program instructions are configured to correlate a determined reactivity of the reactor core section 1004 with a displacement volume necessary to compensate for the measured reactivity of the reactor core section 1004. For example, as discussed above, the reactivity parameter sensor 1030 may acquire a reactivity parameter associated with the molten fuel salt 1006 within the reactivity core section 1004. In settings where the reactivity parameter is indicative of a reactivity larger than a selected tolerance level, the actuator controller 1014 may determine the displacement volume to compensate for the elevated reactivity and direct the actuator 1012 to insert enough of the volumetric displacement element 1010 to achieve at least this level of volumetric salt displacement. In another implementation, in settings where complete reactor shutdown is required, the actuator controller 1014 may direct the actuator 1012 to insert the entire volumetric displacement element 1010 into the reactor core section 1004 in order to achieve maximum volumetric salt displacement. FIG. 11 illustrates a molten salt reactor 1100 equipped with a volumetric displacement element assembly 1102 and a molten fuel salt spill-over system 1130 with a volumetric displacement element 1110 not submerged in molten fuel salt. In one implementation, the molten fuel salt spill-over system 1130 includes one or more fuel salt uptakes 1132 and one or more spill-over reservoirs 1134. It is noted that in some cases the volumetric displacement of the molten fuel salt 1106 by the volumetric displacement element 1110 may cause a rise in the fuel salt level above a desired level. In one implementation, the molten fuel salt spill-over system 1130 is configured to transport molten fuel salt 1106 that is displaced above the maximum tolerated fill level of the reactor core section 1104, as shown in FIG. 12. By way of non-limiting example, the fuel salt uptake 1132 may be placed approximately 10 cm above a nominal fuel salt level. In this regard, when the volumetric displacement element 1110 is engaged, it may, in some cases, cause the molten fuel salt level to rise above normal salt level. Molten salt that reaches the fuel salt uptake 1132 is then transported to the spill-over reservoir 1134. It should be appreciated that multiple volumetric displacement assemblies may be used in various configurations within the molten salt reactor 1100. FIG. 12 illustrates a molten salt reactor 1200 equipped with a volumetric displacement element assembly 1202 and a molten fuel salt spill-over system 1230 with a volumetric displacement element 1210 submerged in molten fuel salt. While the molten fuel salt spill-over system 1230 depicted of FIG. 12 is depicted in the context of the volumetric displacement element assembly 1202 and volumetric displacement element 1210, this is not a requirement on the molten fuel salt spill-over system 1230. In this regard, the molten fuel salt spill-over system 1230 of the present disclosure may be implemented in a context that does not include the volumetric displacement assembly 1202 and volumetric displacement element 1202. In one implementation, the molten fuel salt spill-over system 1230 may be implemented in order to account for thermal expansion of the molten fuel salt 1206. By way of non-limiting example, in the case where the fuel salt uptake 1232 is place at 10 cm above the normal salt level a mere 50° C. increase in temperature of the fuel salt 108 may cause the molten fuel salt 1206 to reach the fuel salt uptake 1232. By way of another non-limiting example, approximate increase of 200° C. in temperature of the molten fuel salt 1206 may cause the molten fuel salt 1206 to spill over through the fuel salt uptake 1232 and lead to 1-5 m3 of fuel salt to spill into one or more spill-over reservoirs 1234. Spilled-over fuel salt 1236 is shown in the one or more spill-over reservoirs 1234. It is recognized herein that the combination of very low excess reactivity and the strong thermal feedback of the molten fuel salt 1206 may allow for nearly passive operation. In this sense, use of the displacement element 1210 may be limited. As the demand on the turbine (not shown) of the nuclear reactor plant varies, the temperature(s) associated with the primary cooling loop will vary slightly. This, in turn, will vary the temperature of the molten fuel salt 1206. As a result, the molten fuel salt 1206 will obtain a new average temperature, and thus, density, causing the fluid level of the molten fuel salt 1206 to increase or decrease. By way of non-limiting example, in the event that demand for electricity increase, the steam of the turbine comes out at a reduced temperature. As a result, temperatures throughout the nuclear reactor system are reduced, causing the molten fuel salt 1206 to decrease in temperature and increase in density. This increase in density results in an increase in reactivity. In addition, the fluid level of the molten fuel salt 1206 is decreases, while increased reactivity causes the power of the molten salt reactor 1200 to increase, thereby meeting the increased demand on the turbine. In turn, increase in power causes the temperature of the molten fuel salt 1206 to increase and the fluid level of the molten fuel salt 1206 to return to (or near) its original level. It is further recognized that, in the event of a loss of heat sink or a turbine trip, temperatures throughout the molten salt reactor 1200 would increase. As a result of increased temperatures in the molten fuel salt 1206, the molten fuel salt 1206 would decrease in density, causing the molten fuel salt 1206 to become less reactive. The decrease in density would cause the fluid level to rise and, in some instances (e.g., +50° C. temperature rise) the fluid level of the molten fuel salt 1206 reach the level of the fuel salt uptake 208. Such a rise in fluid level may then cause some molten fuel salt 1206 to spill over into the one or more spill-over reservoirs 1234, which would serve to further reduce reactivity in the reactor core section 1204. As a result, the molten salt reactor 1200 may go into a sub-critical state and remain in that state, even upon cooling. In another implementation, the molten fuel salt spill-over system 1230 may include a return pathway (e.g., one or more pipes, one or more pumps and one or more valves), where fuel salt stored in the one or more spill-over reservoirs 1234 may be actively pumped out of the one or more spill-over reservoirs 1234 and back into the reactor core section 1204 in order to reestablish a critical state. In another implementation, the displacement element 1210 may be used to accelerate the above process as well as control or shape changes in reactivity/density/temperature during normal operation. It should also be understood that various structural modifications to the displacement element 1210 may be employed to enhance control performance and manage influence that molten fuel salt turbulence may have on the placement and stability of the displacement element 1210 within the reactor core section 1204. Such structural modifications may include without limitation different shapes, sizes, and numbers of displacement elements 1210, dynamic shape change features in displacement element 1210, baffles and/or nozzles in the displacement element 1210, and other flow-friendly features to the displacement element 1210. It should be appreciated that multiple volumetric displacement assemblies may be used in various configurations within the reactor core section 1204. FIG. 13 illustrates various example stages of a fuel displacement cycle 1300. In stage 1302, the displacement element 1301 includes a hollow or solid displacement rod 1303 inserted through rod inlet 1305 and a displacement body 1307 having a width w that is wider than both the displacement rod 1303 and the rod inlet 1305 and a height h that is less than the height y of the reactor core section 1311. As a result, the maximum volume of displacement can be vertically selected/located within the reactor core section 1311 by raising or lowering the displacement body 1307 to a desired height in the molten fuel salt 1309 within the reactor core section 1311. The dashed line 1320 indicates the molten fuel salt level when the displacement element has not yet been lowered into the molten fuel salt 1309. It should be understood that the displacement rod 1303 and/or the displacement body 1307 may be formed of or filled with various materials, including non-neutron absorbing materials and neutron absorbing materials. In stage 1302, the displacement element has been partially lowered into the molten fuel salt, resulting in a raising of the molten fuel salt level. The subsequent stages 1304, 1306, 1308, 1310, and 1312 show progressively lower insertions of the displacement body 1307 into the molten fuel salt 1309, resulting in increasingly higher levels of the molten fuel salt 1309, although such increasing levels of molten fuel salt 1309 may be mitigated by a spill-over system. Stage 1312 illustrates a fully immersed displacement body 1307. By displacing the volume of molten fuel salt 1309 at a particular location within the reactor core section, the reactivity within the reactor core section 1311 can be controlled. Even after the displacement body 1307 is fully immersed within the molten fuel salt 1309, the vertical location within the reactor core section 1311 can further influence the reactivity (e.g., the lower the displacement body 1307, the more negative influence on reactivity) in the illustrated implementations. See FIG. 14 and the associated discussion. It should be appreciated that multiple volumetric displacement assemblies may be used in various configurations within the reactor core section 1311. FIG. 14 illustrates two example stages 1402 and 1404 of a fuel displacement cycle 1400. In stage 1402, the displacement element 1401 includes a hollow or solid displacement rod 1403 and a displacement body 1407 inserted deep into molten fuel salt 1409 within a reactor core section 1411. In stage 1404, the displacement body 1407 inserted less deeply into the molten fuel salt 1409 within the reactor core section 1411. As a result, the maximum volume of displacement can be vertically selected/located within the reactor core section 1411 by raising or lowering the displacement body 1407 to a desired height in the molten fuel salt 1409 within the reactor core section 1411. It should be understood that the displacement rod 1403 and/or the displacement body 1407 may be formed of or filled with various materials, including non-neutron absorbing materials and neutron absorbing materials. Accordingly, in one implementation, the reactivity control may be characterized as more negative in the stage 1402 than in the stage 1404 because the displacement body 1407 is inserted more deeply into the reactor core section 1411, displaying more fuel at an input region of the reactor core section 1411, where the molten fuel salt 1409 first enters the active fission reaction region at each circulation cycle. It should be appreciated that multiple volumetric displacement assemblies may be used in various configurations within the reactor core section 1411. FIG. 15 illustrates example operations 1500 for a molten fuel salt displacement process. A system provisioning operation 1502 provides a molten chloride fast reactor (which is an example molten salt reactor) with a molten fuel salt exchange system. A monitoring operation 1504 monitors for a control condition for the molten fuel salt (e.g., k-effective meets or exceeds a threshold, such as 1.005). For example, one or more reactivity parameter sensors may monitor the reactivity within the molten chloride fast reactor. The control condition may be set for monitored reactivity or some other operating parameter to trigger a molten fuel salt displacement event. If the control condition has not been satisfied, then a decision operation 1506 returns processing to the monitoring operation 1504. If the control condition has been satisfied, then the decision operation 1506 progresses processing to an insertion operation 1508, which inserts a displacement body into molten fuel salt within a reactor core section. A positioning operation 1510 positions the displacement body into the molten fuel salt of the molten chloride fast reactor to remove a selected volume of molten fuel salt from the reactor core section to obtain desired reactivity parameters in the molten chloride fast reactor. Processing returns to the monitoring operation 1504. In one implementation, an example molten salt reactor includes a nuclear reactor core configured to contain a nuclear fission reaction fueled by a molten fuel salt. A molten fuel salt control system is coupled to the nuclear reactor core and is configured to remove a selected volume of the molten fuel salt from the nuclear reactor core to maintain a parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity. Another example molten salt reactor of any preceding reactor provides a molten fuel salt control system that includes a molten fuel salt exchange system fluidically coupled to the nuclear reactor core and configured to exchange a selected volume of the molten fuel salt with a selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that includes a feed-fuel supply unit configured to transfer the feed material into the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that a feed-fuel supply unit configured to transfer a selected volume of the feed material into the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that the molten fuel salt exchange system that includes a feed-fuel supply unit configured to transfer a selected composition of the feed material into the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that includes a used-fuel transfer unit configured to transfer the selected volume of the molten fuel salt as used-fuel from the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that is configured to transfer concurrently the selected volume of the molten fuel salt from the nuclear reactor core and the feed material into the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that controls reactivity of the nuclear fission reaction by exchanging the feed material with the selected volume of the molten fuel salt in the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system that controls composition of the molten fuel salt in the nuclear fission reaction by exchanging the feed material with the selected volume of the molten fuel salt in the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a fast spectrum fission reactor and the molten fuel salt includes a chloride salt. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system controls a composition of UCl3-UCl4-NaCl in the spectrum fission reaction by exchanging the feed material with the selected volume of the molten fuel salt in the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides a molten fuel salt exchange system is configured to exchange repeatedly a selected volume of the molten fuel salt with a selected volume of the feed material to maintain the parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity over time. Another example molten salt reactor of any preceding reactor further includes a reactivity parameter sensor positioned proximate the nuclear reactor core. The nuclear parameter sensor is configured to monitor one or more parameters indicative of reactivity of the nuclear reactor core. A controller communicatively couples to the reactivity parameter sensor to receive the one or more parameters indicative of reactivity of the nuclear reactor core. The controller is configured to control exchange of the selected volume of the molten fuel salt with the selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt based on the one or more parameters. Another example molten salt reactor of any preceding reactor provides the molten fuel salt control system to further include a volumetric displacement control system having one or more volumetric displacement assemblies insertable into the nuclear reactor core. Each volumetric displacement assembly is configured to volumetrically displace a selected volume molten fuel salt from the nuclear reactor core when inserted into the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides the molten fuel salt control system to further include a volumetric displacement control system having one or more volumetric displacement bodies insertable into the nuclear reactor core, each volumetric displacement body being configured to volumetrically displace a selected volume of molten fuel salt from the nuclear reactor core when inserted into the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides the molten fuel salt control system to further include a volumetric displacement control system having one or more volumetric displacement bodies insertable into the nuclear reactor core, each volumetric displacement body being configured to volumetrically displace a selected volume of molten fuel salt from the nuclear reactor core when inserted into the nuclear reactor core, the volumetric displacement control system further having molten fuel salt spill-over system configured to transport molten fuel salt that is displaced by the volumetric displacement body above a tolerated fill level of the nuclear reactor core. Another example molten salt reactor of any preceding reactor provides the molten fuel salt control system to further include a volumetric displacement control system having one or more volumetric displacement bodies insertable into the nuclear reactor core, each volumetric displacement body being configured to volumetrically displace a selected volume of molten fuel salt from the nuclear reactor core when inserted into the nuclear reactor core, the volumetric displacement control system being insertable at multiple insertion depths into the nuclear reactor core to maintain the parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity over time. Another molten salt nuclear reactor includes a nuclear reactor core configured to sustain a nuclear fission reaction fueled by a molten fuel salt and means for exchanging a selected volume of the molten fuel salt with a selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt. Another molten salt nuclear reactor includes a nuclear reactor core configured to sustain a nuclear fission reaction fueled by a molten fuel salt and means for removing a selected volume of the molten fuel salt from the nuclear reactor core to maintain a parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity. An example method includes sustaining a nuclear fission reaction fueled by a molten fuel salt within a nuclear reactor core and removing a selected volume of the molten fuel salt from the nuclear reactor core to maintain a parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity. Another example method of any preceding method further includes replacing the selected volume of the molten fuel salt with a selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt. Another example method of any preceding method wherein the replacing operation includes transferring the feed material into the nuclear reactor core. Another example method of any preceding method wherein the replacing operation includes transferring a selected volume of the feed material into the nuclear reactor core. Another example method of any preceding method wherein the replacing operation includes transferring a selected composition of the feed material into the nuclear reactor core. Another example method of any preceding method wherein the replacing operation includes controlling the reactivity of the nuclear reactor core based on the selected volume of the feed material. Another example method of any preceding method wherein the replacing operation includes controlling the composition of the molten fuel salt fueling the nuclear fission reaction within the nuclear reactor core based on the selected composition of the feed material. Another example method of any preceding method wherein the replacing operation includes controlling the composition of the UCl3-UCl4-NaCl fueling the nuclear fission reaction within the nuclear reactor core based on the selected composition of the feed material. Another example method of any preceding method wherein the method further includes monitoring satisfaction of an exchange condition by the molten fuel salt and controlling exchange of the selected volume of the molten fuel salt with the selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt responsive to satisfaction of the exchange condition. Another example method of any preceding method wherein the method further includes monitoring one or more reactivity parameters indicative of reactivity of the nuclear reactor core and controlling exchange of the selected volume of the molten fuel salt with the selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt based on the one or more reactivity parameters. Another example method of any preceding method wherein the method further includes monitoring one or more composition parameters indicative of composition of the molten fuel salt of the nuclear reactor core and controlling exchange of the selected volume of the molten fuel salt with the selected volume of a feed material containing a mixture of a selected fertile material and a carrier salt based on the one or more composition parameters. Another example method of any preceding method wherein the removing operation includes volumetrically displacing the selected volume molten fuel salt from the nuclear reactor core by inserting one or more volumetric displacement bodies into molten fuel salt within the nuclear reactor core. Another example method of any preceding method wherein the removing operation includes transporting the volumetrically displaced volume of molten fuel salt from the nuclear reactor core via a molten fuel salt spill-over system when the volumetrically displaced volume of molten fuel salt is displaced by the volumetric displacement body above a tolerated fill level of the nuclear reactor core. Another example method of any preceding method provides a method wherein each volumetric displacement body is configured to volumetrically displace a selected volume of molten fuel salt from the nuclear reactor core when inserted into the nuclear reactor core, the volumetric displacement control system being insertable at multiple insertion depths into the nuclear reactor core to maintain the parameter indicative of reactivity of the molten salt reactor within a selected range of nominal reactivity over time. An example fast spectrum molten salt nuclear reactor includes a reactor core section including a fuel input and a fuel output, the fuel input and the fuel output arranged to flow a molten chloride salt nuclear fuel through the reactor core section. The molten chloride salt nuclear fuel including a mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt, the mixture of UCl4 and at least one additional metal chloride salt having a UCl4 content greater than 5% by molar fraction. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the uranium concentration in the mixture of UCl4 and at least one additional metal chloride salt is greater than 61% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the additional uranium chloride salt including UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt has a composition of 82UCl4-18UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt has a composition of 17UCl3-71UCl4-12NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt has a composition of 50 UCl4-50NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the additional metal chloride including at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 or NdCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt has an additional metal chloride salt concentration at or below the precipitation concentration for the an additional metal chloride salt. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt having a melting temperature below a temperature of 800 degrees Celsius. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt having the selected melting temperature above a temperature of 330 degrees Celsius. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides breed-and-burn behavior established within the molten chloride salt nuclear fuel with a uranium-plutonium cycle. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the fuel input located on a first side of the reactor core section and the fuel output located on a second side of the reactor core section opposite to the fuel input. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides a protective layer disposed on at least one surface facing the molten chloride salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides that the at least one surface exposed to the molten chloride salt nuclear includes an internal surface of the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the protective layer that is substantially resistant to at least one of corrosion or radiation. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the protective layer including at least one of a refractory alloy, a nickel alloy, a refractory metal or silicon carbide. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a reflector assembly configured to reflect at least a portion of neutrons emanating from the reactor core section back to the molten chloride salt nuclear fuel within the reactor core section, the reflector assembly including a plurality of reflector modules, at least some of the reflector modules containing a liquid reflector material. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides at least one of the reflector modules formed from at least one of a molybdenum alloy, a nickel alloy or a carbide. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the liquid reflector material including at least one of liquid lead or liquid lead-bismuth. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a displacement assembly operably coupled to the reactor core section and configured to selectively displace a volume of the molten salt nuclear fuel in order to control reactivity within the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the displacement assembly configured to displace a volume of the molten salt nuclear fuel in order to reduce reactivity within the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the displacement assembly that includes a displacement element, an actuator operably coupled to the displacement element, and a controller. The controller is configured to selectively direct the actuator to control a position of the displacement element in order to control the reactivity within the molten salt nuclear fuel contained within the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the displacement element that is formed from a substantially non-neutron-absorbing material. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a molten salt transfer assembly. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt transfer assembly to include a molten salt transfer unit fluidically coupled to the reactor core section and configured to transfer a selected portion of the molten chloride salt fuel from a portion of the fast spectrum molten salt nuclear reactor to a reservoir. The molten salt transfer unit is further configured to transfer a feed material including at least some fertile material from a feed material supply to a portion of the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least some fertile material of the feed material that includes at least one fertile fuel salt. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least one fertile fuel salt in include a salt containing at least one of depleted uranium, natural uranium or thorium. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least one fertile fuel salt to include a salt containing at least one metal from a used nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a fission product removal unit configured to remove at least one fission product from the molten chloride salt fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a primary coolant loop fluidically coupled to the input of the nuclear core section and the output of the nuclear core section. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a primary heat exchanger and a secondary coolant loop, the primary coolant loop and the secondary coolant loop thermally coupled via the primary heat exchanger. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes at least one pump disposed along the primary coolant loop to circulate the molten chloride salt nuclear fuel through the primary coolant loop. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least pump that circulates the molten chloride salt nuclear fuel through the primary coolant loop at or below a selected flow velocity limit. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a gas sparging unit configured to remove one or more noble gases from the molten chloride salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a filter unit configured to remove at least one of a noble metal or a semi-noble metal from the molten salt nuclear fuel. A example method of fueling a fast spectrum molten salt nuclear reactor includes providing a volume of UCl4, providing a volume of at least one of an additional uranium chloride salt or an additional metal chloride salt, mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction, and supplying the molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction to at least a reactor core section of the fast spectrum molten salt nuclear reactor. Another example method of any preceding method includes providing a volume of at least one of an additional uranium chloride salt or an additional metal chloride salt by providing a volume of UCl3. Another example method of any preceding method includes providing a volume of at least one of an additional uranium chloride salt or an additional metal chloride salt by providing a volume of at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 or NdCl3. Another example method of any preceding method includes providing the mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction by mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction and a melting temperature between 330 and 800° C. Another example method of any preceding method includes providing the mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction by mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a composition of 82UCl4-18UCl3. Another example method of any preceding method includes providing the mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction by mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a composition of 17UCl3-71UCl4-12NaCl. Another example method of any preceding method includes providing the mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction by mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a composition of 50 UCl4-50NaCl. Another example method of any preceding method includes providing the mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt by mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt inside of the fast spectrum molten salt nuclear reactor. Another example method of any preceding method includes providing the mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt by mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt outside of the fast spectrum molten salt nuclear reactor. An example molten chloride salt fuel for use in a fast spectrum molten salt nuclear reactor prepared by a process including providing a volume of UCl4, providing a volume of at least one of an additional uranium chloride salt or an additional metal chloride salt, and mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction. An example fast spectrum molten salt nuclear reactor includes a reactor core section including a fuel input and a fuel output. The fuel input and the fuel output are arranged to flow a mixture of molten salt nuclear fuel and at least one lanthanide through the reactor core section at start-up of the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least one lanthanide that includes at least one of La, Ce, Pr or Nd. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of molten salt nuclear fuel and at least one lanthanide that includes a mixture of molten salt nuclear fuel and at least one lanthanide formed by mixing the molten salt nuclear fuel with at least one lanthanide chloride. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least one lanthanide chloride that includes at least one of LaCl3, CeCl3, PrCl3 or NdCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of molten salt nuclear fuel and at least one lanthanide that includes a mixture of molten salt nuclear fuel and at least one lanthanide having a lanthanide concentration between 0.1 and 10% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of molten salt nuclear fuel and at least one lanthanide having a lanthanide concentration between 0.1 and 10% by weight that includes a mixture of molten salt nuclear fuel and at least one lanthanide having a lanthanide concentration between 4 and 8% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of molten salt nuclear fuel and the at least one lanthanide that is formed outside of the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of molten salt nuclear fuel and the at least one lanthanide that is formed inside of the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the fuel input and the fuel output that are arranged to flow a mixture of molten salt nuclear fuel and at least one lanthanide through the reactor core section prior to achieving a selected reactivity threshold in the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the fuel input and the fuel output that are arranged to flow a mixture of molten salt nuclear fuel and at least one lanthanide through the reactor core section prior to achieving criticality in the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the fuel input and the fuel output that are arranged to flow a mixture of molten salt nuclear fuel and at least one lanthanide through the reactor core section prior to generation of a selected amount of plutonium in the fast spectrum molten salt nuclear reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that includes a mixture of at least two of a first uranium chloride, a second uranium chloride or an additional metal chloride. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the additional metal chloride that includes at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3 or AmCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides at least one of the first uranium chloride or the second uranium chloride that includes at least one of UCl4 or UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 82UCl4-18UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 17UCl3-71UCl4-12NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 50 UCl4-50NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 34 UCl3-66NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that includes at least 5% by molar fraction UCl4. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that has a uranium concentration of greater than 61% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that has a melting point between 330 and 800 degrees Celsius. An example method of fueling a fast spectrum molten salt nuclear reactor includes providing a molten salt nuclear fuel and providing at least one lanthanide. Prior to start-up of the fast spectrum molten salt nuclear reactor, the molten salt nuclear fuel is mixed with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel. The lanthanide-loaded molten salt nuclear fuel is supplied to at least a reactor core section of the fast spectrum molten salt nuclear reactor. Another example method of any preceding method provides a molten salt nuclear fuel by providing a mixture of at least two of a first uranium chloride, an additional uranium chloride and an additional metal chloride. Another example method of any preceding method provides a molten salt nuclear fuel by providing a mixture of at least two of UCl4, UCl3 and an additional metal chloride. Another example method of any preceding method provides the additional metal chloride to include at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3 or AmCl3. Another example method of any preceding method provides a molten salt nuclear fuel by providing a molten salt nuclear fuel having at least 5% by molar fraction UCl4. Another example method of any preceding method provides a molten salt nuclear fuel by providing a molten salt nuclear fuel having a uranium concentration of greater than 61% by weight. Another example method of any preceding method provides a molten salt nuclear fuel by providing a molten salt nuclear fuel having a melting point between 330 and 800 degrees Celsius. Another example method of any preceding method provides at least one lanthanide by providing at least one of La, Ce, Pr or Nd. Another example method of any preceding method provides at least one lanthanide by providing at least one lanthanide in the form of a lanthanide chloride. Another example method of any preceding method provides at least one lanthanide in the form of a lanthanide chloride by providing at least one of LaCl3, CeCl3, PrCl3 or NdCl3. Another example method of any preceding method provides mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel by mixing the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel having a lanthanide concentration between 0.1 and 10% by weight. Another example method of any preceding method provides mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel having a lanthanide concentration between 0.1 and 10% by weight by mixing the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel having a lanthanide concentration between 4 and 8% by weight. Another example method of any preceding method provides mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel by mixing the molten salt nuclear fuel with the at least one lanthanide outside of the fast spectrum molten salt nuclear reactor. Another example method of any preceding method provides mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel by mixing the molten salt nuclear fuel with the at least one lanthanide inside of the fast spectrum molten salt nuclear reactor. Another example method of any preceding method provides, prior to start-up of the fast spectrum molten salt nuclear reactor, the mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel by, prior to achieving a selected reactivity threshold in the fast spectrum molten salt nuclear reactor, mixing the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel. Another example method of any preceding method provides, prior to start-up of the fast spectrum molten salt nuclear reactor, mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel by, prior to achieving criticality in the fast spectrum molten salt nuclear reactor, mixing the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel. Another example method of any preceding method provides, prior to start-up of the fast spectrum molten salt nuclear reactor, mixing of the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel by, prior to generation of a selected amount of plutonium in the fast spectrum molten salt nuclear reactor, mixing the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel. An example molten salt fuel for use in a fast spectrum molten salt nuclear reactor prepared by a processing that includes providing a molten salt nuclear fuel, providing at least one lanthanide, and prior to start-up of the fast spectrum molten salt nuclear reactor, mixing the molten salt nuclear fuel with the at least one lanthanide to form a lanthanide-loaded molten salt nuclear fuel. An example fast spectrum molten salt nuclear reactor includes a reactor core section including a fuel input and a fuel output. The fuel input and the fuel output are arranged to flow a molten salt nuclear fuel through the reactor core section. A displacement assembly is operably coupled to the reactor core section and configured to selectively displace a volume of the molten salt nuclear fuel in order to control reactivity within the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement assembly as configured to selectively displace a volume of the molten salt nuclear fuel at a central region of the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement assembly as configured to displace a volume of the molten salt nuclear fuel in order to reduce reactivity within the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement assembly to include a displacement element, an actuator operably coupled to the displacement element, and a controller. The controller is configured to selectively direct the actuator to control a position of the displacement element in order to control the reactivity within the molten salt nuclear fuel contained within the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement element and the reactor section to be centered along a common axis. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the actuator as configured to drive the displacement assembly into the reactor core section in order to reduce the reactivity within the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the actuator as configured to withdraw the displacement assembly from the reactor core section in order to increase the reactivity within the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding claim includes a reactivity parameter sensor configured to sense at least one reactivity parameter of the molten chloride salt nuclear fuel, wherein the reactivity parameter sensor is communicatively coupled to the controller. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the reactivity parameter sensor that includes at least one of a fission detector, a neutron flux monitor, a neutron fluence sensor, a fission product sensor, a temperature sensor, a pressure sensor or a power sensor. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the controller as configured to selectively direct the actuator to control the position of the displacement element within the reactor core section in response to at least one sensed reactivity parameter of the molten chloride salt nuclear fuel from the reactivity parameter sensor. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement element that includes a displacement rod. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement element that includes a plurality of displacement rods. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement element as formed from a substantially non-neutron-absorbing material. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the displacement element as formed from at least one of a substantially neutron-transparent material or a substantially neutron-reflective material. Another example fast spectrum molten salt nuclear reactor of any preceding claim includes a spill-over system configured to transport excess molten salt nuclear fuel out of the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the spill-over system that includes a fuel salt uptake. The fuel salt uptake is positioned above a selected maximum molten salt nuclear fuel fill level of the reactor core section and configured to transport excess molten salt nuclear fuel out of the reactor core section. At least one fluid transport element and a spill-over reservoir are also included. The at least one fluid transport element fluidically couples the fuel salt uptake and the spill-over reservoir. The spill-over reservoir is configured to store excess molten salt nuclear fuel received from the at least one fluid transport element. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the molten salt nuclear fuel that includes a mixture of at least two of a first uranium chloride, a second uranium chloride or an additional metal chloride. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the additional metal chloride that includes at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 or NdCl3. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides at least one of the first uranium chloride or the second uranium chloride that includes at least one of UCl4 or UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the molten salt nuclear fuel that has a composition of 82UCl4-18UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the molten salt nuclear fuel that has a composition of 17UCl3-71UCl4-12NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the molten salt nuclear fuel that has a composition of 50 UCl4-50NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the molten salt nuclear fuel that has a composition of 34 UCl3-66NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that includes at least 5% by molar fraction UCl4. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that has a uranium concentration of greater than 61% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that has a melting point between 330 and 800 degrees Celsius. Another example fast spectrum molten salt nuclear reactor of any preceding claim provides the molten salt nuclear fuel that includes a mixture of at least one uranium fluoride and an additional metal fluoride. An example method includes determining a reactivity parameter in a molten salt nuclear fuel of a molten salt nuclear reactor and, responsive to the reactivity parameter in the molten salt nuclear fuel, displacing a selected volume of the molten salt nuclear fuel with at least one displacement element to control the reactivity of the molten salt nuclear fuel. Another example method of any preceding method provides the determining a reactivity parameter in a molten salt nuclear fuel of a molten salt nuclear reactor by acquiring at least one of a neutron production rate, a neutron absorption rate, a neutron flux, a neutron fluence, a temperature, a pressure, a power or a fission product production rate of the molten salt nuclear fuel, and determining a reactivity parameter in the molten salt nuclear fuel of a molten salt nuclear reactor based on the at least one of a neutron production rate, a neutron absorption rate, a neutron flux, a neutron fluence, a temperature, a pressure, a power or a fission product production rate. Another example method of any preceding method provides, responsive to a reactivity parameter in the molten salt nuclear fuel, displacing a selected volume of the molten salt nuclear fuel with at least one displacement element to adjust the reactivity of the molten salt nuclear fuel by responsive to a reactivity parameter indicative of excess reactivity in the molten salt nuclear reactor, displacing a selected volume of the molten salt nuclear fuel with at least one displacement element to reduce the reactivity of the molten salt nuclear reactor. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel with at least one displacement element by displacing a selected volume of the molten salt nuclear fuel by driving at least a portion of at least one displacement element into the molten salt nuclear fuel to reduce the reactivity of the molten salt nuclear reactor. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel with at least one displacement element by displacing a selected volume of the molten salt nuclear fuel by withdrawing at least a portion of at least one displacement element from the molten salt nuclear fuel to increase the reactivity of the molten salt nuclear reactor. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel by driving at least a portion of at least one displacement element into the molten salt nuclear fuel by displacing a selected volume of the molten salt nuclear fuel by driving a selected amount of at least one displacement element into the molten salt nuclear fuel, wherein the selected amount is based on the determined reactivity parameter. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel by driving at least a portion of at least one displacement element into the molten salt nuclear fuel by displacing a selected volume of the molten salt nuclear fuel by driving at least a portion of at least one displacement element into a volume of the molten salt nuclear fuel within a reactor core section of the molten salt nuclear reactor. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel by driving at least a portion of at least one displacement element into a volume of the molten salt nuclear fuel within a reactor core section of the molten salt nuclear reactor by displacing a selected volume of the molten salt nuclear fuel by driving at least a portion of at least one displacement element into a volume of the molten salt nuclear fuel at a central region of the reactor core section of the molten salt nuclear reactor. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel with at least one displacement element by displacing a selected volume of the molten salt nuclear fuel with at least one displacement rod. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel with at least one displacement rod by displacing a selected volume of the molten salt nuclear fuel with at least one hollow displacement rod. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel with at least one displacement rod by displacing a selected volume of the molten salt nuclear fuel with at least one solid displacement rod. Another example method of any preceding method provides displacing a selected volume of the molten salt nuclear fuel with at least one displacement rod by displacing a selected volume of the molten salt nuclear fuel with a plurality of displacement rods. Another example method of any preceding method provides the at least one displacement rod that is formed from at least one of lead or tungsten. Another example method of any preceding method provides the displacing a selected volume of the molten salt nuclear fuel with at least one displacement element by displacing a selected volume of the molten salt nuclear fuel with at least one displacement rod formed from a substantially non-neutron-absorbing material. Another example method of any preceding method provides the displacing a selected volume of the molten salt nuclear fuel with at least one displacement element by displacing between 0.1 and 10 cubic meters of the molten salt nuclear fuel with at least one displacement element. Another example method of any preceding method provides determining a reactivity parameter in a molten salt nuclear fuel of a molten salt nuclear reactor by determining a reactivity parameter in a molten salt nuclear fuel including a mixture of at least two of a first uranium chloride, an additional uranium chloride or an additional metal chloride. Another example method of any preceding method provides determining a reactivity parameter in a molten salt nuclear fuel including a mixture of at least two of a first uranium chloride, an additional uranium chloride or an additional metal chloride by determining a reactivity parameter in a molten salt nuclear fuel including a mixture of at least two of a first uranium chloride, an additional uranium chloride or an additional metal chloride a mixture of at least two of UCl4, UCl3 and an additional metal chloride. Another example method of any preceding method provides the additional metal chloride that includes at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 or NdCl3. Another example method of any preceding method provides determining a reactivity parameter in a molten salt nuclear fuel by determining a reactivity parameter in a molten salt nuclear fuel having at least 5% by molar fraction UCl4. Another example method of any preceding method provides the determining a reactivity parameter in a molten salt nuclear fuel by determining a reactivity parameter in a molten salt nuclear fuel having a uranium concentration of greater than 61% by weight. Another example method of any preceding method provides determining a reactivity parameter in a molten salt nuclear fuel by determining a reactivity parameter in a molten salt nuclear fuel having a melting point between 330 and 800 degrees Celsius. An example fast spectrum molten salt nuclear reactor includes a reactor core section including a fuel input and a fuel output, the fuel input and the fuel output arranged to flow a molten salt nuclear fuel through the reactor core section and a molten fuel salt exchange assembly operably coupled to the reaction core section and configured to replace a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor. The molten salt nuclear fuel includes at least some fissile material. The feed material includes at least some fertile material. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the selected volume of feed material that is substantially equal in volume to the selected volume of the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the replaced selected volume of the molten salt nuclear fuel that includes at least some fission products. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least some fission products that includes one or more lanthanides. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the replaced selected volume of the molten salt nuclear fuel that includes a carrier salt. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten fuel salt exchange assembly that includes a used-fuel transfer unit fluidically coupled to the reactor core section and configured to transfer a selected volume of the molten salt fuel from the reactor core section to a reservoir and a feed-fuel supply unit fluidically coupled to the reactor core section and configured to transfer a selected volume of feed material including at least some fertile material from a feed material source to the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding reactor that includes a controller is configured to selectively direct the used-fuel unit to transfer a selected volume of the molten salt fuel from the reactor core section to a reservoir and to selectively direct the feed-fuel supply unit to transfer a feed material including at least some fertile material from a feed material source to a portion of the reactor core section. Another example fast spectrum molten salt nuclear reactor of any preceding reactor that includes a reactivity parameter sensor configured to sense at least one reactivity parameter of the molten salt nuclear fuel, wherein the reactivity parameter sensor is communicatively coupled to the controller. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the controller as configured to selectively direct the used-fuel transfer unit to transfer a selected volume of the molten salt fuel from the reactor core section to a reservoir and the controller is further configured to selectively direct the feed-fuel supply unit to transfer a feed material including at least some fertile material from a feed material source to a portion of the reactor core section in response to at least one sensed reactivity parameter of the molten salt nuclear fuel from the reactivity parameter sensor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the reactivity parameter sensor that includes at least one of a fission detector, a neutron flux monitor, a neutron fluence sensor, a fission product sensor, a temperature sensor, a pressure sensor or a power sensor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the reservoir that includes at least one of a storage reservoir. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the reservoir that includes at least one second generation molten salt reactor. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least some fertile material of the feed material that includes at least one fertile fuel salt. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least one fertile fuel salt that includes a salt containing at least one of depleted uranium, natural uranium or thorium. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the at least one fertile fuel salt that includes a salt containing at least one metal from a used nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that includes a mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the additional metal chloride that includes at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 or NdCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides at least one of the first uranium chloride or the second uranium chloride that includes at least one of UCl4 or UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 82UCl4-18UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 17UCl3-71UCl4-12NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 50 UCl4-50NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that has a composition of 34 UCl3-66NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that includes at least 5% by molar fraction UCl4. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that has a uranium concentration of greater than 61% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of at least a first uranium chloride, a second uranium chloride and an additional metal chloride that has a melting point between 330 and 800 degrees Celsius. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the molten salt nuclear fuel that includes a mixture of at least one uranium fluoride and an additional metal fluoride. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a gas sparging unit configured to remove a noble gas from the molten salt nuclear fuel. Another example fast spectrum molten salt nuclear reactor of any preceding reactor includes a filter unit configured to remove at least one of a noble metal or a semi-noble metal from the molten salt nuclear fuel. An example method includes operating a molten salt fast spectrum nuclear reactor including a molten salt nuclear fuel and replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor. The molten salt nuclear fuel includes at least some fissile material. The feed material includes at least some fertile material. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material by replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material equal in volume to the selected volume of the molten salt nuclear reactor. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material by replacing a selected volume of the molten salt nuclear fuel including at least some fission products with a selected volume of feed material. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel including at least some fission products with a selected volume of feed material by replacing a selected volume of the molten salt nuclear fuel including one or more lanthanides with a selected volume of feed material. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material by replacing a selected volume of the molten salt nuclear fuel including a carrier salt with a selected volume of feed material. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor by replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to maintain the reactivity of the molten salt nuclear fuel of molten salt nuclear reactor. Another example method of any preceding method includes measuring a reactivity parameter of the molten salt nuclear fuel of the molten salt fast spectrum nuclear reactor. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor by, responsive to the measured reactivity parameter of the molten salt nuclear fuel, replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor. Another example method of any preceding method provides measuring a reactivity parameter of the molten salt nuclear fuel of the molten salt fast spectrum nuclear reactor by measuring at least one of a neutron production rate, a neutron absorption rate, a neutron flux, a neutron fluence, a temperature, a pressure, a power or a fission product production rate of the molten salt nuclear fuel of the molten salt fast spectrum nuclear reactor. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor by continuously replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor by repeatedly replacing a selected batch volume of the molten chloride salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor. Another example method of any preceding method provides replacing a selected volume of the molten salt nuclear fuel with a selected volume of feed material to control the reactivity of the molten salt nuclear reactor, the molten salt nuclear fuel including at least some fissile material, the feed material including at least some fertile material by removing a selected volume of the molten salt nuclear fuel from the fast spectrum molten salt nuclear reactor, the removed selected volume of molten salt nuclear fuel including at least some fissile material, and supplying a selected volume of feed material to the fast spectrum molten salt nuclear reactor, the supplied selected volume of feed material including at least some fertile material. Another example method of any preceding method provides a rate of supply of the selected volume of feed material that is selected to match a rate of addition of fertile material into the molten salt nuclear reactor to a rate of burning of fissile material within the molten salt nuclear reactor. Another example method of any preceding method provides the removed selected volume of the molten salt nuclear fuel that further includes at least one of a fission product, a fertile material or a carrier salt. Another example method of any preceding method provides the at least some fertile material of the feed material that includes at least one fertile fuel salt. Another example method of any preceding method provides the at least one fertile fuel salt that includes a salt containing at least one of depleted uranium, natural uranium or thorium. Another example method of any preceding method provides the at least one fertile fuel salt that includes a salt containing at least one metal from a used nuclear fuel. Another example method of any preceding method provides the at least one fertile fuel salt that maintains a chemical composition of the molten salt reactor fuel. Another example method of any preceding method includes removing a noble gas from the molten salt nuclear fuel via a gas sparging process. Another example method of any preceding method includes removing at least one of a noble metal or a semi-noble metal from the molten salt nuclear fuel via a plating process. An example system includes at least one first generation molten salt nuclear reactor including a molten salt nuclear fuel, at least one second generation molten salt nuclear reactor, and a molten salt transfer unit configured to transfer a volume of molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor to at least one second generation molten salt nuclear reactor. The volume of the molten salt nuclear fuel includes at least some fissile material enriched in the at least one first generation molten salt nuclear reactor. Another example system of any preceding system provides the volume of the molten salt nuclear fuel including at least some fissile material that is enriched in the at least one first generation molten salt nuclear reactor to so as to achieve criticality in the at least one second generation molten nuclear reactor. Another example system of any preceding system provides the volume of the molten salt nuclear fuel including at least some fissile material that is enriched in the at least one first generation molten salt nuclear reactor to so as to achieve criticality in the at least one second generation molten nuclear reactor without enrichment of the volume of the molten salt nuclear fuel in the at least one second generation molten nuclear reactor. Another example system of any preceding system provides operation of the at least one first generation molten salt nuclear reactor to enrich at least some uranium to generate Pu-239 within the at least one first generation molten salt nuclear reactor. Another example system of any preceding system provides the volume of molten salt nuclear fuel transferred from the at least one first generation molten salt nuclear reactor to the at least one second generation molten salt nuclear reactor that includes Pu-239 generated within the at least one first generation molten salt nuclear reactor. Another example system of any preceding system provides the molten salt transfer unit that includes a fission product removal system configured to remove one or more fission products from the volume of molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor. Another example system of any preceding system provides the at least one first generation molten salt nuclear reactor that includes: a plurality of first generation molten salt nuclear reactors. Another example system of any preceding system provides the at least one second generation molten salt nuclear reactor that includes a plurality of second generation molten salt nuclear reactors. Another example system of any preceding system provides the molten salt nuclear fuel of the at least one first generation molten salt nuclear reactor that includes a mixture of at least two of a first uranium chloride, a second uranium chloride or an additional metal chloride. Another example system of any preceding system provides the additional metal chloride that includes at least one of NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, PuCl3, AmCl3, LaCl3, CeCl3, PrCl3 or NdCl3. Another example system of any preceding system provides at least one of the first uranium chloride or the second uranium chloride that includes at least one of UCl4 or UCl3. Another example system of any preceding system provides the molten salt nuclear fuel that has a composition of 82UCl4-18UCl3. Another example system of any preceding system provides the molten salt nuclear fuel that has a composition of 17UCl3-71UCl4-12NaCl. Another example system of any preceding system provides the molten salt nuclear fuel that has a composition of 50 UCl4-50NaCl. Another example system of any preceding system provides the molten salt nuclear fuel that has a composition of 34 UCl3-66NaCl. Another example system of any preceding system provides the mixture of at least two of a first uranium chloride, a second uranium chloride or an additional metal chloride that includes at least 5% by molar fraction UCl4. Another example system of any preceding system provides the mixture of at least two of a first uranium chloride, a second uranium chloride or an additional metal chloride that has a uranium concentration of greater than 61% by weight. Another example system of any preceding system provides the mixture of at least two of a first uranium chloride, a second uranium chloride or an additional metal chloride that has a melting point between 330 and 800 degrees Celsius. Another example system of any preceding system provides the molten salt nuclear fuel of the at least one first generation molten salt nuclear reactor that includes a mixture of at least one uranium fluoride and an additional metal fluoride. An example method includes enriching at least a portion of a molten salt nuclear fuel in at least one first generation molten salt nuclear reactor, removing a volume of the enriched molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor, and supplying at least a portion of the removed volume of molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor to at least one second generation molten salt nuclear reactor. Another example method of any preceding method provides enriching at least a portion of a molten salt nuclear fuel in at least one first generation molten salt nuclear reactor by enriching at least a portion of a molten salt nuclear fuel in at least one first generation molten salt nuclear reactor so as to achieve criticality in the at least one second generation molten nuclear reactor. Another example method of any preceding method provides enriching at least a portion of a molten salt nuclear fuel in at least one first generation molten salt nuclear reactor so as to achieve criticality in the at least one second generation molten nuclear reactor by enriching at least a portion of a molten salt nuclear fuel in at least one first generation molten salt nuclear reactor so as to achieve criticality in the at least one second generation molten nuclear reactor without enrichment of the volume of the molten salt nuclear fuel in the at least one second generation molten nuclear reactor. Another example method of any preceding method provides enriching at least a portion of a molten salt nuclear fuel in at least one first generation molten salt nuclear reactor by enriching at least some uranium within a volume of the molten salt nuclear fuel of the at least one first generation molten salt nuclear reactor to generate Pu-239. Another example method of any preceding method incudes removing one or more fission products from the at least a portion of the volume of molten salt nuclear fuel removed from the at least one first generation molten salt nuclear reactor. Another example method of any preceding method provides supplying at least a portion of the removed volume of molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor to at least one second generation molten salt nuclear reactor by supplying a portion of the removed volume of molten salt nuclear fuel from the at least one first generation molten fast spectrum salt nuclear reactor to a first second generation molten salt nuclear reactor and supplying at least one additional portion of the removed volume of molten salt nuclear fuel from the at least one first generation fast spectrum molten salt nuclear reactor to at least one additional second generation molten salt nuclear reactor. Another example method of any preceding method provides removing a volume of the enriched molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor by removing a volume of molten salt nuclear fuel from at least one first generation molten salt nuclear reactor to control reactivity of the at least one first generation molten salt nuclear reactor. Another example method of any preceding method provides removing a volume of the enriched molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor by continuously removing a volume of the enriched molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor. Another example method of any preceding method provides removing a volume of the enriched molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor by repeatedly removing a selected batch of a volume of the enriched molten salt nuclear fuel from the at least one first generation molten salt nuclear reactor. Another example method of any preceding method that includes supplying a selected volume of feed material to the at least one first generation molten salt nuclear reactor, the feed material including at least some fertile material. Another example method of any preceding method provides the at least some fertile material of the feed material that includes at least one fertile fuel salt. Another example method of any preceding method provides the at least one fertile fuel salt that includes a salt containing at least one of depleted uranium, natural uranium or thorium. Another example method of any preceding method provides the at least one fertile fuel salt that includes a salt containing at least one metal from a used nuclear fuel. Another example method of any preceding method provides the at least one fertile fuel salt that maintains a chemical composition of the molten salt reactor fuel. Another example method of any preceding method includes supplying a selected volume of feed material to the at least one second generation molten salt nuclear reactor, the feed material including at least some fertile material. An example fast spectrum molten salt nuclear reactor includes a reactor core section including a fuel input and a fuel output. The fuel input and the fuel output are arranged to flow a molten chloride salt nuclear fuel through the reactor core section. The molten chloride salt nuclear fuel includes a mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt, the mixture of UCl4 and at least one additional metal chloride salt having a UCl4 content greater than 5% by molar fraction. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the uranium concentration in the mixture of UCl4 and at least one additional metal chloride salt that is greater than 61% by weight. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the additional uranium chloride salt that includes UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt has a composition of 82UCl4-18UCl3. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt that has a composition of 17UCl3-71UCl4-12NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt that has a composition of 50 UCl4-50NaCl. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt that has an additional metal chloride salt concentration at or below the precipitation concentration for the additional metal chloride salt. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the mixture of UCl4 and at least one of an additional uranium chloride salt or an additional metal chloride salt that has a melting temperature below a temperature of 800 degrees Celsius. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides the selected melting temperature that is above a temperature of 330 degrees Celsius. Another example fast spectrum molten salt nuclear reactor of any preceding reactor provides breed-and-burn behavior that is established within the molten chloride salt nuclear fuel with a uranium-plutonium cycle. An example method of fueling a fast spectrum molten salt nuclear reactor includes providing a volume of UCl4, providing a volume of at least one of an additional uranium chloride salt or an additional metal chloride salt, mixing the volume of UCl4 with the volume of the at least one of an additional uranium chloride salt or an additional metal chloride salt to form a molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction, and supplying the molten chloride salt nuclear fuel having a UCl4 content greater than 5% by molar fraction to at least a reactor core section of the fast spectrum molten salt nuclear reactor. Another example method of any preceding method provides a volume of at least one of an additional uranium chloride salt or an additional metal chloride salt by providing a volume of UCl3. Another example method of any preceding method provides the chlorine in the UCl4 that is enriched with 37Cl. Another example method of any preceding method provides the chlorine in the salt that is enriched to at least 75% 37Cl. The above specification, examples, and data provide a complete description of the structure and use of exemplary implementations of the invention. Since many implementations of the invention can be made without departing from the spirit and scope of the invention, the invention resides in the claims hereinafter appended. Furthermore, structural features of the different implementations may be combined in yet another implementation without departing from the recited claims. |
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claims | 1. A portable X-ray scanner, comprising:an X-ray detector configured to generate images based on incident X-ray radiation;an X-ray tube configured to output X-ray radiation directed toward the X-ray detector; anda frame configured to:hold the X-ray detector;hold the X-ray tube such that the X-ray tube directs the X-ray radiation to the X-ray detector;enable a single user to position the X-ray detector and the X-ray tube while carrying the frame during output of the X-ray radiation; andenable attachment of a Compton scatter shielding device to the frame; anda switch configured to detect attachment of the shielding device to the frame and enable activation of the X-ray tube in response to detecting the attachment of the shielding device. 2. The portable X-ray scanner as defined in claim 1, further comprising a collimator configured to filter the output of the X-ray radiation, the switch configured to detect attachment of the shielding device adjacent the collimator. 3. The portable X-ray scanner as defined in claim 1, wherein the switch comprises a sensor configured to detect the presence of the shielding device. 4. The portable X-ray scanner as defined in claim 3, wherein the sensor comprises at least one of a mechanical switch, a capacitive sensor, an inductive sensor, a magnetic sensor, or an optical sensor. 5. The portable X-ray scanner as defined in claim 1, wherein the switch is configured to enable activation of the X-ray tube based on whether the X-ray tube is configured to use a tube voltage that satisfies a threshold tube voltage. 6. The portable X-ray scanner as defined in claim 5, wherein the threshold tube voltage is 70 kV, and the switch is configured to disable the X-ray tube when the X-ray tube is configured to use at least the threshold tube voltage and attachment of the shielding device to the frame is not detected. 7. The portable X-ray scanner as defined in claim 1, wherein the frame comprises an attachment rail configured to hold the shielding device. 8. A portable X-ray scanner, comprising:an X-ray detector configured to generate images based on incident X-ray radiation;an X-ray tube configured to output X-ray radiation; anda frame configured to:hold the X-ray detector;hold the X-ray tube such that the X-ray tube directs the X-ray radiation to the X-ray detector; andenable a single user to position the X-ray detector and the X-ray tube while carrying the frame during output of the X-ray radiation;a backscatter shield configured to provide shielding from Compton scatter radiation when placed in contact with an object to be scanned; anda switch configured to detect attachment of the backscatter shield to the frame and enable activation of the X-ray tube in response to detecting the attachment of the backscatter shield. 9. The portable X-ray scanner as defined in claim 8, wherein the switch is configured to enable activation of the X-ray tube based on whether the X-ray tube is configured to use a tube voltage that satisfies a threshold tube voltage. 10. The portable X-ray scanner as defined in claim 9, wherein the threshold tube voltage is 70 kV, and the switch is configured to disable the X-ray tube when the X-ray tube is configured to use at least the threshold tube voltage and attachment of the shielding device to the frame is not detected. |
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048872830 | abstract | An X-ray mask includes a membrane formed of a material which transmits X-rays, a mask pattern formed on the surface of the membrane, the mask pattern being made of an X-ray absorbing material, and a supporting frame formed of a material which is mechanically deformed by an external signal. The supporting frame supports the membrane. In an exposure method which employs this X-ray mask, the X-ray mask is first disposed above a substrate in alignment therewith. Subsequently, distortion in the mask pattern is corrected by the application of the external signal to the supporting frame of the mask, and the substrate is then irradiated with X-rays through the mask so as to transfer the mask pattern of the mask to the substrate. |
description | The subject matter disclosed herein relates generally to apparatus and methods for diagnostic medical imaging, such as Nuclear Medicine (NM) imaging. In NM imaging, systems with multiple detectors or detector heads may be used to image a subject, such as to scan a region of interest. For example, the detectors may be positioned adjacent the subject to acquire NM data, which is used to generate a three-dimensional (3D) image of the subject. Single Photon Emission Computed Tomography (SPECT) systems may have moving detector heads, such as gamma detectors positioned to focus on a region of interest. For example, a number of gamma cameras may be moved (e.g., rotated) to different angular positions for acquiring image data. The acquired image data is then used to generate the 3D images. The size of the detector heads may limit an available usable area for the placement of detectors, such as Cadmium Zinc Telluride (CZT) wafers. The sensitivity (e.g., the proportion of radiation received relative to the radiation emitted) may be limited by the size of the detector heads and/or the arrangement of CZT wafers. Conventional approaches to improving sensitivity may use thicker detectors, or detectors arranged in generally identical or similar layers stacked directly one on top of each other. Such conventional approaches may not provide a desired or required sensitivity. In one embodiment, a detector assembly is provided that includes a semiconductor detector, a pinhole collimator, and a processing unit. The semiconductor detector has a first surface and a second surface opposed to each other. The first surface includes pixelated anodes, and the second surface includes a cathode electrode. The pinhole collimator includes an array of pinhole openings corresponding to the pixelated anodes. Each pinhole opening is associated with a single pixelated anode of the semiconductor detector, and the area of each pinhole opening is smaller than a corresponding area of the corresponding pixel, which is exposed to radiation (or the pixel area less the radiation blocking area of the collimator immediately above the pixel). The processing unit is operably coupled to the semiconductor detector and configured to identify detected events within virtual sub-pixels distributed along a length and width of the semiconductor detector. Each pixelated anode includes (e.g., has associated therewith) a plurality of corresponding virtual sub-pixels (as interpreted by the processing unit), wherein absorbed photons are counted as events in a corresponding virtual sub-pixel. In another embodiment, a detector assembly is provided that includes a semiconductor detector, a collimator, and a processing unit. The semiconductor detector has a first surface and a second surface opposed to each other. The first surface includes pixels (which in turn comprise corresponding pixelated anodes), and the second surface includes a cathode electrode. The collimator includes openings. Each opening is associated with a single corresponding pixelated anode of the semiconductor detector. The processing unit is configured to identify detected events within virtual sub-pixels distributed along a length and width of the semiconductor detector. Each pixel includes (e.g., has associated therewith) a plurality of corresponding virtual sub-pixels. Absorbed photons are counted as events in a corresponding virtual sub-pixel, with absorbed photons counted as events within a thickness of the semiconductor detector at a distance corresponding to one over an absorption coefficient of the detector. In another embodiment, a detector assembly includes a semiconductor detector, a collimator and a processing unit. The semiconductor detector has a first surface and a second surface opposed to each other. The first surface includes pixels (which in turn comprise corresponding pixelated anodes), and the second surface includes a cathode electrode. The collimator includes openings, with each opening associated with a single corresponding pixel of the semiconductor detector. The processing unit is configured to identify detected events within virtual sub-pixels distributed along a length and width of the semiconductor detector. Each pixel includes (e.g., has associated therewith) a plurality of corresponding virtual sub-pixels, with absorbed photons are counted as events in a corresponding virtual sub-pixel. Absorbed photons are counted as events within a thickness of the semiconductor detector at a distance corresponding to an energy window width used to identify the events as photon impacts. The following detailed description of certain embodiments will be better understood when read in conjunction with the appended drawings. To the extent that the figures illustrate diagrams of the functional blocks of various embodiments, the functional blocks are not necessarily indicative of the division between hardware circuitry. For example, one or more of the functional blocks (e.g., processors or memories) may be implemented in a single piece of hardware (e.g., a general purpose signal processor or a block of random access memory, hard disk, or the like) or multiple pieces of hardware. Similarly, the programs may be stand alone programs, may be incorporated as subroutines in an operating system, may be functions in an installed software package, and the like. It should be understood that the various embodiments are not limited to the arrangements and instrumentality shown in the drawings. As used herein, the terms “system,” “unit,” or “module” may include a hardware and/or software system that operates to perform one or more functions. For example, a module, unit, or system may include a computer processor, controller, or other logic-based device that performs operations based on instructions stored on a tangible and non-transitory computer readable storage medium, such as a computer memory. Alternatively, a module, unit, or system may include a hard-wired device that performs operations based on hard-wired logic of the device. Various modules or units shown in the attached figures may represent the hardware that operates based on software or hardwired instructions, the software that directs hardware to perform the operations, or a combination thereof. “Systems,” “units,” or “modules” may include or represent hardware and associated instructions (e.g., software stored on a tangible and non-transitory computer readable storage medium, such as a computer hard drive, ROM, RAM, or the like) that perform one or more operations described herein. The hardware may include electronic circuits that include and/or are connected to one or more logic-based devices, such as microprocessors, processors, controllers, or the like. These devices may be off-the-shelf devices that are appropriately programmed or instructed to perform operations described herein from the instructions described above. Additionally or alternatively, one or more of these devices may be hard-wired with logic circuits to perform these operations. As used herein, an element or step recited in the singular and preceded with the word “a” or “an” should be understood as not excluding plural of said elements or steps, unless such exclusion is explicitly stated. Furthermore, references to “one embodiment” of are not intended to be interpreted as excluding the existence of additional embodiments that also incorporate the recited features. Moreover, unless explicitly stated to the contrary, embodiments “comprising” or “having” an element or a plurality of elements having a particular property may include additional elements not having that property. Various embodiments provide systems and methods for improving the sensitivity of image acquisition, for example in Nuclear Medicine (NM) imaging applications. Various embodiments provide one or more different approaches for improving sensitivity and/or other aspects of detector performance. For example, in one approach, an array of pinhole openings are used in a collimator for a detector system. As another example, additionally or alternatively, in a second approach, all events are identified as being absorbed at a location and/or within a range corresponding to an absorption coefficient of the detector (e.g., one over the absorption coefficient of the detector). As one more example, in a third approach, all events are identified as being absorbed at a location and/or within a range that ensures that the energy of the events is measured within the energy window used for imaging. It may be noted that each of the three approaches discussed above in this paragraph may be employed with the use of virtual sub-pixels (or virtual division of the detector) along X and Y directions (or along the width and length of the detector). In various embodiments, a pinhole collimator includes an array of pinholes that defines multiples cells. Each cell includes or corresponds to only a single physical pixel of the detector and only a single pinhole of the array. In various embodiments, in contrast to certain conventional approaches, radiation from a given pinhole (also referred to herein as a pinhole opening) only arrives at one particular physical pixel corresponding to the given pinhole. It may be noted that the physical pixel may be viewed as including a number of virtual sub-pixels by a processing unit. Each event detected in the physical pixel is counted as related to one of the virtual sub-pixels into which the physical pixel is divided to. The virtual sub-pixel to which the event belongs to is determined by the location of the event within the physical pixel that includes the virtual sub-pixels. There are known methods to derive the location of the event within the physical pixel, such as the method described in U.S. patent application Ser. No. 15/280,640 entitled “SYSTEMS AND METHODS FOR SUB-PIXEL LOCATION DETERMINATION” filed Sep. 29, 2016. virtual sub-pixels Further, use of a pinhole array instead of a parallel hole array in various embodiments provides for a smaller area of opening that collects radiation. For example, the size of a pinhole opening may be ⅓ of the width (or 1/9 of the area) of a pixel, whereas an opening of a parallel hole array may be the pixel size less the septa thickness. Accordingly, the opening size in a parallel hole array may be dictated by the pixel size and wall (or septa) thickness; however, in various embodiments employing a pinhole array, the opening size may be selected as desired (e.g., to provide a desired sensitivity and/or collimator height). In some embodiments, physical pixels may be divided in multiple virtual sub-pixels (e.g., an associated processing unit may assign virtual sub-pixels to each physical pixel) along X and Y directions (or length and width of a detector), while having a single layer in the Z direction (or thickness). Alternatively, in other embodiments, multiple virtual sub-pixels may be employed along the Z direction (or thickness). Accordingly, various embodiments provide flexibility to selected collimator height. Such flexibility is especially advantageous when using a collimator within a swiveling detector head. For example, the collimator height may dictate or influence the radius reserved for each head for swiveling motion. The shorter the collimator is, the smaller is the required radius. Accordingly, for a shorter collimator provided by a pinhole array, more heads may be placed around the object being imaged providing improved sensitivity and image quality in comparison with a taller parallel hole collimator. Additionally, the collimator openings produce solid angles through which the virtual pixels observe the object being imaged, with the smaller size of the pinhole openings providing larger and more separated solid angles for the virtual sub-pixels, with less overlap between the solid angles for the virtual sub-pixels of a physical pixel. Accordingly, the use of such pinhole openings improves spatial resolution, or may be used to maintain a desired resolution with a shorter collimator to increase sensitivity and improve image quality. Further, still, the use of thicker septa or walls helps prevent radiation penetration from a given collimator opening to non-associated pixels (or pixels other than a pixel immediately below the opening), thereby improving image quality. Further still, it may be noted that use of many sub-voxels over the thickness of a detector may reduce the number of events for each volume of interest, thereby increasing statistical noise and degrading image quality. In various embodiments, using only one location (or range) or layer along the Z direction (or thickness) reduces statistical noise and improves image quality. It may be noted that in some embodiments, in connection with sub-pixelization along the X and Y directions, a single Z layer, location, or range may be used to identify events along a thickness of a detector. For example, a Z position-range for all events may be defined at or around an average absorption depth of 1/μ, where μ is the absorption coefficient for a specific photon energy for a particular detector material. For example, events may be distributed linearly, as one example, or exponentially, as another example, within a range centered about or otherwise corresponding to distance of 1/μ from the cathode. As another example, a Z position-range for all events may be defined within a range corresponding to energies of the energy window used for imaging. For example, in some embodiments, an absorption location for each absorbed photon within the thickness of the semiconductor detector is defined within a range such that ΔL/D=ΔE/E, where ΔL is a distance from the cathode, D is the detector thickness, ΔE is an energy window width, and E is a photopeak energy of an absorbed photon. Again, the events may be distributed linearly, as one example, or exponentially, as another example. In various embodiments, use of such Z position-ranges (in contrast, for example, to multiple virtual sub-pixels along a detector thickness) helps to reduce statistical noise and to improve image quality. Also, it may be noted that use of such Z position-ranges may be accomplished with simpler hardware or software (in comparison to, for example, use of multiple virtual sub-pixels along a detector thickness), providing for easier implementation and/or lower cost. A technical effect provided by various embodiments includes increased sensitivity of a detector system, such as a NM imaging detector system. The detector system may be provided in a rotating head detector module that may be used as part of a group of similar rotating head detector modules in an imaging system. A technical effect of various embodiments includes improved image quality and spatial resolution. A technical effect of various embodiments includes reduced collimator height allowing for less room needed to allow a detector head to pivot, allowing more detector heads to be placed closely to an object being imaged. A technical effect of various embodiments includes reduced penetration by radiation to pixels other than a pixel associated with (e.g., located directly below) a collimator opening. A technical effect of various embodiments included reduced statistical noise. FIG. 1 provides a schematic block view of a detector assembly 100 in accordance with various embodiments, FIG. 2 provides an exploded view of aspects of the detector assembly 100, and FIG. 3 provides a sectional view taken along line 3-3 of FIG. 1. As seen in FIGS. 1-3, the detector assembly 100 includes a semiconductor detector 110, a pinhole collimator 130, and a processing unit 150, which for the clarity of the drawings is shown only in FIG. 1. Generally, the semiconductor detector 110 produces signals in response to absorption events (e.g., photons produced in response to a radiopharmaceutical that has been administered to an object being imaged that impact the semiconductor detector 110). The signals are provided to the processing unit 150, which uses identified events to reconstruct an image of the object being imaged and to derive the location of the event inside the physical pixel as described, for example in U.S. patent application Ser. No. 15/280,640. The pinhole collimator 130 guides photons to the semiconductor detector 110, and limits the angular range of approach of photons to a given pixel or portion of the semiconductor detector 110, helping to allow for accurate determination of the portion of the object being imaged from which a given detected event originated. As best seen in FIG. 3, the semiconductor detector 110 has a first surface 112 and a second surface 114. The second surface 114 is opposed to the first surface 112 (and, likewise, the first surface 112 is opposed to the second surface 114). The semiconductor detector 110 is configured to generate electrical signal in response to photon impacts, and may be made of, for example, Cadmium Zinc Telluride (CZT). The second surface 114 includes pixelated anodes 116 disposed thereon, and the first surface 112 includes a cathode electrode 118 disposed thereon. In some embodiments, the cathode electrode 118 may be a monolithic, or single, cathode. The cathode electrode 118 collects an opposite electrical charge of the pixelated anodes 116, and the pixelated anodes 116 are used to generate signals in response to charges generated by the semiconductor detector 110 responsive to photon impacts. The pixelated anodes 116 may be arranged in a grid, with the location of one or more pixelated anodes 116 at which a signal is generated responsive to a photon impact used to determine a corresponding location in the object corresponding to the photon impact. As seen in FIGS. 1-3, the pinhole collimator 130 is interposed between the semiconductor detector 110 and an object being imaged (not shown), and is used to control passage of radiation from the object being imaged to the semiconductor detector 110 via the pinhole collimator 130. For example, the pinhole collimator 130 guides photons to the semiconductor detector 110, limiting an angular range of approach for photons that impact the semiconductor detector 110. The pinhole collimator 130 includes an array 132 of pinhole openings 134 corresponding to the pixelated anodes 116 on the second surface 114. In the illustrated embodiment, the array 132 of pinhole openings 134 has a 1:1 correspondence with an array or grid of pixelated anodes 116, with both the pixelated anodes 116 and array 132 of pinhole openings 134 arranged in an 8×8 layout when the projections of openings 132 on the second surface 114 of the detector 110 are centered in the pixelated anodes 116. Accordingly, in the illustrated embodiment, each pinhole opening 134 is associated with a single pixelated anode 116 of the semiconductor detector 110. Accordingly, radiation that passes through a given pinhole opening 134 is confined within a single cell of collimator 130 and is absorbed at a location corresponding to one and only one pixelated anode 116 that is associated with the given pinhole opening 134 (e.g., located directly beneath the pinhole opening 134). Further, each pinhole opening 134 defines an area, with the area of each pinhole opening 134 smaller than a corresponding area of the corresponding pixelated anode 116. For example, as seen in FIG. 3, the width of each pixelated anode 116 is greater than the width of a corresponding pinhole opening 134. Accordingly, if the pixelated anode 116 and pinhole opening 134 are generally square-shaped, the area of the pixelated anode 116 is greater than the area of the pinhole opening 134. It may be noted that the depicted examples have generally square-shaped cross-sections. Other shapes of opening (e.g., circular, rectangular, or triangular, among others), may be utilized in alternate embodiments. As best seen in FIGS. 1 and 3, in various embodiments the pinhole collimator 130 includes a top plate 140 through which the pinhole openings 134 pass. The top plate 140 is mounted to a collimator base 131. The pinhole collimator 130 also includes plural septa 142 (or walls) extending along a height of the collimator base 131 that define collimator cells 144 corresponding to the pinhole openings 134. In the illustrated embodiment, each pinhole opening 134 is associated with a particular collimator cell 144 and a particular pixelated anode 116, with photons that pass through the pinhole opening 134 passing through the corresponding collimator cell 144 toward the corresponding pixelated anode 116. Each collimator cell 144 defines a cavity between the corresponding pinhole opening 134 and the corresponding pixelated anode 116. The septa 142 act to reduce or eliminate passage of a photon through a pinhole opening 134 to non-corresponding pixelated anodes (e.g., pixelated anodes adjacent to the particular pixelated anode that corresponds to the particular pinhole opening). A cell width 148 defined by the septa 142 (e.g., a width between neighboring septa 142) is greater than an opening width 146 defined by the pinhole openings 134. For example, in some embodiments, the cell width 148 is 3 times or more greater than the opening width 146. In the example illustrated in FIGS. 1 to 3, the cell width 148 is d, and the opening width 146 is d/3, or the cell width 148 is 3 times greater than the opening width 146. It may be noted that, in contrast, an opening width and width between neighboring walls may be identical for a parallel-hole collimator. In the example illustrated in FIGS. 1 to 3, the septa 142 are parallel to each other and define square-shaped cross sections for each collimator cell 144; however, it may be noted that different configurations may be employed in alternate embodiments. As best seen in FIGS. 2 and 3, the top plate 140 has a thickness 141, and the septa 142 have a thickness 143. In the depicted example, the thickness 141 of the top plate 140 is greater than the thickness 143 of the septa 142. In the example illustrated in FIGS. 1 to 3, the septa 142 are parallel to each other and define square-shaped cross sections for each collimator cell 144; however, it may be noted that different configurations may be employed in alternate embodiments. FIG. 4 illustrates a cross-section of an example embodiment of a top plate 400 (e.g., which may be used as top plate 140) that may be used with pinhole collimator 130 in various embodiments. The top plate 400 includes a first surface 402 configured to be positioned proximate a semiconductor detector (e.g., semiconductor detector 110), or oriented toward an interior 410 of a collimator (e.g., pinhole collimator 130) including the top plate 400. The top plate 400 also includes a second surface 404 that is opposed to the first surface 402. The second surface 404 is farther away from the semiconductor detector than the first surface 402 is, or the second surface 404 is oriented toward an object 420 being imaged from which photons 422 are emitted. The top plate 400 includes pinhole openings 430 through which photons 422 pass toward the semiconductor detector. The depicted pinhole openings 430 each have a first width 432 at the first surface 402, and a second width 434 at the second surface 404. The first width 432 is greater than the second width 434. Accordingly, the pinhole openings 430 are tapered, and are larger at the first surface 402 than at the second surface 404. It may be noted that the taper orientation of the openings 430 in the plate 400 of FIG. 4 is opposite to the taper orientation of the openings 134 in the plate 140 of FIGS. 1-3. The tapered shape in various embodiments is configured to facilitate passage of photons over a preferred or desired angular range. Alternatively or additionally, it may be noted that pinhole collimators in various embodiments may include tapered walls. FIG. 5 illustrates a cross-section of an example embodiment of pinhole collimator septa 500 that may be used with pinhole collimator 130 in various embodiments. Septa 500 include a first surface 502 proximate to a top plate (e.g., top plate 141, top plate 400; top plate not shown in FIG. 5), and a second surface 504 proximate to a semiconductor detector (e.g., semiconductor detector 110; semiconductor detector not shown in FIG. 5). Cells 506, through which photons pass, are defined between neighboring septa 500. The cells 506 are a first width 512 at the first surface 502, and a second width 514 at the second surface 504, with the second width 514 greater than the first width 512. Accordingly, a first width 503 of the septa 500 at the first surface 502 is greater than a second width 505 of the septa 500 at the second surface 504. In the illustrated embodiment, a pitch 530 is defined by the septa 500, with the first width 512 less than the pitch 530. The tapered septa 500 in various embodiments may cooperate or be complementary with tapered openings (e.g., openings 434). In various embodiments, the tapered septa 500 may be formed by 3D printing a collimator block. The tapered septa 500 provide additional thickness (e.g., relative to septa thickness of a parallel hole collimator) for improved reduction of penetration by photons into adjacent collimator cells. The tapered shape in various embodiments is configured to facilitate passage of photons over a preferred or desired angular range. It may be noted that while the description above includes top plate 141 or 400, the pinholes-array collimator of FIG. 5 may not include a top plate at all when opening 512 is in the desired size of the pinholes openings such as the size of openings 134 and 434 of FIGS. 4 and 1-3, respectively. Examples of solid angles corresponding to the pinhole collimator 130 may be seen in FIG. 3. As seen in FIG. 3, the solid angles defined by virtual sub-pixels may vary based on a depth (or depths) within the semiconductor detector 110 assigned to events. For example, solid angles 320a, 320b, and 320c result from using a common absorption depth (e.g., 1/μ, where μ is an absorption coefficient) for events from three adjacent virtual sub-pixels 117. As another example, solid angles 330a, 330b, and 330c result from using varying absorption depths for events from three adjacent virtual sub-pixels 117. In FIG. 3, the cell width 148 between septa 142 of the pinhole collimator equals d, a height 149 of the pinhole collimator 130 is h/3, and the opening width 146 defined by the pinhole openings 134 is d/3. By way of comparison, examples of solid angles corresponding to use of a parallel hole collimator may be seen in FIG. 6. In FIG. 6, a detector system 600 includes a parallel hole collimator 602 that includes walls 610 having a common opening width 612 therebetween. The opening width 612 defines the width of openings 620. The detector system 600 also includes a detector unit 621 that includes pixelated anodes 616 having virtual sub-pixels 617 associated therewith (e.g., by a processing unit). As seen in FIG. 6, solid angles 660a, 660b, and 660c result from using a common absorption depth (e.g., at the surface of the detector unit 621) for events from three adjacent virtual sub-pixels 617. As seen in FIGS. 3 and 6, the solid angles (solid angles 320a, 320b, and 320c and/or solid angles 330a, 330b, and 330c) for the pinhole collimator 130 have noticeably less overlap than the solid angles for parallel hole collimator 602 (solid angles 660a, 660b, and 660c). In FIG. 6, the width 612 between walls (as well as width of openings 620) is d, and the height 652 of the parallel hole collimator 602 is h. Returning to FIGS. 1-3, the processing unit 150 is operably coupled to the semiconductor detector 110, and is configured to identify detected events, deriving the location of events within physical pixels 119 and based on their location, assigning them to virtual sub-pixels 117 distributed along a length 190 and width 191 of the semiconductor detector 110 to be counted there. In FIG. 3, the virtual sub-pixels 117 are represented by dashed lines passing through the semiconductor detector corresponding to the location of the virtual sub-pixels 117. It may be noted that in the illustrated embodiment, the semiconductor detector 110 includes pixels 119. In FIG. 3, there are 3 virtual sub-pixels across a width of each pixel 119. Each pixel 119 may be understood as including a pixelated anode 116, with each pixelated anode 116 smaller (having a smaller area) than the corresponding pixel 119. In the illustrated embodiment, there are 9 virtual sub-pixels 117 per pixelated anode 116 or pixel 119 (e.g., a grid of 3×3 virtual sub-pixels 117 per pixelated anode 116 or pixel 119). Each pixel 119 includes a plurality of corresponding virtual sub-pixels, with absorbed photons in the semiconductor detector 110 counted as events in a corresponding virtual sub-pixel. Additional discussion regarding virtual sub-pixels and the use of virtual sub-pixels, and the use of collected and non-collected charge signals may be found in U.S. patent application Ser. No. 14/724,022, entitled “Systems and Method for Charge-Sharing Identifcation and Correction Using a Single Pixel,” filed 28 May 2015 (“the 022 Application); U.S. patent application Ser. No. 15/280,640, entitled “Systems and Methods for Sub-Pixel Location Determination,” filed 29 Sep. 2016 (“the 640 Application”); and U.S. patent application Ser. No. 14/627,436, entitled “Systems and Methods for Improving Energy Resolution by Sub-Pixel Energy Calibration,” filed 20 Feb. 2015 (“the 436 Application). The subject matter of each of the 022 Application, the 640 Application, and the 436 Application are incorporated by reference in its entirety. In various embodiments the processing unit 150 includes processing circuitry configured to perform one or more tasks, functions, or steps discussed herein. It may be noted that “processing unit” as used herein is not intended to necessarily be limited to a single processor or computer. For example, the processing unit 150 may include multiple processors, ASIC's, FPGA's, and/or computers, which may be integrated in a common housing or unit, or which may distributed among various units or housings. It may be noted that operations performed by the processing unit 150 (e.g., operations corresponding to process flows or methods discussed herein, or aspects thereof) may be sufficiently complex that the operations may not be performed by a human being within a reasonable time period. For example, the determination of values of collected, non-collected, and/or combined charge signals within the time constraints associated with such signals may rely on or utilize computations that may not be completed by a person within a reasonable time period. As discussed, herein, signals are generated by one or more pixelated anodes 116 in response to a photon impact, with the location of the pixelated anode(s) 116 generating a signal used to determine a corresponding location in the object for which an event is counted. In various embodiments, as also discussed in the 022 Application, the 640 Application, and the 436 Application, signals from adjacent pixels may be used to assign a virtual sub-pixel location within a given pixelated anode 116. In some embodiments, the processing unit 150 is configured to determine an absorption location for a given absorbed phon based on non-collected signals received from pixelated anodes adjacent to a pixelated anode absorbing the given absorbed photon. Additionally or alternatively to the use of virtual pixels along a length and/or width of the semiconductor detector 110, in various embodiments virtual pixels may be employed along a thickness of the semiconductor detector 110. Virtual pixels employed along a thickness of the semiconductor detector 110 may be used to represent different depths of absorption of photons. For example, in various embodiments, as best seen in FIG. 3, the semiconductor detector 110 has a thickness 396. Three rows of virtual pixels are distributed along the thickness 396—a first row 390, a second row 392, and a third row 394. The processing unit 150 in various embodiments is configured to identify detector events with the virtual sub-pixels in the first row 390, second row 392, and third row 394 distributed along the thickness 396. Accordingly, in various embodiments, different virtual sub-pixels along a thickness may be used to provide different absorption depths for identifying event locations. For example, as seen in FIG. 3, event 331a is shown at a depth corresponding to the first row 390, event 331b is shown at a depth corresponding to the third row 394, and event 331c is shown at a depth corresponding to the second row 392. However, it may be noted that, in other embodiments that may or may not include a pinhole-array collimator, a single absorption depth may be employed. For example, in some embodiments, the processing unit 150 is configured to count absorbed photons as events within the thickness 396 of the semiconductor detector 110 at a location (e.g., a distance from the cathode 118) corresponding to one over an absorption coefficient of the semiconductor detector 110. For example, with μ as the absorption coefficient, photons (e.g., photons at a given energy corresponding with the absorption coefficient) may be counted as events at a location in the semiconductor detector a distance 395 from the second surface 112 (and/or cathode 118) along the thickness 396, as shown for event locations 321a, 321b, and 321c of FIG. 3. The distance 395 in various embodiments is 1/μ. It may be noted that μ may vary based on photon energy. It may further be noted that use of a single absorption depth as discussed herein and in the next paragraph may be used in connection with a pinhole collimator (e.g., pinhole collimator 130) in various embodiments, or may be used in connection with a parallel hole collimator (e.g., parallel hole collimator 602) in other embodiments. In some embodiments, the absorption location for each photon is defined within a range of 1/μ±1 millimeter. As another example of use of a single absorption depth, in some embodiments, the processing unit 150 is configured to count absorbed photons as events within the thickness 396 of the semiconductor detector 110 at a distance corresponding to an energy window width used to identify the events as photon impacts. For example, in some embodiments, an absorption location for each absorbed photon within the thickness 396 of the semiconductor detector 110 is defined within a range such that ΔL/D=ΔE/E, where ΔL is the distance 395 from the first surface 112 (and/or the cathode 118), D is the detector thickness (e.g., thickness 396), ΔE is an energy window width, and E is a photopeak energy of an absorbed photon. The energy window width in various embodiments is a range of energies around the photopeak energy which are considered as true events. FIG. 7 is a schematic illustration of a NM imaging system 1000 having a plurality of imaging detector head assemblies mounted on a gantry (which may be mounted, for example, in rows, in an iris shape, or other configurations, such as a configuration in which the movable detector carriers 1016 are aligned radially toward the patient-body 1010). In particular, a plurality of imaging detectors 1002 are mounted to a gantry 1004. Each detector 1002 may include, for example, collimators and detectors arranged generally similarly to the arrangements discussed in connection with FIGS. 1-6. In the illustrated embodiment, the imaging detectors 1002 are configured as two separate detector arrays 1006 and 1008 coupled to the gantry 1004 above and below a subject 1010 (e.g., a patient), as viewed in FIG. 7. The detector arrays 1006 and 1008 may be coupled directly to the gantry 1004, or may be coupled via support members 1012 to the gantry 1004 to allow movement of the entire arrays 1006 and/or 1008 relative to the gantry 1004 (e.g., transverse translating movement in the left or right direction as viewed by arrow T in FIG. 7). Additionally, each of the imaging detectors 1002 includes a detector unit 1014 (which may include collimator and/or detector assemblies as discussed herein in connection with FIGS. 1-6), at least some of which are mounted to a movable detector carrier 1016 (e.g., a support arm or actuator that may be driven by a motor to cause movement thereof) that extends from the gantry 1004. In some embodiments, the detector carriers 1016 allow movement of the detector units 1014 towards and away from the subject 1010, such as linearly. Thus, in the illustrated embodiment the detector arrays 1006 and 1008 are mounted in parallel above and below the subject 1010 and allow linear movement of the detector units 1014 in one direction (indicated by the arrow L), illustrated as perpendicular to the support member 1012 (that are coupled generally horizontally on the gantry 1004). However, other configurations and orientations are possible as described herein. It should be noted that the movable detector carrier 1016 may be any type of support that allows movement of the detector units 1014 relative to the support member 1012 and/or gantry 1004, which in various embodiments allows the detector units 1014 to move linearly towards and away from the support member 1012. Each of the imaging detectors 1002 in various embodiments is smaller than a conventional whole body or general purpose imaging detector. A conventional imaging detector may be large enough to image most or all of a width of a patient's body at one time and may have a diameter or a larger dimension of approximately 50 cm or more. In contrast, each of the imaging detectors 1002 may include one or more detector units 1014 coupled to a respective detector carrier 1016 and having dimensions of, for example, 4 cm to 20 cm and may be formed of Cadmium Zinc Telluride (CZT) tiles or modules. For example, each of the detector units 1014 may be 8×8 cm in size and be composed of a plurality of CZT pixelated modules (not shown). For example, each module may be 4×4 cm in size and have 16×16=256 pixels. In some embodiments, each detector unit 1014 includes a plurality of modules, such as an array of 1×7 modules. However, different configurations and array sizes are contemplated including, for example, detector units 1014 having multiple rows of modules. It should be understood that the imaging detectors 1002 may be different sizes and/or shapes with respect to each other, such as square, rectangular, circular or other shape. An actual field of view (FOV) of each of the imaging detectors 1002 may be directly proportional to the size and shape of the respective imaging detector. The gantry 1004 may be formed with an aperture 1018 (e.g., opening or bore) therethrough as illustrated. A patient table 1020, such as a patient bed, is configured with a support mechanism (not shown) to support and carry the subject 1010 in one or more of a plurality of viewing positions within the aperture 1018 and relative to the imaging detectors 1002. Alternatively, the gantry 1004 may comprise a plurality of gantry segments (not shown), each of which may independently move a support member 1012 or one or more of the imaging detectors 1002. The gantry 1004 may also be configured in other shapes, such as a “C”, “H” and “L”, for example, and may be rotatable about the subject 1010. For example, the gantry 1004 may be formed as a closed ring or circle, or as an open arc or arch which allows the subject 1010 to be easily accessed while imaging and facilitates loading and unloading of the subject 1010, as well as reducing claustrophobia in some subjects 1010. Additional imaging detectors (not shown) may be positioned to form rows of detector arrays or an arc or ring around the subject 1010. By positioning multiple imaging detectors 1002 at multiple positions with respect to the subject 1010, such as along an imaging axis (e.g., head to toe direction of the subject 1010) image data specific for a larger FOV may be acquired more quickly. Each of the imaging detectors 1002 has a radiation detection face, which is directed towards the subject 1010 or a region of interest within the subject. In various embodiments, multi-bore collimators may be constructed to be registered with pixels of the detector units 1014, which in one embodiment are CZT detectors. However, other materials may be used. Registered collimation may improve spatial resolution by forcing photons going through one bore to be collected primarily by one pixel. Additionally, registered collimation may improve sensitivity and energy response of pixelated detectors as detector area near the edges of a pixel or in-between two adjacent pixels may have reduced sensitivity or decreased energy resolution or other performance degradation. Having collimator septa directly above the edges of pixels reduces the chance of a photon impinging at these degraded-performance locations, without decreasing the overall probability of a photon passing through the collimator. As discussed herein, in various embodiments parallel-hole and/or pin-hole collimation may be employed. A controller unit 1030 may control the movement and positioning of the patient table 1020, imaging detectors 1002 (which may be configured as one or more arms), gantry 1004 and/or the collimators 1022 (that move with the imaging detectors 1002 in various embodiments, being coupled thereto). A range of motion before or during an acquisition, or between different image acquisitions, is set to maintain the actual FOV of each of the imaging detectors 1002 directed, for example, towards or “aimed at” a particular area or region of the subject 1010 or along the entire subject 1010. The motion may be a combined or complex motion in multiple directions simultaneously, concurrently, or sequentially as described in more detail herein. The controller unit 1030 may have a gantry motor controller 1032, table controller 1034, detector controller 1036, pivot controller 1038, and collimator controller 1040. The controllers 1030, 1032, 1034, 1036, 1038, 1040 may be automatically commanded by a processing unit 1050, manually controlled by an operator, or a combination thereof. The gantry motor controller 1032 may move the imaging detectors 1002 with respect to the subject 1010, for example, individually, in segments or subsets, or simultaneously in a fixed relationship to one another. For example, in some embodiments, the gantry controller 1032 may cause the imaging detectors 1002 and/or support members 1012 to move relative to or rotate about the subject 1010, which may include motion of less than or up to 180 degrees (or more). The table controller 1034 may move the patient table 1020 to position the subject 1010 relative to the imaging detectors 1002. The patient table 1020 may be moved in up-down directions, in-out directions, and right-left directions, for example. The detector controller 1036 may control movement of each of the imaging detectors 1002 to move together as a group or individually as described in more detail herein. The detector controller 1036 also may control movement of the imaging detectors 1002 in some embodiments to move closer to and farther from a surface of the subject 1010, such as by controlling translating movement of the detector carriers 1016 linearly towards or away from the subject 1010 (e.g., sliding or telescoping movement). Optionally, the detector controller 1036 may control movement of the detector carriers 1016 to allow movement of the detector array 1006 or 1008. For example, the detector controller 1036 may control lateral movement of the detector carriers 1016 illustrated by the T arrow (and shown as left and right as viewed in FIG. 7). In various embodiments, the detector controller 1036 may control the detector carriers 1016 or the support members 1012 to move in different lateral directions. Detector controller 1036 may control the swiveling motion of detectors 1002 together with their collimators 1022. The pivot controller 1038 may control pivoting or rotating movement of the detector units 1014 at ends of the detector carriers 1016 and/or pivoting or rotating movement of the detector carrier 1016. For example, one or more of the detector units 1014 or detector carriers 1016 may be rotated about at least one axis to view the subject 1010 from a plurality of angular orientations to acquire, for example, 3D image data in a 3D SPECT or 3D imaging mode of operation. The collimator controller 1040 may adjust a position of an adjustable collimator, such as a collimator with adjustable strips (or vanes) or adjustable pinhole(s). It should be noted that motion of one or more imaging detectors 1002 may be in directions other than strictly axially or radially, and motions in several motion directions may be used in various embodiment. Therefore, the term “motion controller” may be used to indicate a collective name for all motion controllers. It should be noted that the various controllers may be combined, for example, the detector controller 1036 and pivot controller 1038 may be combined to provide the different movements described herein. Prior to acquiring an image of the subject 1010 or a portion of the subject 1010, the imaging detectors 1002, gantry 1004, patient table 1020 and/or collimators 1022 may be adjusted, such as to first or initial imaging positions, as well as subsequent imaging positions. The imaging detectors 1002 may each be positioned to image a portion of the subject 1010. Alternatively, for example in a case of a small size subject 1010, one or more of the imaging detectors 1002 may not be used to acquire data, such as the imaging detectors 1002 at ends of the detector arrays 1006 and 1008, which as illustrated in FIG. 7 are in a retracted position away from the subject 1010. Positioning may be accomplished manually by the operator and/or automatically, which may include using, for example, image information such as other images acquired before the current acquisition, such as by another imaging modality such as X-ray Computed Tomography (CT), MRI, X-Ray, PET or ultrasound. In some embodiments, the additional information for positioning, such as the other images, may be acquired by the same system, such as in a hybrid system (e.g., a SPECT/CT system). Additionally, the detector units 1014 may be configured to acquire non-NM data, such as x-ray CT data. In some embodiments, a multi-modality imaging system may be provided, for example, to allow performing NM or SPECT imaging, as well as x-ray CT imaging, which may include a dual-modality or gantry design as described in more detail herein. After the imaging detectors 1002, gantry 1004, patient table 1020, and/or collimators 1022 are positioned, one or more images, such as three-dimensional (3D) SPECT images are acquired using one or more of the imaging detectors 1002, which may include using a combined motion that reduces or minimizes spacing between detector units 1014. The image data acquired by each imaging detector 1002 may be combined and reconstructed into a composite image or 3D images in various embodiments. In one embodiment, at least one of detector arrays 1006 and/or 1008, gantry 1004, patient table 1020, and/or collimators 1022 are moved after being initially positioned, which includes individual movement of one or more of the detector units 1014 (e.g., combined lateral and pivoting movement) together with the swiveling motion of detectors 1002. For example, at least one of detector arrays 1006 and/or 1008 may be moved laterally while pivoted. Thus, in various embodiments, a plurality of small sized detectors, such as the detector units 1014 may be used for 3D imaging, such as when moving or sweeping the detector units 1014 in combination with other movements. In various embodiments, a data acquisition system (DAS) 1060 receives electrical signal data produced by the imaging detectors 1002 and converts this data into digital signals for subsequent processing. However, in various embodiments, digital signals are generated by the imaging detectors 1002. An image reconstruction device 1062 (which may be a processing device or computer) and a data storage device 1064 may be provided in addition to the processing unit 1050. It should be noted that one or more functions related to one or more of data acquisition, motion control, data processing and image reconstruction may be accomplished through hardware, software and/or by shared processing resources, which may be located within or near the imaging system 1000, or may be located remotely. Additionally, a user input device 1066 may be provided to receive user inputs (e.g., control commands), as well as a display 1068 for displaying images. DAS 1060 receives the acquired images from detectors 1002 together with the corresponding lateral, vertical, rotational and swiveling coordinates of gantry 1004, support members 1012, detector units 1014, detector carriers 1016, and detectors 1002 for accurate reconstruction of an image including 3D images and their slices. It should be noted that the various embodiments may be implemented in hardware, software or a combination thereof. The various embodiments and/or components, for example, the modules, or components and controllers therein, also may be implemented as part of one or more computers or processors. The computer or processor may include a computing device, an input device, a display unit and an interface, for example, for accessing the Internet. The computer or processor may include a microprocessor. The microprocessor may be connected to a communication bus. The computer or processor may also include a memory. The memory may include Random Access Memory (RAM) and Read Only Memory (ROM). The computer or processor further may include a storage device, which may be a hard disk drive or a removable storage drive such as a solid-state drive, optical disk drive, and the like. The storage device may also be other similar means for loading computer programs or other instructions into the computer or processor. As used herein, the term “computer” or “module” may include any processor-based or microprocessor-based system including systems using microcontrollers, reduced instruction set computers (RISC), ASICs, logic circuits, and any other circuit or processor capable of executing the functions described herein. The above examples are exemplary only, and are thus not intended to limit in any way the definition and/or meaning of the term “computer”. The computer or processor executes a set of instructions that are stored in one or more storage elements, in order to process input data. The storage elements may also store data or other information as desired or needed. The storage element may be in the form of an information source or a physical memory element within a processing machine. The set of instructions may include various commands that instruct the computer or processor as a processing machine to perform specific operations such as the methods and processes of the various embodiments. The set of instructions may be in the form of a software program. The software may be in various forms such as system software or application software and which may be embodied as a tangible and non-transitory computer readable medium. Further, the software may be in the form of a collection of separate programs or modules, a program module within a larger program or a portion of a program module. The software also may include modular programming in the form of object-oriented programming. The processing of input data by the processing machine may be in response to operator commands, or in response to results of previous processing, or in response to a request made by another processing machine. As used herein, a structure, limitation, or element that is “configured to” perform a task or operation is particularly structurally formed, constructed, or adapted in a manner corresponding to the task or operation. For purposes of clarity and the avoidance of doubt, an object that is merely capable of being modified to perform the task or operation is not “configured to” perform the task or operation as used herein. Instead, the use of “configured to” as used herein denotes structural adaptations or characteristics, and denotes structural requirements of any structure, limitation, or element that is described as being “configured to” perform the task or operation. For example, a processing unit, processor, or computer that is “configured to” perform a task or operation may be understood as being particularly structured to perform the task or operation (e.g., having one or more programs or instructions stored thereon or used in conjunction therewith tailored or intended to perform the task or operation, and/or having an arrangement of processing circuitry tailored or intended to perform the task or operation). For the purposes of clarity and the avoidance of doubt, a general purpose computer (which may become “configured to” perform the task or operation if appropriately programmed) is not “configured to” perform a task or operation unless or until specifically programmed or structurally modified to perform the task or operation. As used herein, the terms “software” and “firmware” are interchangeable, and include any computer program stored in memory for execution by a computer, including RAM memory, ROM memory, EPROM memory, EEPROM memory, and non-volatile RAM (NVRAM) memory. The above memory types are exemplary only, and are thus not limiting as to the types of memory usable for storage of a computer program. It is to be understood that the above description is intended to be illustrative, and not restrictive. For example, the above-described embodiments (and/or aspects thereof) may be used in combination with each other. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the various embodiments without departing from their scope. While the dimensions and types of materials described herein are intended to define the parameters of the various embodiments, they are by no means limiting and are merely exemplary. Many other embodiments will be apparent to those of skill in the art upon reviewing the above description. The scope of the various embodiments should, therefore, be determined with reference to the appended claims, along with the full scope of equivalents to which such claims are entitled. In the appended claims, the terms “including” and “in which” are used as the plain-English equivalents of the respective terms “comprising” and “wherein.” Moreover, in the following claims, the terms “first,” “second,” and “third,” etc. are used merely as labels, and are not intended to impose numerical requirements on their objects. Further, the limitations of the following claims are not written in means-plus-function format and are not intended to be interpreted based on 35 U.S.C. § 112(f) unless and until such claim limitations expressly use the phrase “means for” followed by a statement of function void of further structure. This written description uses examples to disclose the various embodiments, including the best mode, and also to enable any person skilled in the art to practice the various embodiments, including making and using any devices or systems and performing any incorporated methods. The patentable scope of the various embodiments is defined by the claims, and may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if the examples have structural elements that do not differ from the literal language of the claims, or the examples include equivalent structural elements with insubstantial differences from the literal languages of the claims. |
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abstract | A method of determining the R-factor for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind. The R-factor is a factor that accounts for the weighted local power influence on a fuel rod. A local R-factor (Ri(z)) is determined for each fuel rod (i) in said bundle and for each of a plurality of levels (z) in an axial direction. The individual axial heat generation profile for a certain fuel rod (i) is taken into account when determining the local R-factors (Ri(z)) for said fuel rod (i). A processor is configured for automatically determining the R-factor. A computer program product, a method of determining the critical power for a bundle of fuel rods, a nuclear energy plant, and a method of operating a nuclear energy plant are also described. |
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abstract | A method for determining a reduced ICI patterns is provided to achieve core monitoring and surveillance and other required functions using fewer ICIs. Candidate ICI patterns having a reduced number of ICIs relative to the existing pattern are first selected according to predetermined selection considerations. After selection, the candidate patterns are evaluated according to predetermined evaluation criteria. |
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claims | 1. An injection system for a neutron poison for an emergency shutdown of a nuclear reactor, the injection system comprising:an absorber liquid having an absorption cross section that captures free neutrons;a pressure vessel containing a propelling fluid being essentially water;an overflow line communicating with a core region of a reactor pressure vessel of a nuclear reactor, the core region being external to said pressure vessel;a reservoir vessel storing said absorber liquid at room temperature, said reservoir vessel connected, via said overflow line, to said pressure vessel; andan outflow line connected to said bottom region of said reservoir vessel; anda shut-off apparatus having a closed position that closes said outflow line and an open position that opens said outflow line thereby causing the absorber liquid to be pushed through said outflow line by the propelling fluid;said reservoir vessel being spatially separated from said pressure vessel, and only said overflow line connecting said reservoir vessel to said pressure vessel such that pressure is equalized between said reservoir vessel and said pressure vessel;said pressure vessel formed with a ceiling region and a bottom region, said pressure vessel including a heating device configured to heat and partially evaporate the propelling fluid in order to form a vapor cushion located in said ceiling region of said pressure vessel, wherein an un-evaporated portion of the propelling fluid defines a liquid portion located below the vapor cushion;said overflow line connecting said bottom region of said pressure vessel to said ceiling region of said reservoir vessel, said overflow line including a riser; andsaid overflow line including a plurality of nozzles discharging the propelling fluid near said ceiling region of said reservoir vessel. 2. The injection system according to claim 1, wherein said bottom region of said pressure vessel has an outflow opening formed therein for removal of said absorber liquid. 3. The injection system according to claim 1, wherein said pressure vessel, said reservoir vessel and said overflow line are pressure-loaded components constructed to operate at a pressure of more than 100 bar. 4. The injection system according to claim 1, wherein said pressure vessel and said heater cooperate such that said propelling fluid is always available at a temperature of more than 300° C. 5. The injection system according to claim 1, wherein said reservoir vessel is dimensioned to hold a quantity of said absorber liquid which is sufficient for shutting down the nuclear reactor. 6. The injection system according to claim 1, wherein said pressure vessel is dimensioned to hold a quantity of said propelling fluid which is sufficient to completely displace said absorber liquid from said reservoir vessel. 7. The injection system according to claim 1, wherein said absorber liquid is an aqueous boron solution stored in said reservoir vessel. 8. The injection system according to claim 1, wherein said pressure vessel, said reservoir vessel and said overflow line are pressure-loaded components constructed to operate at a pressure of approximately 150 bar. 9. The injection system according to claim 1, wherein said pressure vessel and said heater cooperate such that said propelling fluid is always available at a temperature of approximately 340° C. 10. The injection system according to claim 1, wherein said plurality of nozzles discharge the propelling fluid such that the propelling fluid is prevented from mixing with the absorber liquid after the propelling fluid exits said plurality of nozzles. |
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description | The subjects of the present invention are: a casing internal part (a device for the internal arrangement of a casing; frame or rack internal to a casing) suited to the dry intermediate storage of at least one irradiated fuel element; a dry intermediate storage casing equipped with such a casing internal part; and a method for the dry intermediate storage of at least one irradiated fuel element, based on the use of such a casing and such a casing internal part. The casing internal part according to the invention characteristically comprises a hybrid structure within which the thermal function is dissociated from the mechanical integrity function. This is explained later on. In the following description of the invention now claimed, an irradiated fuel element is to be understood to mean an irradiated fuel element or a set of at least two superposed irradiated fuel elements. Such an irradiated fuel element has a vertical structure. The invention has been developed in the context of the dry intermediate storage of irradiated fuel elements. Such irradiated fuel elements are conventionally temporarily stored: either while they await conditioning suited to definitive permanent storage thereof (they are then processed as waste); or awaiting processing/recycling in order to realize their asset value. To date, dry intermediate storage of irradiated fuel elements is performed only on low power irradiated fuel elements. Managing the conduction of heat, in order to ensure cooling, is therefore not a critical issue. Said low-power elements, individually or superposed in twos, are placed in casings. Said casings are not equipped with retaining structures. Said casings are placed in frames, which frames are introduced into the ventilated pits. In a related context, the problems of the dissipation of heat are managed. Thus, for transporting high-power irradiated fuel elements, use is made of special-purpose shipping casks. These are big, heavy, structures in which there are blocks of cast aluminum superposed on each other; said blocks stabilizing the irradiated fuel elements. Within said casks said blocks perform a thermal function and a mechanical integrity function. The technical problem addressed in the context of the present invention has chiefly been that of the dry intermediate storage, individually or in batches, of fuel elements of a higher power (notably each having a power liable to be as high as 5 kW) than those temporarily stored in the prior art. It will be seen later on that the system of the invention is suited to the dry intermediate storage of several fuel elements in one and the same casing. Irrespective of the power involved, it is quite obviously necessary to ensure that the irradiated fuel elements being temporarily stored are correctly cooled in order to avoid any degradation due to excessive temperature, so that the cladding of said fuel elements will maintain its integrity in the long term. The higher the power, the more critical this problem becomes. The good conduction of the heat of the stored irradiated fuel element(s) toward the cooling system of the storage system becomes, in a high-power context, a very important or even crucial parameter. This conduction of heat has to be provided, quite obviously, while at the same time keeping the irradiated fuel element(s) in place, and stable within the temporary storage casing, and this means that the structure of the casing, including its casing internal part (internal arrangement) needs to maintain its geometry throughout the duration of the intermediate storage. Faced with such a technical problem, the inventors are proposing an innovative and high-performance solution based on a novel casing internal part structure, this structure being a tiered hybrid structure with modules made of a material that is a good conductor of heat (advantageously aluminum) with hollowed-out heat sinks (heat dissipators). According to its first object, the present invention therefore proposes novel casing internal parts. The concept of casing internal part is familiar to those skilled in the art. This may also be termed the internal arrangement device of a casing suited in this instance, to the stable positioning of the fuel elements that are to be temporarily stored inside it. The casing internal parts according to the invention are therefore suited to the dry intermediate storage of at least one irradiated fuel element. They have a novel structure. Characteristically, they have a tiered structure which comprises at least two superposed modules; each of said modules, made of a material that is a good conductor of heat, with compartment(s) for accommodating said fuel element(s) in the central part and with at least one hollowed-out heat sink at the periphery (hereafter known as a hollowed-out peripheral heat sink), being arranged on a support plate that is perforated to allow said fuel element(s) through; the perforated support plates of said structure being positioned and assembled by means of a retaining system, with a clearance left between the top of the module of one tier and the perforated support plate of the tier immediately above. As already mentioned, said structure is hybrid: the thermal function is performed by means of modules made of a material that is a good conductor of heat (good conductor (good dissipater) of heat). What is meant in this text by materials which are good conductors of heat are materials the thermal conductivity of which is higher than 150 W·m−1·K−1. Such a definition is commonly accepted by materials specialists. Materials of this type (which have such thermal conductivity) are known, and commonly used at this time; other materials of this type (having such thermal conductivity) are likely to be developed in the future. A person skilled in the art is aware in particular of silver, copper, gold and aluminum. None of these metals is a priori set aside, by way of material of which the modules can be made. However, it will be appreciated that the use of silver or gold presents more particular problems in terms of cost and that the use of aluminum is sure to be preferred. Copper can of course be used, but the reason why aluminum is preferred will be explained hereafter. Copper and aluminum are both good conductors of heat but copper has the disadvantage of being more expensive to source and to use. This is because copper is far more difficult to adapt to suit the desired geometries than is aluminum. Aluminum is therefore the preferred material of which to form the modules of casing internal part according to the invention. It is a good conductor of heat (can undergo a surface treatment to improve its emissivity still further), it can be extruded and what is more it is light weight. However, it is well known that its mechanical strength decreases significantly with increasing temperature; hence the recourse to a support structure (see later). This remark may be entirely valid for some other material that is a good conductor of heat; the function of mechanical integrity is afforded by means of a support structure: perforated support plates and retaining system. Said support structure is obviously made of a material with high mechanical strength, resistant to heat. It is advantageously made of stainless steel. It will be appreciated that other materials such as titanium may be suitable. It will be understood that, in terms of cost, stainless steel is preferred. The compartment(s) of the modules and the opening(s) in the support plates and also the positioning of said modules and support plates are quite obviously suited to the vertical positioning of the fuel elements. Each module comprises at least one compartment in the central part and at least one heat sink at the periphery. Characteristically, said peripheral heat sink is not a solid element but is a hollowed-out part (said peripheral sinks are not solid elements but are hollowed-out parts). Generally, there are several heat sinks arranged at the periphery of each module. In the context of the invention, the thermal function is performed with the amount of material needed being optimized, hence leading to substantial savings, particularly in a mass-production context. According to an alternative form, the modules of the casing comprise a single compartment in the central part. Said single compartment is designed to accept an irradiated fuel element, a priori a high-power fuel element, of a power such that storing it together with another element will present problems (in terms of power). According to another alternative form which is preferred, several compartments are arranged in the structure of each module that makes up the casing internal part. Said casing internal part is therefore designed for the intermediate storage of a batch of fuel elements so that several fuel elements can be temporarily stored at one and the same level . . . and in one and the same casing. This is novel: the prior art combines one casing internal part for one fuel element (or two superposed elements) with one casing, therefore one fuel element (or two superposed fuel elements) per casing. Incidentally, it might be noted that a casing internal part designed for the intermediate storage of n fuel elements (comprising modules with n compartments) is entirely suited to the intermediate storage of m fuel elements: m≦n. The choice of the value of m will be made as a function of numerous parameters particularly as a function of the thermal power involved. In either one of the above alternative forms, each vertical housing created by the alignment of the compartments in the superposed modules is therefore able to accept a single irradiated fuel element or a set of at least two superposed irradiated fuel elements. A functional clearance is left between the top of one module and the perforated support plate situated immediately above. This arrangement allows free (stressless) expansion of this module (advantageously made of aluminum), the expansion of which may exceed that of some of the elements of the retaining system (of the stainless steel through bolt type for example (see later)). Incidentally, it might be noted here that the upper perforated “support” plate does not support any module. It is termed a support plate simply by analogy. It is advantageously identical to the other support plates, although this is not compulsory. The casing internal part according to the invention therefore comprises superposed modules, each of said modules comprising one (or more) compartment(s) suited to receiving the irradiated fuel element(s) and at least one hollowed-out peripheral heat sink which removes the heat from said fuel element(s) toward the wall of the casing (in which said casing internal part is to be placed). Advantageously, the constituent elements of a module—compartment(s)+heat sink(s)—are extruded elements (elements obtained by extrusion). The height of the modules is determined by mechanical calculation and by production feasibility. The heights involved, of the order of one meter per module, make for ease of manufacture of the modules (notably made of aluminum) by extrusion. Such an industrial implementation is particularly advantageous in that it makes it possible to limit the cost of said modules. The number of modules of a casing internal part is obviously tailored to the length of the fuel element(s) to be temporarily stored within said casing internal part. This number is therefore generally dependent on said length and on the manufacturing technologies involved. By way of an indication, the tiered structure of a casing internal part according to the invention may thus comprise 2, 3, 4, 5 or 6 tiers (=2, 3, 4, 5 or 6 superposed modules). In any event, this indication is entirely non-limiting. In general, and more particularly in this advantageous context of extruded elements used to form the modules, the at least one peripheral heat sink (the (only) peripheral heat sink or the peripheral heat sinks) of said module is very advantageously based on an extruded element with or without fin(s) and with radiating surface(s). Said at least one peripheral heat sink can thus be optimized in terms of weight, performance, cost, ease of manufacture, etc. Within a casing of circular cross section there is advantageously a casing internal part which is likewise of circular cross section, and therefore very advantageously at least one peripheral heat sink with curved radiating surface(s) (and with or without fin(s)). The modules of the casing internal parts of the invention are thus advantageously made up of several extruded elements which are joined together by screw fastening or by any other method of assembly (for example welding) and which are positioned precisely, for example using pegs. One of said (pre-existing) extruded elements may thus consist of at least one prefabricated compartment. Starting from said at least one prefabricated compartment it is possible, by adding further extruded elements, to construct the module (with central compartment(s) and peripheral heat sink(s)). For smaller (simple) geometries, the module can be produced as a single extruded part. The wall thickness of the modules, or more specifically, the thickness of the walls of the at least one compartment (of the (only) compartment or of the compartments) and of the heat sink(s) is advantageously optimized to ensure good conduction of heat for the lowest possible mass (notably of aluminum). The geometry and the capacity for the intermediate storage of fuel elements of a casing internal part of the invention is determined by thermal calculations: these are dependent on the power of the fuels to be temporarily stored. The number of compartments is therefore tailored to the power of the fuel to be stored. The shape of said compartments is quite obviously tailored to, or even optimized in relation to, the shape of said fuels to be temporarily stored. A casing internal part of the invention (of circular cross section) thus advantageously comprises 1, 4, 7, 12 or 15 compartments (of square cross section); this corresponds to the optimum filling of a circle with compartments of square cross section. Such an optimum is desired in order to obtain the best capacity for intermediate storage. Incidentally, it might be noted, however, that the casing internal part of the invention and its associated casing are not necessarily of circular cross section, that the compartments are not necessarily of square cross section and it will also incidentally be recalled that the intermediate storage may be carried out with some compartments empty (in which case it is not optimized). Obviously, optimization is advantageously in terms of the number of compartments and the shape of said compartments, on the one hand (with reference to the structure of the casing internal part) and in terms of the number of compartments occupied, on the other hand (with reference to the employed method of intermediate storage in a casing equipped with a casing internal part according to the invention). The person skilled in the art has the ability to make these optimizations. The support structure of a casing internal part according to the invention therefore encompasses the perforated support plates (each module rests on such a plate, which is perforated to allow the fuel element(s) to pass) and the retaining system that holds the set of said plates. In order to perform its function, said support structure is made of a material with very good mechanical strength. It was seen earlier that it is advantageously made of stainless steel. It was also seen earlier that the material of which the modules are made is advantageously aluminum. Very advantageously, aluminum modules are combined in a casing internal part of the invention with a support structure made of stainless steel. The retaining system advantageously comprises through bolts which perform a dual function. Firstly, they position and assemble the various support plates, and secondly, they can be used to transfer load from the casing internal part to the casing in which it is intended to be housed. Very advantageously, these same through bolts are used for (stably) positioning each module on its support plate. In the context of this highly advantageous alternative form, said through bolts therefore perform a triple function. Such through bolts are therefore advantageously made of stainless steel. In the context of a preferred embodiment, through bolts, the length of which corresponds to that of a tier (of a module) are plugged together by screw fastening and have to be stabilized in the same way (by screw fastening) at the bottom and top parts of the casing internal part. There are generally three through bolts involved in each tier. In the structure of the casing internal part according to the invention, the load of a module is reacted via each support plate and via the retaining system (the through bolts in the advantageous alternative form explained hereinabove). As a result, a module of said structure does not experience the weight of the module situated above it and the mechanical stress applied to it is thus lower, and compatible with the mechanical properties of the material (preferably aluminum) of which it is made. The load of the entire casing internal part is to be borne by the casing (the bottom of the casing) within which said casing internal part is intended to be positioned. It was indicated hereinabove entirely non-limitingly that the casing internal part according to the invention may notably comprise 2, 3, 4, 5 or 6 tiers (2, 3, 4, 5 or 6 superposed modules). Entirely non-limitingly, it may also be indicated here that the compartments of the modules are of square, hexagonal or circular cross section. In any event, said cross section needs to be suitable for said compartments to accommodate the fuel elements. The openings in the support plates have likewise to be suitable. Said compartments and said openings are generally the same shape and the same size. The above is illustrated non-limitingly by the attached figures. A second subject of the present invention is a casing for the dry intermediate storage of at least one irradiated fuel element. Characteristically, said casing contains a casing internal part like those described hereinabove (first subject of the present invention) and comprises a bottom capable of withstanding the mass of said casing internal part and that of said at least one fuel element intended to rest thereon (that of the (only) fuel element intended to rest thereon or that of the fuel elements intended to rest thereon). What is meant by the bottom of said casing is the structure thereof on which the casing internal part and the lower part of said at least one stored fuel element rest once said fuel element has been passed through the lower support plate (the support plate of the first tier of the first module) of the casing internal part. Opposite said bottom, said casing can obviously be fitted with a lid so as to form a sealed casing. Advantageously, said casing of the invention and its casing internal part have cross sections of substantially the same dimension(s) (of substantially the same diameter assuming the most frequent case where said casing and casing internal part are right cylinders of circular cross section). In any event, said casings and casing internal parts advantageously have the shape of right cylinders. The clearance between the casing of the invention and its casing internal part is advantageously as small as possible, with a view to optimizing the dissipation of heat. However, it must allow said casing internal part to be introduced into and positioned in said casing, i.e. it must allow the modules to be fitted into the casing. The casing internal part is advantageously stabilized in the casing. Thus, retaining systems suited to preventing said casing internal part (or the modules thereof) from moving around inside the casing, more particularly if said casing is knocked over, may thus be provided. The third subject of the present invention relates to a method for the dry intermediate storage of at least one irradiated fuel element (a single) irradiated fuel element or irradiated fuel elements. Conventionally, said method comprises: the stable conditioning of said at least one irradiated fuel element in a casing within which the heat released by said at least one element is dissipated with no adverse effect on the structure of said at least one element, the intermediate storage of said sealed casing in a vertical pit cooled by the circulation of air. Characteristically, the casing used in the context of said method is a casing of the invention (second subject of the present invention), that is to say a casing equipped with a casing internal part of the invention (first subject of the present invention). Such a casing optimizes the intermediate storage by allowing the maximum of fuel (the maximum power in fact) to be stored temporarily in the smallest amount of space. According to some particularly preferred alternative forms of embodiment: the method of the invention is implemented using a casing containing several fuel elements (at one and the same level); and/or said method of the invention involves the intermediate storage of at least one high-power fuel element (of a power up to 5 kW; for example from a light water reactor). The method of the invention is not restricted to its particularly preferred alternative forms of implementation. Thus, it may be as suited to the intermediate storage of a single high-power fuel element as it is to the intermediate storage of a single or of several low or lower power fuel elements. If it is capable of heavy duty then it is also capable of light duty. It will be recalled that the structure of the casing involved has been optimized for the maximum removal of released heat. FIG. 1 therefore shows a casing internal part 1 arranged inside a casing 10. Said casing internal part 1 and casing 10 are of cylindrical shape (have the shape of right cylinders of circular cross section). The casing internal part 1 comprises four tiers. Its structure comprises four modules M1, M2, M3 and M4 each positioned on a perforated support plate 4. A clearance 3 is left between the top of each module (such a clearance allows said module, notably made of aluminum, free expansion) and the support plate of the module of the tier above. The compartments 2a (central) and 2b (1st ring) of the four modules M1, M2, M3 and M4 and the corresponding openings 2 (central) and 2′ (1st ring) in the support plate 4 (see FIG. 2) are obviously made to coincide so as to generate within the structure of the casing internal part 1 vertical housings for the fuel elements. The through bolts 5 are used, in the context of the advantageous embodiment depicted: to position and assemble the support plates 4 with respect to one another, to transfer load from the casing internal part 1 to the casing 10. A fixing screw for a lower through bolt 5 positioned on the bottom part 9 of the casing 10 has been depicted as 5′ and a fixing nut for fixing an upper through bolt 5 to the upper “support” plate 4 of the casing internal part 1 has been depicted as 5″. It will thus be appreciated that load from one module M1 is reacted via each support plate 4 and the through bolts 5 and that in this way the load of the entire casing internal part 1 is borne by the casing 10, to position each of the modules M1, M2, M3 and M4 on its respective support plate 4 (see FIG. 2). FIG. 2 therefore shows the lower module M1 (identical to the other modules M2, M3 and M4) and the lower support plate 4 (the support plate for said module M1) and the intermediate support plate 4 (support plate for the module M2 formed between said modules M1 and M2). The module M1 comprises seven compartments: the central compartment 2a, prefabricated, and six ring compartments 2b obtained by construction, following the fitting of the peripheral heat sinks 3a and 3b. These heat sinks have hollowed-out structures with two types of geometry. They are shown in said FIG. 2 and in FIGS. 3A and 3B. The heat sinks 3a, of which there are four, are corner heat sinks; the heat sinks 3b, of which there are two, are T-shaped. Said heat sinks 3a and 3b have geometries that are optimized in relation to their performance, for the weight/performance/ease of construction compromise. These are therefore hollowed-out structures obtained by extrusion, with fin 31 and curved radiating surface 32 which are particularly beneficial with respect to thermal radiation. In the structure of said heat sinks 3a and 3b there are housings 8 for the through bolts 5. Said through bolts 5 have their ends 5b and 5a threaded so that they respectively collaborate at the bottom with the fixing screws 5′ (if M1 is the module of the first tier of the casing internal part 1) and, at the top, with the threaded lower end 5b of the through bolt 5 of the module M2 (of the tier above, not depicted). The end 5a of the through bolt 5 of the module M4 is itself called upon to collaborate with the fixing nut 5″ (refer to FIG. 1 and the above description of said figure). The central compartment 2a is positioned at the center of the other elements (heat sinks 3a and 3b) using centering pegs 6. It will be realized, on studying FIGS. 3A, 3B and 3C, that, within the wall thickness of said central compartment 2a there are in fact housings 6′ for said centering pegs 6, provided on the heat sinks 3a and 3b. Said heat sinks 3a and 3b (six peripheral elements) are joined together by screw fastening. The screws involved are depicted schematically as 7. FIG. 2 clearly shows that, according to the advantageous embodiment depicted, the through bolts 5 provide the stable positioning of the module M1 (of its constituent compartments 2a and 2b) on the (lower) support plate 4. The support plates 4 (lower plate, support for the module M1, and upper plate, support for the module M2) are obviously perforated to allow the fuel elements that are to be stored in the casing 1 to pass, that is to say that said plates 4 have openings 2 and 2′ which, respectively, are in register with the compartments 2a and 2b. Said plates 4 also have holes 41 for the passage of the through bolts 5. These holes 41 more specifically, in the case of the lower plate, are where the lower ends of the through bolts 5 (depicted) are secured to the fixing screws 5′ and where, in the case of the upper plate, the upper ends of said through bolts 5 are secured to the lower ends of the through bolts 5 which have not been depicted in FIG. 2 (but which are depicted in FIG. 1) of the module M2 of the tier above. Said plates 4 finally have slots 42 providing communication between the tiers of the casing internal part 1 positioned inside the casing 10. From studying FIGS. 4 and 5 it will be appreciated that, according to the second embodiment depicted, the casing internal part of the invention comprise fifteen compartments: three so-called central compartments 20a, ten compartments said to be of a first ring 20b and two said to be of a second ring 20c. The set of compartments is produced from simple pieces of section piece, obtained by extrusion then assembled by spot welds at their ends. In the case of the compartments located around the periphery of the set, devices, also assembled by spot welds, are installed to join the compartments together and ensure correct positioning of the set in relation to the peripheral heat sinks 30. The module depicted therefore is essentially the result of assembling a first component comprising the compartments 20a and certain walls of the compartments 20b and 20c and the four peripheral heat sinks 30. Said heat sinks 30 are hollowed-out extruded elements. They have fins 301 and curved radiating surfaces 302 and the passages 80 for the through bolts (not depicted). The fixing screws that join said heat sinks 30 together have been depicted as 70. On the support plate 40 (FIG. 5) we again find: the openings 20, 20′ and 20″ which are respectively in register with the compartments 20a, 20b and 20c; the holes 401 (in register with the passages 80) for the passage of the through bolts; the slots 402 providing communication between the tiers. The invention is finally illustrated through the following two examples. A casing internal part 1 like the one depicted in FIGS. 1 and 2 is produced. Said casing internal part comprises 4 aluminum modules 1080 mm tall, each resting on a perforated horizontal plate made of stainless steel. Each module is therefore made of 7 elements, with the three different geometries: a central compartment, two T-shaped peripheral heat sinks, and four corner peripheral heat sinksand comprises 7 compartments (see FIGS. 1, 2, 3A, 3B and 3C and the above description thereof). Said casing internal part has been positioned in a stainless steel casing with a clearance of 3 mm on the radius. The casing internal part/casing arrangement is as depicted in FIG. 1. The casing obtained can therefore accept up to seven irradiated fuel elements, notably of the PWR (pressurized water reactor) type. Such a casing, with 7-comparatment casing internal part, is perfectly suited to the dry intermediate storage of PWR 900 irradiated fuel elements; the thermal power of such an irradiated fuel element can be as high as 1.6 kW. Such a power of 1.6 kW corresponds to the residual thermal power of a fuel element of 900 MW PWR type stored 5 years after entering the reactor. Such a casing is also able to accommodate sets (of appropriate overall height) of superposed irradiated fuel elements. A casing internal part is produced with modules of the kind depicted in FIG. 4 and with support plates of the kind depicted in FIG. 5 (casing internal part “analogous” to those depicted in FIG. 1). Said casing internal part comprises four aluminum modules 1080 mm tall, each resting on a perforated horizontal plate made of stainless steel. Each module therefore comprises: a set making up the body of 15 compartments, and four hollowed-out peripheral heat sinks completing the belt around the peripheral compartments. Said casing internal part has been positioned in a stainless steel casing with a clearance of 3 mm on the radius. The casing internal part/casing arrangement is of the kind depicted in FIG. 1. The casing obtained is therefore able to accommodate up to 15 irradiated fuel elements, notably BWR (boiling water reactor) fuel of 8×8 type. such a casing, with a 15-compartment casing internal part, is perfectly suited to the dry intermediate storage of BWR irradiated fuel of 8×8 type; the thermal power of such an irradiated fuel element may be as much as 0.6 kW. Such a casing is also able to accommodate sets (of suitable overall height) of superposed irradiated fuel elements. |
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052895106 | description | Preferred embodiments of the invention have been chosen for the purpose of illustration and description. The preferred embodiments illustrated are not intended to be exhaustive nor to limit the invention to the precise forms shown. The preferred embodiments are chosen and described in order to best explain the principles of the invention and their application and practical use to thereby enable others skilled in the art to best utilize the invention in various embodiments and modifications as are best adapted to the particular use contemplated. DETAILED DESCRIPTION OF THE INVENTION As described in assignee's co-pending U.S. Patent application entitled, "On-Line Tritium Production Monitor", Ser. No. 07/924,125, filed Aug. 3, 1992, the detection of nuclear activity is achieved by coating an scintillation optical fiber with a nuclear reactive material that produces two particles with active energy upon being contacted by a nuclear particle or ray. One of these energetic particles is received by the optical fiber for producing a pulse of scintillation light therein which is transmitted via a photomultiplier tube to a pulse counting circuit. The coating on this optical fiber can be formed of enriched .sup.6 Li so that upon being contacted by a neutron a nuclear .sup.6 Li(n.alpha.)T reaction is induced in the coating for producing energetic .alpha. particles and tritons which are ejected from the coating in directions 180.degree. apart from one another with one of the energetic particles entering the optical fiber to provide therein the pulse of scintillation light. Inasmuch as the scintillation optical fibers, the coatings on the optical fibers, and the energetic particle producing reactions described this co-pending U.S. patent application generally correspond to the type of optical fibers, the planar layer of nuclear reactive materials contained between the planar arrays of the side-by-side optical fibers oriented in an X-Y plane, and the scintillation light producing reactions in the planar layers of nuclear reactive material as employed in the present invention, especially in the embodiment of FIGS. 1-4 of the present invention, this co-pending U.S. patent application is incorporated herein by reference. However, it will appear clear that the embodiments of the nuclear reactive detectors of the present invention as shown in FIGS. 1-4 and FIGS. 5 and 6 differ considerably from the coated optical fiber arrangement described in assignee's aforementioned patent application which provides of position sensitivity in only one direction. In the present invention the X-Y orientation of the planar arrays of the side-by-side optical fibers with the planar layer of nuclear reactive material therebetween as in the embodiment of FIGS. 1-4 or in contact with one of the X-Y arrays of optical fibers as in the embodiment of FIGS. 5 and 6, provides position sensitivity in two-dimensions, i.e., in both the X and the Y direction, for determining essentially the exact X-Y coordinates of each of a plurality of sequentially occurring nuclear events. A preferred utilization of the nuclear reaction detectors of the present invention is directed to the position sensitive detection of nuclear reactions between neutrons and a planar layer of .sup.6 Li with submillimeter resolution in two dimensions. Thus, while the description below is primarily directed to neutron detectors having position sensitivity in two directions, it is to be understood that the planar layer of .sup.6 Li can be replaced by another nuclear reactive material with or without a phosphor dopant. For example, the planar layer of nuclear reactive material can be formed of a .sup.10 B compound so that the location of .sup.10 B (n.alpha.) reactions can be determined in a position sensitive manner in two dimensions, or of a .sup.235 U compound for a position sensitivity determination of fission density distribution such as in a reactor or a storage vault for fissile material, or of a .sup.238 U compound for yielding the two dimensional locations of fast fission occurrences in a nuclear reactor. Also, if desired the planar layer of the .sup.6 Li used for neutron detection can be replaced by compounds of elements other than .sup.6 Li that are capable of reacting with neutrons to produce one or more energetic particles for providing pulses of light in the optical fibers of the optical fiber arrays. With reference to the embodiment of the detector illustrated in FIGS. 1-4, the detector 10 incorporates a nuclear reaction sensing assembly 12 for providing the position sensitive detection of nuclear events in two dimensions. This nuclear reaction sensing assembly 12 is shown formed of a first plurality of commercially available glass or plastic scintillation optical fibers 14 arranged side-by-side in a single plane to form a first planar array 16 of optical fibers which, for the purpose of this description, is shown oriented in the X direction of a two-dimensional X-Y plane. A second plurality of optical fibers 18 similar in construction and length to the optical fibers 14 are arranged side-by-side in a single plane to form a second planar array 20 of optical fibers oriented in the Y direction or at 90.degree. to the optical fiber array 16. A planar layer 22 of .sup.6 Li, preferably .sup.6 LiF, as best shown in FIGS. 2-4, is positioned between the two planar optical fiber arrays 16 and 20 with this two-dimensional layer 22 of .sup.6 LiF extending a sufficient distance in both the X and the Y directions so as to be between and in contact with surface regions of the optical fibers 14 and 18 at all of the nexuses formed by the crossing optical fibers. For example, if two optical fiber arrays 16 and 20 contained ten optical fibers, then the planar layer 22 of .sup.6 LiF would be of an area sufficiently large so as to be in contact with the optical fibers 14 and 18 at the one hundred nexuses or cross-over points. Inasmuch as the nuclear reaction sensing assembly 12 can be readily fabricated with various X and Y dimensions, it is expected that this assembly 12 of optical fiber arrays can be positioned in any appropriate location within a nuclear reactor or in an experimental configuration wherein neutrons or other nuclear particles or rays are produced. The planar layer 22 of .sup.6 LiF, preferably with the .sup.6 Li enriched to 95%, positioned between the two optical fiber arrays 16 and 20 is of a maximum thickness less than about 10 microns so that when a neutron contacts the layer 22 the resulting nuclear reaction (.sup.6 Li(n.alpha.)T) will eject the energetic .alpha. particle and the triton from the layer 22 in opposite directions. These energetic particles simultaneously enter a single optical fiber in each optical fiber array 16 and 20 to produce a pulse of scintillation light in only these two optical fibers. These light pulses propagate to the end of the optical fibers 14 and 18 which are coupled to a suitable reaction-position signal producing circuit 24 which includes light amplifying means such as a photomultiplier tubes 26 and 28 or an array of photo diodes (not shown) and a suitable coincidence processor counting circuit 30 for providing an output signal indicative of the location of the nuclear reaction in a two-dimensional plane. The photomultiplier tubes 26 and 28 used to individually receive the light pulses from the optical fibers 14 and 18 may be provided by using an array of individual photomultiplier tubes as shown in the aforementioned publication or by using photomultiplier tubes having position sensitivity such as provided by crossed wire anodes. One such commercially available photomultiplier tube with multiple position sensitive locations for receiving signals is the "Hamamatsu R2486 Series Type" photomultiplier tube as described in the publication "Position-Sensitive Photomultiplier Tubes with Crossed Wire Anodes, R2486 Series", Hamamatsu Technical Data, August 1989, Hamamatsu Corporation, 360 Foothill Road, Ridgewater, N.J. Such position sensitive photomultiplier tubes using crossed wire anodes as well as arrays of individual photo diodes or photomultiplier tubes, such as described in the aforementioned publication, are capable of providing over 200 useful locations of signal. In accordance with the present invention, each optical fiber 14 of the planar array 16 oriented in the X direction would be attached to a single photomultiplier tube or to a selected location on the position sensitive photomultiplier tube 26 while each optical fiber 18 of the planar 20 oriented in the Y direction would be coupled to a selected location on the position sensitive photomultiplier tube 28. By employing such photomultiplier tubes 26 and 28, the X-Y position of each nuclear event may be accurately determined by using X and Y location position analyzers 32 and 34, which are respectively coupled to the photomultiplier tubes 26 and 28 and are used for determining the exact X and Y positions on the photomultiplier tubes 26 and 28 receiving the light pulse generated from a single nuclear reaction. Such a position analyzer is commercially available for use with the aforementioned "Hamamatsu R2486 Series Type" photo amplifier tube from the Hamamatsu Corporation noted above as "Position Analyzer C1816." The coincidence counting circuit 30 can be provided by a collection of 1, 20 coincidence circuit modules such as the ORTEC Model OR/102B, available from EG&G ORTEC, Oak Ridge, Tenn., or an equivalent counting module. For a relatively large number of optical fiber arrays or optical fibers in each array the use of conventional coincident counting techniques and systems may be more space efficient. The coincidence counting circuit 30 is coupled to the X and Y position analyzers 32 and 34 for associating the X and Y fiber locations where the nuclear reactions occur as a function of time. Thus, by using coincidence processing techniques, the timed association of the light pulses from specific X and Y optical fiber locations is used to accurately derive the exact location of the X-Y intersection of the two optical fibers 14 and 18 where each nuclear reaction occurred. A suitable data processor such as a computer processor 35 is shown coupled to the coincidence processor 30 of analyzing the signals therefrom for data indicative of the X-Y locations of the nuclear events. The optical fibers 14 and 18 employed in the optical fiber arrays are preferably fast scintillation plastic fibers formed of a polymer such as polystyrene. Such plastic scintillation optical fibers are available from Bicron Corporation, Newberry, Ohio 44065. These commercially available plastic scintillation fibers can produce pulses with a total pulse width of less than 20 nsec. Alternatively, the optical fibers 14 and 18 can be formed of glass such as fused silica. The glass or plastic optical fibers utilized in each optical fiber array 16 and 20 may be of any desired length with the layer 22 of nuclear reactive material fully contacting all the intersections or cross over points provided by the optical fibers in both optical fiber arrays 14 and 20 so as to provide a two-dimensional nuclear reaction detecting assembly of a selected size. Moreover, the number of optical fibers used in the optical fiber arrays 16 and 20 is unlimited by the positions at the photomultiplier tubes 26 and 28 since more than one such photomultiplier tube can be used with any optical fiber array containing greater than 200 optical fibers. As shown in FIG. 2, the optical fibers are of a generally rectangular configuration so that when they are placed in an abutting side-by-side relationship, a greater area of each fiber is contacted by the layer 22 of nuclear reactive material at each nexus of the optical fibers. Alternatively, as shown in FIG. 3, the optical fibers 14 and 18 may be of a cylindrical cross section. In using either rectangular or cylindrical optical fibers, the surface area of the fibers in contact with the layer 22 of nuclear reactive material at the nexuses of the optical fibers is sufficient to provide accurate position sensitivity of the nuclear event with submillimeter resolution in two dimensions. The position sensitivity of the nuclear reaction detector system 12 of the present invention is limited only by the cross-sectional thickness of the glass or plastic optical fibers which can be readily formed with a cross-sectional thickness or diameter of less than 0.1 mm so that spatial resolution of the sequential nuclear reactions can be provided in the X-Y plane with a point accuracy to less than 0.1 mm. As shown in FIG. 4, several pairs of optical fiber arrays disposed in the X and Y directions are placed together in the Z direction to provide a stacked assembly 36 of pairs of X and Y optical fiber arrays 16 and 20 with a layer 22 of nuclear reactive material contained between each X-Y pair of optical fiber arrays in the stacked assembly 36 so as to provide a level of X-Y sensitivity over a greater volume of a nuclear reactor, fissile storage vault, or experimental apparatus than can be provided by a single X and Y construction of optical fiber arrays as shown in FIGS. 1-3. When using such a stacked assembly 36, the individual optical fibers in the stacked pairs of optical fiber arrays that are located in common locations in the X direction and the Y direction are coupled at the same point to the photomultiplier tubes 26 and 28, respectively, so that a nuclear event occurring in any of the paired X-Y optical fiber arrays in the stacked assembly 36 will provide an indication or signal indicative of the X-Y position of the nuclear event to the coincidence counting circuit 30. With reference to the embodiment of the detector of the present invention illustrated in FIG. 5, the nuclear reaction sensing assembly 40 provides for the position sensitive detection of nuclear events in a two-dimensional X-Y plane as in the detector embodiment of FIGS. 1-4. In this embodiment the nuclear reaction sensing assembly 40 is shown formed of two planar arrays 42 and 44 of wave length shifting optical fibers with one array oriented in the X-direction and with the other array oriented in the Y-direction. Each array 42 and 44 of optical fibers is formed of a plurality of side-by-side wave length shifting optical fibers 46 and 48, respectively. The planar layer 50 of nuclear reactive material is provided by nuclear reactive materials of the type previously described for use in the nuclear reaction sensing assembly 12 of the detector embodiment of FIGS. 1-4 except for being doped with a suitable phosphor. This planar layer 50 of the phosphor-doped nuclear reactive material is shown placed over the optical fiber array 42 in a contacting relationship with outer surface regions of the optical fibers 46. Also, as shown, the planar layer 50 of the phosphor-doped nuclear reactive material is coextensive with the lengths of the optical fibers 46 in the optical fiber array 42 that are in a contacting nexus-forming relationship with the optical fibers 48 in optical fiber array 44. This embodiment of the detector is used for the position sensitive detection of nuclear reactions where only one of the two particles produced from the reactions is provided with sufficient active energy to escape from the layer of nuclear reactive material. Thus, in order to assure that the X-Y coordinates of each nuclear reaction can be determined when only one energetic particle is produced by each nuclear reaction, the nuclear reactive material is doped with a phosphor which reacts with the energetic particle to produce a light pulse in the layer 50 at the nuclear reaction site. This light pulse will be emitted from the nuclear reactive material and virtually simultaneously pass into a single optical fiber 46 in the optical fiber array 42 and a single optical fiber 46 in the optical fiber array 44 at the nexus of these two optical fibers. The light pulses provided in the single wave length shifting optical fibers in each array 42 and 44 are transmitted therethrough to a reaction-position signal producing circuit corresponding to previously described circuit 24 used in connection with the detector embodiment of FIGS. 1-4. The wave length shifting fibers 46 and 48 are of the type capable of transmitting light such as at frequencies in the green or blue light spectrum and are of rectangular or round configurations with cross sections or diameters less than 1 mm, preferably less than 0.1 mm. Wave length shifting optical fibers having such features are available from the above mentioned Bicron Corporation. The nuclear reactive material is doped with a suitable phosphor such as calcium tungstate, magnesium tungstate, zinc silicate, zinc sulfide, cadmium tungstate, and cadmium borate. The concentration of the phosphor dopant in the layer of nuclear reactive material provided by a compound such as .sup.6 LiF or glass or plastic scintillators containing .sup.235 U, .sup.10 B, or .sup.238 U is sufficient to assure that an adequate distribution of phosphor is obtained throughout the layer of nuclear reactive material to be contacted by and react with the energetic particle produced from each nuclear reaction. Normally, a concentration of the phosphor dopant in the range of about 100 ppm to about 2 percent by volume is adequate to achieve this objective. Inasmuch as the light pulse is produced within the layer 50 of nuclear reactive material the thickness of this layer is not as critical to the successful operation of the detector embodiment of FIG. 5 as it is to that of the detector embodiment of FIGS. 1-4. In this nuclear reaction sensing assembly 40 the layer 50 of nuclear reactive material can be of a thickness of up to about one half the diameter or cross section of the fibers in order to maintain the spatial resolution and assure that the light will escape from the layer 50 of nuclear reactive material. With reference to FIG. 6, the detector embodiment of FIG. 5 is modified by placing several pairs of the wave length shifting optical fiber arrays 42 and 44 together in the Z-direction to provide a stacked assembly 52 of pairs of X and Y optical arrays. A layer 50 of the phosphor-doped nuclear reactive material is placed between each of the adjacent pairs of optical fiber arrays 42 and 44. As with the detector embodiment of FIG. 4 the individual optical fibers in the stacked pairs of optical fiber arrays that are located in common locations in the X direction and the Y direction are coupled at the same point to the photomultiplier tubes 26 and 28, respectively, so that a nuclear event occurring in any of the paired X-Y optical fiber arrays 42 and 44 in the stacked assembly 52 will provide an indication or signal indicative of the X-Y position of the nuclear event to the coincidence counting circuit. By employing the nuclear reaction detecting assembly 12 of the present invention and utilizing a coincidence counting circuit operatable at relatively high rates of greater than about 10.sup.6 counts per second, the number of instruments in the signal producing circuit 24 required for the position sensitive measurements of nuclear events by using techniques described in the aforementioned publication is reduced by a factor of n/2 wherein n.sup.2 is the number of positions of interest in the X-Y plane. The techniques described in the aforementioned article when used in 10.times.10 array requires n.sup.2 electronics whereas the present invention only requires the use of 2n electronics. Thus, for a 10.times.10 array defined by ten optical fibers in each optical fiber array, one-hundred individual photomultiplier tubes or tube positions would be required using previously known techniques whereas in the present invention a 10.times.10 array with ten optical fibers in each optical fiber array 16 and 20 of FIGS. 1-4 or 42 and 44 of FIGS. 5 and 6 would require only twenty photomultiplier tubes or tube positions which could easily produce pulses at 10.sup.7 rates. It will be seen that by using the nuclear reaction detectors of the present invention, the X and Y locations of a relatively high number of sequentially occurring nuclear events can be readily determined with submillimeter resolution in a manner considerably more efficient than heretofore available. Further, the determination of the X and Y locations of the nuclear events with a resolution of less than 0.1 mm is feature of the present invention not previously obtainable. |
description | The present invention relates to a cushioning body attached to a cask. At a final stage of a nuclear fuel cycle, spent fuel (a radioactive material) contains a highly radioactive material and is cooled in a cooling pit of a nuclear power plant for a predetermined period of time because of the need of thermally cooling the spent fuel after the occurrence of decay heat to the radioactive material. The spent fuel is then stored in a cask that is a container, transported to an intermediate storage facility, a reprocessing plant, or the like by a transport method such as a truck, and stored or processed at the facility or the like. The spent fuel is hereinafter referred to as “spent fuel assemblies”, because the spent fuel is normally stored in the cask that serves as a container in a state of fuel assemblies. Generally, the cask is configured to include a main body having one end that is open and the other end that is closed, resin that is provided on the outer circumference of this main body and that is a neutron shield, an outer shell provided on the outer circumference of this resin, a basket that is arranged within the main body and that accommodates the spent fuel assemblies, and a primary lid and a secondary lid that are fixed to one end of the main body. As described above, the cask stores therein the spent fuel assemblies, and therefore it is necessary to keep the cask shielded, subcritical, and hermetically sealed. Therefore, during the transport of casks, the casks are protected with the both ends of each cask covered with a cask cushioning body. Even if the casks are subjected to drop or the like, the shielding, subcritical, and the hermetically sealing performances of the casks are prevented from being degraded. Conventionally, there is an example of such cask cushioning body disclosed in Patent Literature 1. This conventional cask cushioning body includes a housing having a hollow interior, and a cushioning material (such as wood) with which the interior of the housing is filled, and partitions made of metal plates that independently partition the cushioning material are provided in the interior of the housing. The partitions are provided horizontally to the height direction of the housing. Patent Literature 1: Japanese Patent Application Laid-open No. 2005-321304 In the cask cushioning body described in Patent Literature 1 mentioned above, the steel housing serves as the outer shell of the cushioning body, and is abutted on the ends of the cask and arranged on the surface of the cask in a state of being fixed to the cask. However, the conventional cask cushioning body has the following problems. The impact generated when the cask drops deforms the impact absorber such as the wood charged into the cushioning body and the impact absorber absorbs the impact. Depending on drop conditions, the impact even deforms the housing, the deformed housing abuts on the lid or the side surface of the cask, a local load is applied to the cask, and the impact is thereby transferred to the cask. As a result, it is impossible to sufficiently reduce the load on the cask, particularly on the lid in all assumable cask drop events. While a metal gasket or the like keeps the hermetically sealing performance of the lid, there is a probability that the impact exceeds a follow-up deformation permissible amount and deforms the metal gasket, which makes it impossible to keep a sufficient hermetically sealing performance. The present invention has been achieved to solve the above problems, and an object of the present invention is to provide a cask cushioning body that can exhibit an improved impact absorbing performance in various assumable drop events. According to an aspect of the present invention, a cask cushioning body attached to both ends of a cask storing therein spent fuel and absorbing an impact applied to the cask, includes: an end-surface side member in which a plurality of plates made of steel are formed at a distance between plate surfaces of the plates that face each other, and in which the plate surfaces of the plates are arranged along an end surface of the cask; and a circumferential-surface side member that forms a cylindrical body made of the steel, one end of which is connected to a periphery of the end-surface side member, and that is arranged along an end-portion outer-circumferential surface of the cask. An impact absorber that absorbs the impact by deforming is provided outside of the end-surface side member and the circumferential-surface side member. According to the cask cushioning body, at the time of a drop or an impact, the end-surface side member made of steel does not have a large deformation while the outer impact absorber deforms and absorbs the impact. Accordingly, a large deformation does not cause the end-surface side member to abut on the lid and an impact load that possibly prevents the hermetically sealing performance of the cask from being kept is not transferred to the cask. Therefore, it is possible to reduce a local load on the end surface of the cask. Furthermore, at the time of the drop or the impact, the circumferential-surface side member does not have a large deformation while the outer impact absorber deforms and absorbs the impact. Further, on the lid side, it is possible to prevent a positional deviation of the lid, and to keep the hermetically sealing performance by the metal gasket arranged between the lid and the main body. Accordingly, a large deformation does not cause the circumferential-surface side member to abut on the end-portion outer-circumferential surface of the cask and the impact load that possibly prevents the hermetically sealing performance of the cask from being kept is not transferred to the cask. Therefore, it is possible to reduce a local load on the end-portion outer-circumferential surface of the cask. Advantageously, in the cask cushioning body, the end-surface side member includes a plurality of end-surface reinforcing members provided to connect the plates to each other. The cask cushioning body can realize a reduction in the load on the end surface of the cask because the end-surface reinforcing members make it more difficult to deform the end-surface side member. Advantageously, in the cask cushioning body, an opening is formed in a central portion of the plates other than at least one plate of the end-surface side member, and the end-surface side member includes a plurality of central reinforcing members in a region opened by the opening between the plate and the other plate. The cask cushioning body can realize a reduction in the local load on the central portion of the end surface of the cask by the central reinforcing members when a bar member penetrates the impact absorber. Advantageously, in the cask cushioning body, the impact absorber is inserted into a region surrounded by the central reinforcing members and the plate provided with the central reinforcing members. The cask cushioning body can ensure attaining an impact absorbing effect of the impact absorber because a positional deviation of the impact absorber is prevented when the impact absorber absorbs the impact by allowing portions in which the central reinforcing members are provided to hold the impact absorber. Advantageously, in the cask cushioning body, the circumferential-surface side member includes a protruding portion in which peripheries of the plates in the end-surface side member protrude outward of the cylindrical body over an entire circumference of one end of the cylindrical body, a flange portion protruding outward over an entire circumference of other end of the cylindrical body, and a plurality of circumferential-surface reinforcing members connecting the protruding portion to the flange portion, and arranged on an outside surface of the cylindrical body. The cask cushioning body can realize a further reduction in the local load on the end-portion outer-circumferential surface of the cask because the circumferential-surface reinforcing members make it more difficult to deform the circumferential-surface side member. Advantageously, in the cask cushioning body, the impact absorber is inserted into a region surrounded by the protruding portion, the flange portion, and the circumferential-surface reinforcing members. The cask cushioning body can ensure attaining the impact absorbing effect of the impact absorber because the positional deviation of the impact absorber is prevented when the impact absorber absorbs the impact by allowing the protruding portion, the flange portion, and the circumferential-surface reinforcing members to hold the impact absorber. Advantageously, in the cask cushioning body, a region of the end-surface side member in which the plates face each other covers a bolt for fixing a lid of the cask. The cask cushioning body can ensure maintaining a fastening force of the lid by reducing the concentration of a load on the bolt for fixing the lid of the cask and by preventing damage of the bolt, and can prevent the positional deviation of the lid when the bar member penetrates the impact absorber. Therefore, it is possible to keep the hermetically sealing performance by the metal gasket arranged between the lid and the main body. Advantageously, in the cask cushioning body, the region of the end-surface side member in which the plates face each other covers an air-supply/exhaust, water-feed/drainage or pressure monitoring unit provided on the lid. The cask cushioning body can keep the hermetic sealing performance by covering of the air-supply/exhaust, water-feed/drainage or pressure monitoring unit by reducing the concentration of a load on the air-supply/exhaust, water-feed/drainage or pressure monitoring unit provided on the lid of the cask. Advantageously, in the cask cushioning body, an impact absorbing material that absorbs the impact by deforming is filled in the end-surface side member between regions in which the plates face each other. The cask cushioning body can absorb the impact (particularly, a high frequency impact) applied to the plate when the bar penetrates the impact absorber. Advantageously, in the cask cushioning body, the cask cushioning body is formed so that a size between an outer circumferential surface of the lid constituting the cask and an inner circumferential surface of the circumferential-surface side member is larger than a size between an outer circumferential surface of a main body of the cask to which the lid is fixed and the inner circumferential surface of the circumferential-surface side member. The cask cushioning body can reduce the impact applied to the lid and can further prevent the circumferential-surface side member from abutting on the lid and the impact from being transferred to the lid by causing the circumferential-surface side member and the main body of the cask to receive the impact from the outer circumferential side, can prevent the positional deviation of the lid, and can keep the hermetically sealing performance by the metal gasket arranged between the lid and the main body. Advantageously, in the cask cushioning body, the impact absorber is formed by a combination of a plurality of wood blocks, and the impact absorber includes a first impact absorber group that is provided along a circumference of other end of the circumferential-surface side member, and that is made of a first material that absorbs the impact in a direction parallel to the end surface of the cask, a second impact absorber group that is provided around one end of the circumferential-surface side member, along the outer circumference of the end-surface side member, and adjacent to the first impact absorber group, and that is made of a second material that is lower in a compression strength than the first material and that absorbs the impact in the direction parallel to the end surface of the cask, a third impact absorber group that is provided in an external corner of the impact absorber along the outer circumference of the end-surface side member and adjacent to the second impact absorber group, and that is made of a third material that is lower in the compression strength than the second material and that absorbs the impact in a direction orthogonal to or inclined with respect to the end surface of the cask, a fourth impact absorber group that is provided along inner circumferences of the second and third impact absorber groups and adjacent to the second and third impact absorber groups, and that is made of the third material that absorbs the impact in the direction orthogonal to the end surface of the cask, and a fifth impact absorber group that is provided inside of a circumference of the fourth impact absorber group, and that is made of the third material that absorbs the impact in the direction parallel to the end surface of the cask. The cask cushioning body can appropriately absorb the impact of the drop or collision in assumable drop events of the cask by the impact absorber groups in addition to effects of the end-surface side member and circumferential-surface side member described above. Advantageously, in the cask cushioning body, a mounting hole into which a bolt for fixing the cushioning body to the cask is inserted is provided in the impact absorber, and the mounting hole is expandable and contractable in a depth direction of the mounting hole. According to the cask cushioning body, it is possible to suppress a sudden increase in an impact load caused by deformations of the mounting hole when the impact absorber starts deforming at the time of the vertical drop or vertical collision of the cask. As a result, it is possible to suppress an excessive force from acting on the bolt that fixes the lid to the main body at the time of the vertical drop or vertical collision of the cask, thereby keeping the hermetically sealing performance by the metal gasket. According to the present invention, the cask cushioning body that can exhibit an improved impact absorbing performance in various assumable drop events. Exemplary embodiments of the present invention will be explained below in detail with reference to the accompanying drawings. The present invention is not limited to the embodiments. In addition, constituent elements in the following embodiments include those that can be replaced and easily anticipated by persons skilled in the art, or that are substantially equivalent. FIG. 1 is a partially exploded perspective view of a cask to which a cushioning body according to an embodiment of the present invention is applied, FIG. 2 is a plan view of the cask shown in FIG. 1, and FIG. 3 is a bottom view of the cask shown in FIG. 1. As shown in FIGS. 1 to 3, a cask 100 storing therein radioactive materials is formed into a generally cylindrical shape. The cask 100 includes a cavity 101, a basket 102, a main body 103 serving as a cask main body, a bottom part 104, a lid 105, and a neutron shield 106. The cavity 101 is formed within the cask 100. The cavity 101 is a space surrounded by the main body 103, the bottom part 104, and the lid 105. The basket 102 is provided within the cask 100. The basket 102 includes cells 102a that accommodate spent fuel assemblies that are spent fuel. For example, the basket 102 is made of an aluminum compound material to which powder of a material having a neutron absorption performance is added. An interior of the cavity 101 is formed in conformity with an outer circumferential shape of the basket 102. Alternatively, the cavity 101 can be formed into a simple cylindrical shape, and spacers excellent in a heat transfer performance can be arranged in gaps between the cavity 101 and the basket 102 that are unavailable as the cells 102a. The main body 103 is formed into a generally cylindrical shape. The bottom part 104 is connected to a lower end (the other end) of the main body 103 by welding. The lid 105 is provided on an opening end of an upper portion (one end) of the main body 103. Trunnions 108 are fixed to an exterior of the main body 103. The trunnions 108 function as lifting tools for lifting the cask 100 by a lifting device such as a crane. The neutron shield 106 is filled in surroundings of the main body 103. The neutron shield 106 is a polymer material that contains hydrogen in large amounts and realizes a neutron shielding function. Boron or boron-compound-added resin is typically used as a neutron absorbing agent of the neutron shield 106. This can improve the neutron shielding function of the main body 103. Furthermore, the bottom part 104 is provided to protrude cylindrically from a bottom of the main body 103, and the neutron shield 106 is filled in an interior of the bottom part 104, thereby improving the neutron shielding function of the bottom part 104. A plurality of bolt holes 104c with which bolts 7 for attaching the cushioning body according to the present embodiment to the cask 100 are threaded are provided circumferentially at a bottom of the bottom part 104, as shown in FIG. 3. The lid 105 closes the opening end of the main body 103 that is opposite to the bottom part 104. The lid 105 is configured to include a primary lid 105a and a secondary lid 105b. The primary lid 105a is made of carbon steel or stainless steel that shields γ rays and formed into a disk shape. The secondary lid 105b appears on an exterior of the cask 100 while covering the primary lid 105a. Similarly to the primary lid 105a, the secondary lid 105b is made of carbon steel or stainless steel that shields γ rays and formed into a disk shape. Alternatively, the neutron shielding function of the primary lid 105a can be improved by filling the neutron shield 106 in the primary lid 105a. The primary lid 105a and the secondary lid 105b are attached to the main body 103 by bolts 109a and 109b made of carbon steel or stainless steel. Therefore, a plurality of insertion holes 105c into which the bolts 109a and 109b are inserted are formed circumferentially on the primary lid 105a and the secondary lid 105b. On the other hand, a plurality of bolt holes 110a and 110b into which the bolts 109a and 109b are threaded are formed circumferentially on the main body 103 so as to correspond to the number of the insertion holes 105c. Some of the insertion holes 105c (in the present embodiment, twelve) of the secondary lid 105b that appears on the exterior of the cask 100 are provided to insert the bolts 7 for attaching a cushioning body 1 to the cask 100, and some of the bolt holes 110b (in the present embodiment, twelve) are also provided to tighten the bolts 7. As shown in FIG. 2, a plurality (in the present embodiment, four) of air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d are provided on the lid 105 (the primary lid 105a and the secondary lid 105b). The air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d fill feed/drainage water or helium gas to or from the cavity 101 into the cavity 101 and check pressure of the cavity 101 while attaching the lid 105 to the main body 103. Each of the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d is configured to include a hole that penetrates the lid 105 (the primary lid 105a and the secondary lid 105b) and a cover that closes the hole by locating a plug or the like in the hole. Although not shown in FIG. 2, a metal gasket is provided between the primary lid 105a and the main body 103. Furthermore, although not shown in FIG. 2, metal gaskets are provided between the secondary lid 105b and the main body 103 and between the lid 105 (the primary lid 105a and the secondary lid 105b) and the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d, respectively. The metal gasket secures a hermetically sealing performance between the primary lid 105a and the main body 103. Further, the metal gasket secures the hermetically sealing performance between the secondary lid 105b and the main body 103. In addition, the metal gasket secures the hermetically sealing performance between the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d and the main body 103. FIG. 4 is a plan view of the cushioning body according to the present embodiment, FIG. 5 is a cross-sectional side view of the cushioning body according to the present embodiment, FIG. 6 is a perspective plan view of the cushioning body according to the present embodiment, and FIG. 7 is a perspective bottom view of the cushioning body according to the present embodiment. As shown in FIGS. 4 to 7, the cushioning body 1 applied to the cask 100 described above is attached to both ends of the cask 100 that stores therein the spent fuel assemblies, that is, an end near the lid 105 (lid-105 side end) and that near the bottom part 104 (bottom-part-104 side end), and absorbs impact applied to the cask 100. The cushioning body 1 is configured to include an end-surface side member 2, a circumferential-surface side member 3, and an impact absorber 4. In the cushioning body 1 according to the present embodiment, constituent elements attached to the lid-105 side end are identical to those attached to the bottom-part-104 side end. Therefore, only the constituent elements attached to the lid-105 side end are explained below and explanations of those attached to the bottom-part-104 side end will be omitted. The end-surface side member 2 is arranged along each of end surfaces 100a of the cask 100. The end surfaces 100a of the cask 100 are an upper surface of the lid 105 (the secondary lid 105b) and the bottom of the bottom part 104. The end-surface side member 2 includes a plurality (in the present embodiment, two) of plates 21 and 22 made of steel. Each of the plates 21 and 22 is formed into a circular shape equal to or slightly larger than a circular shape of the secondary lid 105b in a size. Furthermore, the plates 21 and 22 are provided to face each other at a distance kept therebetween, and plate surfaces of the plates 21 and 22 are arranged along each of the end surfaces 100a of the cask 100. In the present embodiment, the plate 21 is arranged to face the end surface 100a of the cask, and the plate 22 is arranged to be away from the end surface 100a of the cask 100 because of the distance from the plate 21. In a case of the end-surface side member 2, at least one plate (the plate 21 in the present embodiment; however, the plate 22 can serve as the plate) covers entirely the end surface 100a of the cask 100. A circular opening 23 is formed in a central portion of the plate other than the at-least-one plate (the plate 22 in the present embodiment; however, the plate 21 can serve as the other plate). A partition member 24 is provided on an edge of the opening 23. The partition member 24 is made of steel, and formed into a cylindrical shape along a circular shape of the opening 23. The partition member 24 is arranged between the plates 21 and 22, thereby partitioning a region between the plates 21 and 22 into a region that the plates 21 and 22 face and a region near the opening 23. The opening 23 is not necessarily formed in all the plates 21 and 22. The partition member 24 is not arranged if the opening 23 is not formed. Furthermore, a plurality of end-surface reinforcing members 5 connecting the plate 21 to the plate 22 are provided in the end-surface side member 2. The end-surface reinforcing members 5 are ribs made of steel and provided to extend radially around a center of circles of the plates 21 and 22. In the cushioning body 1 according to the present embodiment, because of the formation of the opening 23 in the central portion of the plate 22, the end-surface reinforcing members 5 include portions other than the openings 23 and are provided at which the plate surfaces of the plates 21 and 22 face each other between the partition member 24 and the circumferential-surface side member 3. In contrast, if the openings 23 are not formed in all plates 21 and 22, the end-surface reinforcing members 5 include central portions and are provided at which the plate surfaces of the plates 21 and 22 face each other. Further, when the opening 23 is provided, a plurality of central reinforcing members 6 are provided in regions opened by the opening 23 and shared by the other plate. The central reinforcing members 6 are ribs made of steel and provided to extend radially around the center of the circles of the plates 21 and 22. Furthermore, in the end-surface side member 2, locking holes 8 that lock the bolts 7 threaded with the bolt holes 110b of the cask 100 are provided in the plate 21 closest to the end surface 100a of the cask 100, and insertion holes 9 into which the bolts 7 are inserted are provided in the other plate 22. That is, the end-surface side member 2 is attached to the cask 100 by fixing the plate 21 to the cask 100 by the bolts 7. The circumferential-surface side member 3 is arranged along each of end-portion outer-circumferential surfaces 100b of the cask 100. The end-portion outer-circumferential surfaces 100b are an outer circumferential surface of the lid 105 (the secondary lid 105b) and that of the bottom part 104. The circumferential-surface side member 3 includes a cylindrical body 31 made of steel and having a cylindrical shape. One end of the cylindrical body 31 is connected to an edge of the end-surface side member 2 by welding. Specifically, one end of the cylindrical body 31 is connected to the plate surface of the plate 22 and connected to a periphery of the plate 21. With this configuration, the circumferential-surface side member 3 covers the end-portion outer-circumferential surface 100b of the cask 100 in a state of attaching the end-surface side member 2 to the end surface 100a of the cask 100. Note that steel thicker than the other constituent elements made of steel is used as a material of the cylindrical body 31. Because one end of the cylindrical body 31 is connected to the plate surface of the plate 22, the circumferential-surface side member 3 is provided with a protruding portion 10 in which the periphery of the plate 22 in the end-surface side member 2 protrudes outward of the cylindrical body 31 over an entire circumference of one end of the cylindrical body 31. Furthermore, the circumferential-surface side member 3 is provided with a flange portion 11 protruding outward over an entire circumference of the other end of the cylindrical body 31 and made of steel. Further, the circumferential-surface side member 3 is provided with a plurality of circumferential-surface reinforcing members 12 connecting the protruding portion 10 to the flange portion 11 and arranged on an outside surface of the cylindrical body 31. The circumferential-surface reinforcing members 12 are ribs made of steel and provided to extend on an outer circumferential surface of the cylindrical body 31 radially around the circles of the plates 21 and 22. The impact absorber 4 is attached to an exterior of the end-surface side member 2 described above and an exterior of the circumferential-surface side member 3 described above, and arranged on an outermost side of the entire cushioning body 1. The impact absorber 4 absorbs the impact generated when the cask 100 drops or collides by deforming, and is made of wood, for example. The impact absorber 4 is attached to the end-surface side member 2 along an outside plate surface of the outermost plate 22. When the opening 23 is formed in the outermost plate 22, the impact absorber 4 is attached to the end-surface side member 2 while being inserted into a region of the end-surface side member 2 that is within the opening 23 and that is formed by the other plate 21, the partition member 24, and the central reinforcing members 6. The impact absorber 4 is attached to the circumferential-surface side member 3 while a part of the impact absorber 4 is inserted into a region formed by the outer circumferential surface of the cylindrical body 31, the protruding portion 10, the flange portion 11, and the circumferential-surface reinforcing members 12. Furthermore, the impact absorber 4 is attached not only to the end-surface side member 2 and the circumferential-surface side member 3 but also to angular portions outside of external corners formed by the end-surface side member 2 and the circumferential-surface side member 3. Therefore, the impact absorber 4 is attached to outside of the end-surface side member 2 and that of the circumferential-surface side member 3, and arranged on the outermost side of the cushioning body 1. The cushioning body 1 protrudes outward of both end surfaces of the cask 100 and protrudes outward of an outside diameter of the neutron shield 106 that is provided around the main body 103 of the cask 100. The cask cushioning body according to the present embodiment configured as described above includes the end-surface side member 2 in which a plurality of plates 21 and 22 made of steel are formed at a distance between the plate surfaces of the plates 21 and 22 that face each other, and in which these plate surfaces of the plates 21 and 22 are arranged along the end surface 100a of the cask 100, and the circumferential-surface side member 3 that forms the cylindrical body 31 made of the steel, one end of which is connected to the periphery of the end-surface side member 2, and that is arranged along the end-portion outer-circumferential surface 100b of the cask 100, and the impact absorber 4 that absorbs the impact by deforming is provided outside of the end-surface side member 2 and the circumferential-surface side member 3. A drop event with a height of drop of 9 meters (drop test I) and a drop event with a height of drop of 1 meter on an upright round bar having a diameter of 15 centimeters (drop test II) are imposed, as drop tests for demonstrating capabilities (of keeping the hermetic sealing, shielding, and subcritical performances) to withstand accident conditions of transport, on the cask 100 in Regulations for the Safe Transport of Radioactive Material, 2005 Edition, the IAEA (International Atomic Energy Agency) Safety Standard. For example, the drop events include (1) a vertical drop with the end surface 100a (a central axis R of the cask 100 shown in FIG. 1) oriented in a vertical direction, (2) a horizontal drop with the end surface 100a (the central axis R) oriented in a horizontal direction, and (3) a corner drop with the end surface 100a (the central axis R) oriented aslant. In recent years, the metal gasket is adopted for the use between the lid 105 and the main body 103 of the cask 100, it is important to suppress a positional deviation of the lid 105 from the main body 103, and the weight of the cask 100 increases as a result of an increase in the number of stored spent fuel assemblies. Therefore, the demand for a cushioning performance of the cushioning body 1 rises. The cask cushioning body according to the present embodiment can exhibit the improved impact absorbing performance in the various drop events such as the assumable drop events (1) to (3) described above to which the cask 100 is subjected. Specifically, at the time of a drop or an impact, the end-surface side member 2 made of steel does not have a large deformation while the outer impact absorber 4 deforms and absorbs the impact. Therefore, a large deformation does not cause the end-surface side member 2 to abut on the lid 105 and an impact load that possibly prevents the hermetically sealing performance of the cask 100 from being kept is not transferred to the cask 100. Therefore, it is possible to reduce a local load on the end surface 100a of the cask 100 in the drop events (1) and (3). Furthermore, at the time of the drop or the impact, the circumferential-surface side member 3 does not have a large deformation while the outer impact absorber 4 deforms and absorbs the impact. Accordingly, a large deformation does not cause the circumferential-surface side member 3 to abut on the end-portion outer-circumferential surface 100b of the cask 100 and the impact load that possibly prevents the hermetically sealing performance of the cask 100 from being kept is not transferred to the cask 100. Therefore, it is possible to reduce a local load on the end-portion outer-circumferential surface 100b of the cask 100 in the drop event (2). Furthermore, on the lid 105 side, it is possible to prevent a positional deviation of the lid 105, and to keep the hermetically sealing performance by the metal gasket arranged between the lid 105 and the main body 103. As a result, it is possible to improve the impact absorbing performance in the assumable various events. The cylindrical body 31 of the circumferential-surface side member 3 is made of the thicker steel than the other constituent elements made of the steel. This makes it possible to attain appropriate rigidity while preventing the end-portion outer-circumferential surface 100b of the cask 100 on which a strict size restriction is set for the cushioning body 1 from largely protruding outward. Furthermore, in the cask cushioning body according to the present embodiment, the end-surface side member 2 includes a plurality of end-surface reinforcing members 5 provided to connect the plates 21 and 22 to each other. The cask cushioning body can realize a reduction in the load on the end surface 100a of the cask 100 because the end-surface reinforcing members 5 make it more difficult to deform the end-surface side member 2. Further, in the cask cushioning body according to the present embodiment, the opening 23 is formed in the central portion of the plate 22 other than at least one plate of the end-surface side member 2, and the end-surface side member 2 includes a plurality of central reinforcing members 6 in the region opened by the opening 23 between the plate 21 and the other plate 21. The cask cushioning body can realize a reduction in the local load on the central portion of the end surface 100a of the cask 100 by the central reinforcing members 6 when a bar member penetrates the impact absorber 4. Furthermore, in the cask cushioning body according to the present embodiment, the impact absorber 4 is inserted into the region surrounded by the central reinforcing members 6 and the plate 21 provided with the central reinforcing members 6. The cask cushioning body can ensure attaining an impact absorbing effect of the impact absorber 4 because a positional deviation of the impact absorber 4 is prevented when the impact absorber 4 absorbs the impact by allowing portions in which the central reinforcing members 6 are provided to hold the impact absorber 4. It is particularly possible to prevent the positional deviation of the impact absorber 4 and to ensure attaining the impact absorbing effect of the impact absorber 4 at the time of (3) the corner drop with the end surface 100a oriented aslant in the drop event with the height of the drop of 9 meters. Further, in the cask cushioning body according to the present embodiment, the circumferential-surface side member 3 includes the protruding portion 10 in which the periphery of the plate 22 in the end-surface side member 2 protrudes outward of the cylindrical body 31 over the entire circumference of one end of the cylindrical body 31, the flange portion 11 protruding outward over the entire circumference of the other end of the cylindrical body 31, and a plurality of circumferential-surface reinforcing members 12 connecting the protruding portion 10 to the flange portion 11 and arranged on the outside surface of the cylindrical body 31. The cask cushioning body can realize a further reduction in the local load on the end-portion outer-circumferential surface 100b of the cask 100 because the circumferential-surface reinforcing members 12 make it more difficult to deform the circumferential-surface side member 3. Furthermore, in the cask cushioning body according to the present embodiment, the impact absorber 4 is inserted into the region surrounded by the protruding portion 10, the flange portion 11, and the circumferential-surface reinforcing members 12. The cask cushioning body can ensure attaining the impact absorbing effect of the impact absorber 4 because the positional deviation of the impact absorber 4 is prevented when the impact absorber 4 absorbs the impact by allowing the protruding portion 10, the flange portion 11, and the circumferential-surface reinforcing members 12 to hold the impact absorber 4. It is particularly possible to prevent the positional deviation of the impact absorber 4 and to ensure attaining the impact absorbing effect of the impact absorber 4 at the time of (3) the corner drop with the end surface 100a oriented aslant in the drop events with the height of the drop of 9 meters. Further, in the cask cushioning body according to the present embodiment, regions of the end-surface side member 2 in which the plurality of the plates 21 and 22 face each other cover the bolts 109b for fixing the lid 105 (particularly, the secondary lid 105b) of the cask 100. The cask cushioning body can ensure maintaining a fastening force of the lid 105 by reducing the concentration of a load on the bolts 109b for fixing the lid 105 (particularly, the secondary lid 105b) of the cask 100 and by preventing damage of the bolts 109b, and can prevent the positional deviation of the lid 105 when the bar member penetrates the impact absorber 4. As a result, it is possible to keep the hermetically sealing performance by the metal gasket arranged between the lid 105 and the main body 103. It is particularly possible to reduce an excessive load on the bolts 109b at the time of (1) the vertical drop with the end surface 100a oriented in the vertical direction in the drop event with the height of drop of 1 meter on the round boar. Although not shown in the drawings, it is preferable that recesses are formed in the plate surface of the plate 21 that covers the bolts 109b which surface faces the bolts 109b. Alternatively, it is preferable that the bolts 109b are provided to be recessed with respect to the end surface 100a. These recesses can reduce the impact related to the bolts 109b (particularly the impact at the time of (1) the vertical drop with the end surface 100a oriented in the vertical direction in the drop event with the height of drop of 1 meter on the round bar). In the cask cushioning body according to the present embodiment, the regions of the end-surface side member 2 in which the plurality of the plates 21 and 22 face each other cover the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d provided on the lid 105. The cask cushioning body can keep the hermetic sealing performance by cover of the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d by reducing the concentration of a load on the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d provided on the lid 105 of the cask 100, when the bar member penetrates the impact absorber 4. It is particularly possible to reduce an excessive load on the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d at the time of (1) the vertical drop with the end surface 100a oriented in the vertical direction in the drop event with the height of drop of 1 meter on the round boar. Although not shown in the drawings, it is preferable that recesses are formed in the plate surface of the plate 21 that covers the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d which surface faces the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d. Alternatively, it is preferable that the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d are provided to be recessed with respect to the end surface 100a. These recesses can reduce the impact related to the air-supply/exhaust, water-feed/drainage or pressure monitoring units 105d (particularly the impact at the time of (1) the vertical drop with the end surface 100a oriented in the vertical direction in the drop event with the height of drop of 1 meter on the round bar). In the cask cushioning body according to the present embodiment, the impact absorbing material that absorbs the impact by deforming is filled in the end-surface side member 2 between the regions in which the plurality of the plates 21 and 22 face each other. In this case, the impact absorbing material refers to wood, urethane foam, foam metal, or the like. The cask cushioning body can absorb the impact (particularly, a high frequency impact) applied to the plate 22 when the bar penetrates the impact absorber 4 at the time of (1) the vertical drop with the end surface 100a oriented in the vertical direction in the drop event with the height of drop of 1 meter on the round bar. As shown in FIGS. 8 and 9 that are enlarged cross-sectional side views of the cushioning body according to the present embodiment, the cask cushioning body according to the present embodiment is formed so that a size A between the outer circumferential surface of the lid 105 (the secondary lid 105b in the present embodiment) constituting the cask 100 and an inner circumferential surface of the cylindrical body 31 of the circumferential-surface side member 3 is larger than a size B between an outer circumferential surface of the main body 103 of the cask 100 to which the lid 105 is fixed and the inner circumferential surface of the cylindrical body 31 of the circumferential-surface side member 3. Specifically, as shown in FIG. 8, a stepped portion 31a recessed by about 1 millimeter is provided in a portion of the cylindrical body 31 of the circumferential-surface side member 3 that faces the outer circumferential surface of the secondary lid 105b, and the stepped portion 31a is made thinner than portions of the cylindrical body 31 that face the outer circumferential surface of the main body 103. Alternatively, as shown in FIG. 9, an outside diameter of the secondary lid 105b is made smaller than that of the main body 103 by about 1 millimeter. The cask cushioning body can reduce the impact applied to the lid 105 (the secondary lid 105b) and can further prevent the circumferential-surface side member 3 from abutting on the lid 105 (the secondary lid 105b) and the impact from being transferred to the lid 105 by causing the circumferential-surface side member 3 and the main body 103 of the cask 100 to receive the impact from the outer circumferential side (particularly the impact at the time of (2) the horizontal drop with the end surface 100a oriented in the horizontal direction). As a result, it is possible to prevent the positional deviation of the lid 105 and to keep the hermetically sealing performance by the metal gasket arranged between the lid 105 and the main body 103. The cask cushioning body according to the present embodiment can be configured to provide multiple (in the present embodiment, double) cylindrical bodies 31, as shown in FIG. 10 that is a cross-sectional side view of the cushioning body according to the present embodiment. In this case, one end of each cylindrical body 31 is connected to the plate surface of the plate 21 and the protruding portion 10 is provided outward of one end of each cylindrical body 31. The flange portion 11 is provided on the other end of each cylindrical body 31. The circumferential-surface reinforcing members 12 are provided in regions formed by the respective cylindrical bodies 31, the plate 22, and the flange portions 11. The circumferential-surface reinforcing members 12 are also provided in the regions formed by the outermost cylindrical body 31, the protruding portion 10, and the flange portion 11. The cask cushioning body can realize a further reduction in the load on the end-portion outer-circumferential surface 100b of the cask 100 because it is made more difficult to deform the circumferential-surface side member 3. When the multiple cylindrical bodies 31 are provided, each cylindrical body 31 is made of the steel as thick as the other constituent elements made of the steel. This makes it possible to attain appropriate rigidity while preventing the end-portion outer-circumferential surface 100b of the cask 100 on which a strict size restriction is set for the cushioning body 1 from largely protruding outward. When the multiple cylindrical bodies 31 are provided, the impact absorber 4 is inserted into the region surrounded by the outermost cylindrical body 31, the protruding portion 10, the flange portion 11, and the circumferential-surface reinforcing members 12. Therefore, the cask cushioning body according to the present embodiment can ensure attaining the impact absorbing effect of the impact absorber 4 by causing the protruding portion 10, the flange portion 11, and the circumferential-surface reinforcing members 12 to hold the impact absorber 4 and by preventing the positional deviation of the impact absorber 4 when the impact absorber 4 absorbs the impact. When the multiple cylindrical bodies 31 are provided, the impact absorbing material that absorbs the impact by deforming is filled in the region surrounded by each cylindrical body 31, the plate 22, the flange portion 11, and the circumferential-surface reinforcing members 12. In this case, the impact absorbing material refers to wood, urethane foam, foam metal, or the like. Therefore, the cask cushioning body can absorb the impact (particularly, a high frequency impact) applied to the outermost cylindrical body 31 when the bar penetrates the impact absorber 4 at the time of (2) the horizontal drop with the end surface 100a oriented in the horizontal direction in the drop event with the height of drop of 1 meter on the round bar. The impact absorber 4 in the cushioning body 1 according to the present embodiment is described. FIG. 11 is a cross-sectional side view of an impact absorber of the cushioning body according to the present embodiment, and FIG. 12 is a cross-sectional plan view of the impact absorber. In FIGS. 11 and 12, arrows indicate directions of fiber of the wood constituting the impact absorber 4. The cushioning body 1 according to the present embodiment can further fulfill a function required as the cushioning body of the cask 100 by forming the impact absorber 4 out of the wood and arranging the impact absorber 4 while changing types of the impact absorber 4 and directions of the fiber. As shown in FIG. 11, the impact absorber 4 is configured to include a combination of a first impact absorber group 41, a second impact absorber group 42, a third impact absorber group 43, a fourth impact absorber group 44, and a fifth impact absorber group 45. These first to fifth impact absorber groups 41 to 45 are formed by combining a plurality of wood blocks. An outermost side of the impact absorber 4 is covered with an outer shell 46. The first impact absorber group 41 is provided along a circumference of the other end of the circumferential-surface side member 3. The first impact absorber group 41 is made of a first material that absorbs the impact in a direction parallel to the end surface 100a of the cask 100 by arranging the first impact absorber group 41 to set the direction of the fiber along the direction parallel to the end surface 100a of the cask 100. As shown in FIGS. 11 and 12, in the first impact absorber group 41, a plurality of blocks divided in a circumferential direction of the impact absorber 4 are inserted into the region surrounded by the flange portion 11 and the circumferential-surface reinforcing members 12 and held by the flange portion 11 and the circumferential-surface reinforcing members 12. The second impact absorber group 42 is provided around one end of the circumferential-surface side member 3, along an outer circumference of the end-surface side member 2, and adjacent to the first impact absorber group 41. The second impact absorber group 42 is made of a second material having a lower compression strength than the first material and that absorbs the impact in the direction parallel to the end surface 100a of the cask 100 by arranging the second impact absorber group 42 to set the direction of the fiber along the direction parallel to the end surface 100a of the cask 100. As shown in FIGS. 11 and 12, in the second impact absorber group 42, a plurality of blocks divided in the circumferential direction of the impact absorber 4 are inserted into the region surrounded by the protruding portion 10 and the circumferential-surface reinforcing members 12 and held by the protruding portion 10 and the circumferential-surface reinforcing members 12. The third impact absorber group 43 is provided in each external corner of the impact absorber 4 along the outer circumference of the end-surface side member 2 and adjacent to the second impact absorber group 42. The third impact absorber group 43 is made of a third material that is lower in the compression strength than the second material and that absorbs the impact in a direction orthogonal to or inclined with respect to the end surface 100a of the cask 100 by arranging the third impact absorber group 43 to set the direction of the fiber along the direction parallel to the end surface 100a of the cask 100. Although not shown in the drawings, the third impact absorber group 43 is constituted by a plurality of blocks divided in the circumferential direction of the impact absorber 4. The fourth impact absorber group 44 is provided along inner circumferences of the second and third impact absorber groups 42 and 43 and adjacent to the second and third impact absorber groups 42 and 43. The fourth impact absorber group 44 is made of the third material that absorbs the impact in the direction orthogonal to the end surface 100a of the cask 100 by arranging the fourth impact absorber group 44 to set the direction of the fiber along the direction orthogonal to the end surface 100a of the cask 100. Although not shown in the drawings, the fourth impact absorber group 44 is constituted by a plurality of blocks divided in the circumferential direction of the impact absorber 4. The fifth impact absorber group 45 is provided inside of a circumference of the fourth impact absorber groups 44. The fifth impact absorber group 45 is made of the third material that absorbs the impact in the direction parallel to the end surface 100a of the cask 100 by arranging the fifth impact absorber group 45 to set the direction of the fiber along the direction parallel to the end surface 100a of the cask 100. Although not shown in the drawings, the fifth impact absorber group 45 is constituted by a plurality of blocks divided in the circumferential direction of the impact absorber 4. In the cask cushioning body according to the present embodiment, the opening 23 is formed in the central portion of the plate 22 other than at least one plate of the end-surface side member 2, and the end-surface side member 2 includes the plurality of the central reinforcing members 6 in the region opened by the opening 23 between the plate 21 and the other plate 21. The fifth impact absorber group 45 is inserted into the region surrounded by the central reinforcing members 6 and the plate 21 provided with the central reinforcing members 6. The first material that constitutes the first impact absorber group 41 has the highest compression strength among all the materials of the impact absorber 4 and oak, for example, is used as the first material. The second material that constitutes the second impact absorber group 42 is lower in the compression strength than the first material, and red cedar, for example, is used as the second material. The third material that constitutes the third impact absorber group 43, the fourth impact absorber group 44, and the fifth impact absorber group 45 is lower in the compression strength than the second material, and balsa, for example, is used as the third material. The compression strength refers to Young's modulus or a compressive strength when the impact absorber is compressed. The outer shell 46 is made of thinner steel than the end-surface side member 2 and the circumferential-surface side member 3, and provided along outer side surfaces of the first to fifth impact absorber groups 41 to 45 so as to cover the first to fifth impact absorber groups 41 to 45. The outer shell 46 is connected to the flange portion 11 of the circumferential-surface side member 3 by welding. The outer shell 46 protects the first to fifth impact absorber groups 41 to 45 from moist and drops of water, and absorbs the impact by deforming together with the first to fifth impact absorber groups 41 to 45 in the drop events described above. As described above, in the cask cushioning body according to the present embodiment, the impact absorber 4 is formed by a combination of a plurality of wood blocks, and includes the first impact absorber group 41 that is provided to surround the other end of the circumferential-surface side member 3 and that is made of the first material that absorbs the impact in the direction parallel to the end surface 100a of the cask 100, the second impact absorber group 42 that is provided to surround one end of the circumferential-surface side member 3, along the outer circumference of the end-surface side member 2, and adjacent to the first impact absorber group 41 and that is made of the second material that is lower in the compression strength than the first material and that absorbs the impact in the direction parallel to the end surface 100a of the cask 100, the third impact absorber group 43 that is provided in each external corner of the impact absorber 4 along the outer circumference of the end-surface side member 2 and adjacent to the second impact absorber group 42 and that is made of a third material that is lower in the compression strength than the second material and that absorbs the impact in the direction orthogonal to or inclined with respect to the end surface 100a of the cask 100, the fourth impact absorber group 44 that is provided along the inner circumferences of the second and third impact absorber groups 42 and 43 and adjacent to the second and third impact absorber groups 42 and 43 and that is made of the third material that absorbs the impact in the direction orthogonal to the end surface 100a of the cask 100, and the fifth impact absorber group 45 that is provided inside of the circumference of the fourth impact absorber group 44 and that is made of the third material that absorbs the impact in the direction parallel to the end surface 100a of the cask 100. The cask cushioning body can appropriately absorb the impact of the drop or collision in the assumable drop events of the cask 100 by the impact absorber groups 41, 42, 43, 44, and 45 in addition to effects of the end-surface side member 2 and the circumferential-surface side member 3 described above. FIG. 13 is a cross-sectional side view of the impact absorber of the cushioning body according to the present embodiment in another direction. As shown in FIG. 13, mounting holes 13 into which the bolts 7 that fix the cushioning body 1 to the cask 100 are inserted, respectively are provided in the impact absorber 4. These mounting holes 13 are expandable and contractable in a depth direction of the mounting holes 13 that is the direction orthogonal to the end surface 100a of the cask 100. The mounting holes 13 are provided coaxially with the locking holes 8 and the insertion holes 9. Bellows portions 14 are provided in the mounting holes 13. The bellows portions 14 cause the mounting holes 13 to deform in the direction orthogonal to the end surface 100a of the cask 100 with little resistance, when the cask 100 vertically drops or vertically collides. As described above, in the cask cushioning body according to the present embodiment, the mounting holes 13 into which the bolts 7 that fix the cushioning body 1 to the cask 100 are inserted are provided in the impact absorber 4, and the mounting holes 13 are expandable and contractable in the depth direction of the mounting holes 13. According to the cask cushioning body, it is possible to suppress a sudden increase in an impact load caused by deformations of the mounting holes 13 when the impact absorber 4 starts deforming at the time of the vertical drop or vertical collision of the cask 100. As a result, it is possible to suppress an excessive force from acting on the bolts 7 that fix the lid 105 to the main body 103 at the time of the vertical drop or vertical collision of the cask 100, thereby keeping the hermetically sealing performance by the metal gaskets. 1 cushioning body 2 end-surface side member 21, 22 plate 23 opening 24 partition member 3 circumferential-surface side member 31 cylindrical body 31a stepped portion 4 impact absorber 5 end-surface reinforcing member 6 central reinforcing member 7 bolt 8 locking hole 9 insertion hole 10 protruding portion 11 flange portion 12 circumferential-surface reinforcing member 13 mounting hole 14 bellows portion 41 first impact absorber group 42 second impact absorber group 43 third impact absorber group 44 fourth impact absorber group 45 fifth impact absorber group 46 outer shell 100 cask 100a end surface 100b end-portion outer-circumferential surface 103 main body 104 bottom part 105 lid 105a primary lid 105b secondary lid 105c insertion hole 105d air-supply/exhaust, water-feed/drainage or pressure monitoring unit 109a, 109b bolt 110a, 110b bolt hole |
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052971821 | summary | FIELD OF THE INVENTION This invention relates generally as indicated to a method of decommissioning a nuclear reactor and more particularly to a decommissioning process which includes the steps of encapsulating the relevant portions of the reactor into a solid integral reactor capsule and converting this reactor capsule into decommissioned transportable segments. BACKGROUND OF THE INVENTION A nuclear power cycle may be viewed as originating with a nuclear reaction which occurs within a component commonly called a "nuclear reactor." A nuclear reactor typically comprises a vessel which defines a chamber and a fuel core which is situated within this chamber. In commercial reactors for utility applications, the fuel core consists of a number of fuel rods which are arranged so that a self-sustaining nuclear reaction will take place. This nuclear reaction releases an enormous amount of energy which is used to transfer heat to a circulating fluid. In one type of nuclear reactor, a "boiling water reactor" (BWR), the circulating fluid is water which is converted to steam within the reactor vessel and then supplied to the turbine of the power cycle. In another type of nuclear reactor, a "pressurized water reactor" (PWR), the circulating fluid is heated within the reactor vessel and then supplied to a closed heat exchanger, or steam generator, wherein the circulating fluid converts water to steam which is then supplied to the turbine of the power system. In either case, the reactor vessel will include input/output nozzles for the circulating fluid which are connected to the appropriate plant lines. Additionally, both BWR units and PWR units will include reactor internals which interact with the circulating fluid and/or which are used to control the fuel core. As was indicated above, an enormous amount of energy is released during a nuclear reaction. In fact the amount of energy released per atom exceeds by a factor of several million the amount of energy obtainable per atom in a chemical reaction, such as the burning of fossil fuels. Consequently, nuclear power has become a very attractive alternative for the utility industry and more than 500 nuclear reactors are now either on-line or under construction world-wide. However, the nuclear reaction also emits potentially harmful radiation thereby creating sometimes complex construction and operating challenges. Additionally, and of particular interest in the present application, this radiation produces certain obstacles when dismantling or decommissioning a nuclear reactor at the end of its operating life. More specifically, the inside of the vessel, along with the reactor internals, will be considered radioactively contaminated thereby placing certain restrictions on the decommissioning of the reactor. It should be noted at this point that removal and disposal of a nuclear reactor's fuel core is not normally considered part of the decommissioning process. This is due to the fact that it is an accepted industry practice to replace one-third of the fuel rods of a nuclear reactor each operating year. Consequently, the removal/disposal of the fuel core during a decommissioning process will usually not present any problems above and beyond those encountered during annual replacements. However, the decommissioning of the remaining portions the nuclear reactor, i.e., the vessel and the reactor internals, presents challenges not normally confronted during the maintenance of an on-line nuclear reactor. In the past, the decommissioning of nuclear reactors has been accomplished almost exclusively by a "water platform" method. In such a method, the top of the reactor vessel is removed and the vessel is filled with water which thereby functions as a radioactive shield relative to the reactor internals. A platform is placed on top of the water and underwater cutting is performed on the pieces of the reactor internals located immediately beneath the platform. These pieces are then loaded onto the platform and transferred to either a "wet cutting station" in which further underwater cutting is performed or a "dry cutting station" in which further cutting is performed in an air-controlled environment. The resulting pieces of the reactor internals are then placed in "casks" which comprise a casing made of a radioactively shielding material. The water level is then decreased, the platform lowered, and the cutting process is again initiated. This sequence of events is repeated until all of the reactor internals have been removed. Thereafter the cutting of the reactor vessel itself is initiated. This "water platform" method of decommissioning a nuclear reactor, while acceptably effective, places many time, cost and safety constraints on a decommissioning project. For example, although water functions as a radioactive shield, some worker interaction with the sectioned reactor internals will usually be experienced in the transfer between the water platform and the wet/dry cutting stations. Additionally, the cutting process usually produces a significant amount of particles whereby respirators and HEPA ventilation are sometimes necessary to combat the effects of airborne contamination. Furthermore, the "shielding" water in the vessel will absorb particles produced during the cutting process whereby constant circulation and filtration of this fluid is necessary to remove liquid radioactive waste. An alternate method of decommissioning a nuclear reactor was recently used on a retired nuclear reactor at the Shippingport Power Plant. The Shippingport nuclear reactor was an offspring of the Eisenhower presidency and thus its decommissioning was orchestrated by the United States Department of Energy. In decommissioning this unit, the reactor vessel was filled with concrete and then moved in one piece to a disposal site. The Shippingport reactor, which was rated at 72 megawatts, was significantly smaller than most of the commercial nuclear reactors in use or construction today. Nonetheless, the weight of the reactor when filled with concrete required the fabrication of special lifting equipment to lift the reactor from its underground housing. More particularly, the project required the erection of a gigantic frame, the construction of four huge hydraulic jacks, each having an approximately 6,000 ton lifting capacity, and the mounting of these jacks on the frame. In the transfer of the Shippingport reactor to the transport vehicle (a barge in this case), the jacks hoisted the reactor seventy-seven feet into the air, moved it approximately thirty-eight feet horizontally along a track and then lowered it onto a trailer. The Energy Department's decision to decommission the Shippingport reactor in this manner, which avoided cutting apart the radioactive structure, saved an estimated seven million dollars and, perhaps more importantly, dramatically reduced worker exposure to radiation. However, such a procedure is probably not possible for most commercial reactors which possess an average rating of approximately 1000 megawatts, and would weigh over 2500 tons if filled with concrete. Moreover, even if larger reactors could be moved in one piece, the capital cost of fabricating the necessary lifting equipment would probably make such an approach economically unfeasible. Applicant therefore believes that a need remains for a cost effective method of decommissioning a nuclear reactor in which radiation exposure is minimized. SUMMARY OF THE INVENTION The present invention provides a cost effective process of decommissioning a nuclear reactor in which radiation exposure is minimized. More particularly, the present invention provides a method of decommissioning a nuclear reactor which comprises the steps of encapsulating portions of the reactor vessel and reactor internals into a solid reactor capsule and then converting this reactor capsule into a plurality of decommissioned segments. The conversion of the reactor capsule is preferably accomplished by cutting the reactor capsule into transportable-size segments and then encasing these transportable-size segments. The cutting procedure may be performed in two stages, namely cutting the reactor capsule into a series of sections and then subsequently cutting each of these sections into a plurality of transportable-size segments. The encapsulating step preferably includes the step of forming a matrix within the chamber which integrally attaches to the vessel and integrally embeds the reactor internals to create a solid reactor capsule. Such a reactor capsule has an outer shell which is formed from the vessel and which substantially encases the matrix and thus the reactor internals. More particularly, this encapsulating is preferably accomplished by providing a fluidized matrix-creating material which may be predictably solidified and which functions as a radioactive shield in its solid state. The fluidized matrix-creating material is introduced into the reactor chamber and then solidified in such a manner that it integrally attaches to the vessel and integrally embeds the reactor internals to form the reactor capsule. Concrete is the preferred matrix-creating material because it performs well as a matrix, it is compatible with the subsequent cutting steps, and, perhaps most importantly, it functions quite effectively as a radioactive shield. When concrete is used as the matrix-creating material, the "solidifying" step entails simply waiting for the concrete to cure. The cutting of the reactor capsule preferably includes the steps of cutting a first section of the capsule and then transferring it to a location away from the direct locality of the remaining portion of the capsule. Thereafter, a second section is then cut from the remaining portion of the capsule and then this section is transferred to a location away from the direct locality of the now remaining portion of the capsule. This sequence is repeated until all of the relevant portions of the reactor capsule have been transferred. The size of the reactor sections is chosen so that pre-existing in-house equipment, such as the reactor building crane, may be utilized during the transferring step. The series of sections formed from the reactor capsule are preferably transferred to an appropriate secondary cutting station within the reactor building and this appropriate secondary cutting station will in most instances be either a dry cutting station or a wet cutting station. The sectioned reactor capsule is then further cut, if necessary, into transportable-size segments which each include a portion of the outer shell and a portion of the matrix. Sheets, which are dimensioned to cover the exposed surfaces of the matrix, are then attached to the transportable segment to cover these exposed surfaces. In this manner, the reactor internals will be embedded within a radioactively shielding matrix and this matrix will be further encased by the portions of the vessel and the encasing sheets. As such, the need for shipping casks will be eliminated in many situations. Thus decommissioning a nuclear reactor according to the present invention provides several advantages over the prior art processes. For example, the cutting of the vessel and the reactor internals occurs substantially simultaneously whereby the preferred method provides for removal of both the reactor vessel and the reactor internals for essentially the same cost as individually removing either of these components with the water-platform method. Additionally, during the cutting steps and all subsequent steps, the reactor internals are embedded in a material which functions as a radioactive shield whereby worker interaction and environmental exposure to contaminated components is minimized. Furthermore, in contrast to the Shippingport decommissioning method, "in-house" lifting equipment may be used instead of specially fabricated units and this will usually substantially reduce the overall cost of the decommissioning project. Still further, the encasing step of the method will eliminate the need for shipping casks in many situations. These and other features of the invention are fully described and particularly pointed out in the claims. The following descriptive annexed drawings set forth in detail one illustrative embodiment, this embodiment being indicative, however, of but one of the various ways in which the principles of the invention may be employed. |
abstract | An energy-sensitive computed tomography system is provided. The energy-sensitive computed tomography system includes an X-ray source configured to emit an X-ray beam resulting from electrons impinging upon a target material. The energy-sensitive computed tomography system also includes an object positioned within the X-ray beam. The energy-sensitive computed tomography system further includes a detector configured to receive a transmitted beam of the X-rays through the object. The energy-sensitive computed tomography system also includes a filter having an alternating pattern disposed between the X-ray source and the detector, the filter configured to facilitate measuring projection data that can be used to generate low-energy and high-energy spectral information. |
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abstract | An apparatus for generating medical isotopes provides an annular fissile solution vessel surrounding a neutron generator. The annular fissile solution vessel provides for good capture of the emitted neutrons and a geometry that provides enhanced stability in an aqueous reactor. A neutron multiplier and/or a neutron moderator may be used to improve the efficiency and control the criticality of the reaction in the annular fissile solution vessel. |
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claims | 1. An apparatus for forming a solid immersion lens (SIL) in an optical medium, comprising:a milling head configured to generate a focused ion beam; anda binary bitmap milling pattern defining locations at which the focused ion beam is projected onto a surface of the optical medium, wherein the locations at which the focused ion beam impact the surface of the optical medium are located completely within the binary bitmap milling pattern and are randomized over successive raster scans of the surface of the optical medium, wherein the binary bitmap milling pattern comprises a plurality of pixels, a first portion of the pixels defining locations where the focused ion beam impacts the surface of the optical medium and a second portion of the pixels defining locations where the focused ion beam is prevented from impacting the surface of the optical medium, each successive raster scan comprising a subsequent binary bitmap milling pattern having a different distribution of pixels that impact the surface of the optical medium and that are prevented from impacting the surface of the optical medium than a previous binary bitmap milling pattern, each different distribution of pixels comprising a functionally equivalent binary bitmap milling pattern. 2. The apparatus of claim 1, wherein the successive raster scans comprise electronically rotating rastering axes of the focused ion beam relative to the surface of the optical medium for each successive raster scan. 3. The apparatus of claim 1, wherein the successive raster scans comprise rotating the surface of the optical medium relative to the rastering axes of the focused ion beam for each successive raster scan. 4. The apparatus of claim 1, wherein the subsequent binary bitmap milling pattern randomizes a dwell time of the focused ion beam. 5. The apparatus of claim 4, wherein the focused ion beam covers an area of 3 pixels by 3 pixels and 14 binary bitmap milling patterns create the SIL. 6. The apparatus of claim 5, wherein the SIL is created on a backside of the optical medium. 7. A method for forming a solid immersion lens (SIL), comprising:generating a focused ion beam;projecting the focused ion beam onto an optical medium at locations defined by a binary bitmap milling pattern, wherein the locations at which the focused ion beam impact a surface of the optical medium are located completely within the binary bitmap milling pattern and are randomized over successive raster scans of the surface of the optical medium to form at least a portion of a hemispherical structure in the optical medium; andusing the binary bitmap milling pattern to define first locations where the focused ion beam impacts a surface of the optical medium and to define second locations where the focused ion beam is prevented from impacting the surface of the optical medium, each successive raster scan comprising a subsequent binary bitmap milling pattern having a different distribution of pixels that impact the surface of the optical medium and that are prevented from impacting the surface of the optical medium than a previous binary bitmap milling pattern, each different distribution of pixels comprising a functionally equivalent binary bitmap milling pattern. 8. The method of claim 7, further comprising electronically rotating the rastering axes of the focused ion beam relative to the surface of the optical medium for each successive raster scan. 9. The method of claim 7, further comprising rotating the surface of the optical medium relative to the rastering axes of the focused ion beam for each successive raster scan. 10. The method of claim 7, wherein the subsequent binary bitmap milling pattern randomizes a dwell time of the focused ion beam. 11. The method of claim 10, wherein the focused ion beam covers an area of 3 pixels by 3 pixels and 14 binary bitmap milling patterns create the SIL. 12. The method of claim 11, wherein the SIL is created on a backside of the optical medium. 13. An apparatus for forming a solid immersion lens (SIL) in bulk silicon, comprising:a milling head configured to generate a focused ion beam; anda binary bitmap milling pattern defining locations at which the focused ion beam is projected onto a surface of the bulk silicon, wherein the locations at which the focused ion beam impact the surface of the bulk silicon are located completely within the binary bitmap milling pattern and are randomized over successive raster scans of the surface of the bulk silicon by relative rotation between the rastering axes of the focused ion beam and the surface of the bulk silicon, the relative rotation occurring as the focused ion beam follows a boustrophedon pattern. 14. The apparatus of claim 13, further comprising using the binary bitmap milling pattern to define first locations where the focused ion beam impacts a surface of the bulk silicon and to define second locations where the focused ion beam is prevented from impacting the surface of the bulk silicon. 15. The apparatus of claim 13, wherein the successive raster scans comprise electronically rotating the focused ion beam relative to the surface of the optical medium. 16. The apparatus of claim 13, wherein the successive raster scans comprise eucentrically rotating the surface of the optical medium relative to an axis along which the focused ion beam is projected. 17. The apparatus of claim 14, wherein each successive raster scan comprises a binary bitmap milling pattern having a different distribution of pixels that impact the surface of the bulk silicon and that are prevented from impacting the surface of the bulk silicon than a previous binary bitmap milling pattern, each different distribution of pixels comprising a functionally equivalent binary bitmap milling pattern. 18. The apparatus of claim 17, wherein the binary bitmap milling pattern randomizes a dwell time of the focused ion beam. 19. The apparatus of claim 18, wherein the focused ion beam covers an area of 3 pixels by 3 pixels and 14 binary bitmap milling patterns. 20. The apparatus of claim 19, wherein the SIL is created on a backside of the bulk silicon. |
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abstract | The invention concerns a method of estimating when dryout may occur in a nuclear light water reactor of the boiling water reactor kind. The method includes the use of a formula which expresses the local dryout property of the nuclear reactor. The formula includes at least a first and a second factor. The first factor is a first function that describes how the dryout property depends on the flow of the cooling medium through the nuclear fuel arrangement. The second factor is a second function that describes how the dryout property depends on the axial power profile of the nuclear fuel arrangement. The first and the second functions describe said flow dependence and said axial power profile dependence independently of each other. The invention also concerns a nuclear energy plant, a computer program product (23) and a method of operating a nuclear energy plant. |
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051046113 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The numeral 10 generally designates the heat exchanger or steam generator which is part of a nuclear steam supply system. A nuclear reactor, generally designated 2 delivers to fluid through a conduit or hot leg 14 from an active core region 16 in the reactor. The fluid enters the reactor through a primary inlet 18 for circulation through core region 16 and out to the hot leg 14. Other details of reactor 12 are well known and not necessary to the present discussion. The steam generator 10 has a generally cylindrical vessel 20 which includes a horizontal tubesheet 22. The vessel 20 has a primary side cavity divided by a divider wall (not shown) into an inlet cavity 24 and outlet cavity 26. The hot leg 14 communicates fluid to the primary side inlet cavity 24. Tubesheet 22 separates the primary side cavity portions 24 and 26 from a secondary side cavity 28. A plurality of U-shaped tubes 30 called a "bundle" are mounted in and extend through the tubesheet 22 in fluid communication, with the primary side cavity, each tube with one end in communication with the primary side inlet cavity 24 and the other with the primary side outlet cavity 26. THe tubes 30 extend into the secondary side cavity 28 to provide a heat exchanging relationship between fluid in the tubes 30 and fluid in the secondary side cavity 28. The secondary side cavity 28 includes an integral "economizer" generally designated by the numeral 32. Economizer 32 is made up of a plurality of baffle plates 34 and passages formed thereby and includes a vertical flow divider plate 36. The economizer 32 preheats incoming feedwater in the outlet side or cold leg side of the tubes 30 in the secondary side cavity 28 on the opposite side of tubesheet 22 from primary outlet cavity portion 26. The secondary side feedwater enters through an upper feedwater inlet 38a and a lower feedwater outlet 38a. An auxiliary feedwater inlet 38c is of no consequence to the invention description. From the feedwater inlets 38a and 38b, fluid is distributed 180.degree. circumferentially around the cold leg side of the tube bundle made of U-shaped tubes 30 and radially inwardly beneath flow distribution baffles 34. Annular gaps in the baffles 34 distribute flow evenly above the baffles around the tubes 30 and upwardly through the bundle in counterflow to primary coolant inside the tubes 30. Feedwater is thus preheated for the first portion of its travel above tubesheet 22. This economizer 32 of the steam generator 10 is separated from the evaporation section in the hot leg side of the tube bundle by a secondary side flow divider plate 36 of with "X" which is attached to the wall of vessel 20 with tongue-and-groove mechanical connections 39 and extends downwardly to the tubesheet 22. The relocation of secondary side handholes 40 to align with the ends of the secondary side flow divider plate 36 is a significant enhancement for field maintenance. It provides improved access for sludge lancing, for annulus inspection and foreign object search and retrieval. The present invention recognizes that with the relocation of the handholes, a problem of access to the full circumference of the tube bundle at that elevation could only be solved by providing a notch 42 at each of the ends of the secondary side divider plate 36. The present invention s a flow blocker and a heat exchanger with a flow blocker to prevent fluid bypass around the end of plate 36 through the notches 42. Each notch 42 is defined by opposed edges 44, which are arcuate and concentric with the adjacent handhole 40. A flow blocker, generally designated 50, is substantially cylindrical, is concentric with the adjacent handhole 40 and notch arcuate surfaces 42 and fits snugly within them. The flow blocker 50 includes a hollow inner cylindrical member 52 having a skirt 54, and end wall 56, with a recess 58, and a centrally apertured transverse flange 60. On the outside of the member 52, telescopically assembled thereto, it a hollow outer cylindrical member 62 into which a spring 64 has been assembled between an outer end wall 66 and the flange 60 of member 52 to bias them in opposite directions. The spring 64 is preloaded by means of a hex-headed rod 68 threaded in end wall 66 of member 62 and seated in recess 58 of member 52. A bolt flange 70 permits attachment of member 62 by means of bolts 72 to a closure plate 74 for handhole 40. The basic components of the flow blocker device 50 consist of commonly used parts such as cylinders springs, threaded rods and bolts. However, the assembled parts are unique in their arrangement for the purpose being used of allowing previously unattainable ease of access to an economizer secondary side for inspection and maintenance. |
048428108 | claims | 1. A nuclear power plant comprising: a cylindrical prestressed concrete pressure vessel defining a cavity exhibiting a liner; a high temperature reactor arranged eccentrically in said cavity; a pile of spherical fuel elements arranged in a core of said reactor; said reactor including a graphite reflector surrounding and defining said core, and means for inserting a plurality of absorber rods into a lateral portion of said reflector and into said core; a cold gas collector chamber located above said high temperature reactor configured to withstand hot gas temperature levels; thermal insulation and liner cooling means for removal of decay heat at an elevated temperature arranged on an inside surface of said pressure vessel; at least two heat exchangers arranged in parallel, and adjacent to said reactor and elevated with respect to said core wherein said heat exchangers are arranged for removal of operational and decay heat; a blower associated with each heat exchanger arranged in a coolant flow path subsequent to said heat exchanger and configured for downward coolant flow through said pile; a power generation water-steam loop connected to a secondary side of each heat exchanger. an external circulation loop connected to said liner cooling means; an elevated reservoir connected to said external cooling loop; a recooling circulating loop connected to said reservoir; and an additional heat sink connected to said recooling circulating loop. means for introducing fluid to said liner cooling means connected to said external circulating loop. 2. A nuclear power plant according to claim 1, wherein each blower is installed vertically above an associated heat exchanger in a passage in said prestressed concrete pressure vessel. 3. A nuclear power plant according to claim 1, further comprising means for introduction of feed water connected to each heat exchanger located outside said pressure vessel. 4. A nuclear power plant accordng to claim 2, further comprising passage closures located in said passages in said pressure vessel and wherein the prestressed concrete pressure vessel exhibits a configuration and prestressing system arranged to avoid imparting undue stresses on said closures. 5. A nuclear power plant according to claim 1, wherein said thermal insulation is optimized for minimal heat transfer during power operation and optimum heat transfer during the removal of decay heat. 6. A nuclear power plant according to claim 1, configurated so that plant capacity may be increased by addition of heat exchangers and associated blowers. 7. A nuclear power plant according to claim 1, further comprising: 8. A nuclear power plant according to claim 7, further comprising: |
050858254 | abstract | A multiple liquid standby safety injection system for nuclear fission reactor plants comprising means for injecting supplemental coolant water into the nuclear reactor pressure vessel to cool the fuel core and means for injecting a water solution of a neutron absorbing compound into the nuclear reactor pressure vessel about the fuel core to diminish the fission reaction. The coolant water and solution of neutron absorbent each comprise individual systems and are conveyed from their respective supply container by means of pressurized propelling gas. The individual standby safety injection systems for coolant water and solution are integrated with means for transferring propelling gas from one supply container to the other to enhance the source and available volume of liquid propelling gas for either system by drawing from the other. |
042499941 | abstract | An electromagnetic filter of the type provided with a magnetizable packing and placed in the water circuit of a nuclear reactor is cleaned by a method which first consists in isolating the filter from the circuit, then subjecting it to a series of washing and draining-off cycles. The washing operation consists in withdrawing a fraction of water from the circuit and introducing it substantially at the temperature and pressure of withdrawal and under such conditions as to impart turbulent flow to the water within the packing. The draining-off operation consists in discharging the wash water contained in the filter. |
045254966 | abstract | A water-in-oil emulsion of a water-soluble polymer such as acrylamide/acrylic acid copolymer having improved invertibility is prepared by (1) adding a portion of an inverting surfactant such as sorbitan trioleate having 20 polyoxyethylene units to a water-in-oil emulsion of water-soluble monomer, (2) polymerizing the monomer and (3) adding a remaining amount of inverting surfactant to the emulsion. |
039492311 | summary | BACKGROUND OF THE INVENTION The present invention relates generally to an infrared radiator unit, and more particularly to an infrared radiator unit for infrared analyzers. Infrared radiator units of this type are used in infrared analyzers and are electrically heated to approximately 1000K, so that the energy radiated by them is in the infrared portion of the spectrum. When such units are used in infrared analyzers for measuring purposes, specific requirements are made of them with respect, inter alia, to the direction of the infrared radiation and the constancy of the emitted radiation, as well as with respect to the energy requirements of the source. It is known that a particularly high effectiveness of the radiation utilized for measuring purposes can be obtained, if the infrared radiation which leaves the radiator unit issues in axially parallel condition, because this assures that the largest part of the radiation will enter axially parallel into the measuring or reference receptacle, rather than entering into it at an angle to impinge upon its side walls and become partly absorbed therein. To obtain this direction of the radiation it is known in the prior art to configurate the reflecting surface of the reflector of the unit as a paraboloid. However, the energy sources used in the art are not point sources but have an elongated configuration, so that a purely paraboloid-shaped reflecting surface does not adequately condense the radiation into a direct beam. Another prior-art difficulty has been the very substantial influence of heat losses via the mountings of the energy source and the unit per se, upon the constancy of the radiation intensity. The better thermal conductivity there is between the energy source, the mount for the energy source and the housing, the more substantially the temperature of the energy source will be influenced by the ambient temperature, and this in turn leads to a wavelength shift in the major portion of the emitted radiation, so that the intensity of radiation is not constant. Furthermore, it is desired that such infrared radiating units should require as little energy as possible for the operation, a condition which is particularly important if such units are employed in battery-operated infrared analyzers where the available battery energy is strictly limited. Finally, another problem that has not been solved in the art is the mechanical stability of the mounting arrangement for the radiant energy source. If the source is shifted in any way in its position relative to the optical axis of the analyzer, by mechanical vibrations or the like, then the symmetry of the arrangement is disturbed and errors in measurement can and will occur. SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide an improved infrared radiating unit for infrared analyzers which overcomes the disadvantages of the prior art. More particularly, it is an object of this invention to provide such an improved infrared radiating unit wherein the radiant energy is emitted largely in axial parallelism. A further object of the invention is to provide such a unit which has a constant radiation output as a result of low heat losses. In additional object of the invention is to provide such a unit which requires relatively little energy for its operation. Still a further object of the invention is to provide a unit of the type in question wherein the energy source is mounted in such a manner as to be highly stable in its position relative to the optical axis, and in which the maintenance of thermal symmetry about the optical axis of the analyzer is assured even if the device undergoes exceptionally strong vibrations or other agitation. In keeping with these objects, and with others which will become apparent hereafter, one feature of the invention resides in an infrared radiating unit for infrared analyzers which, briefly stated, comprises a cylindrically shaped source of infrared radiation, and a reflector partly surrounding the source and comprising a curved reflector section having a reflecting surface shaped to resemble an axially bisected rotationally symmetrical paraboloid, and a cylindrical reflector section extending from an open side of the curved reflector section. The focal points of the reflecting surface are located on a circle which coincides with the outer circumference of the cylindrically shaped source. The present invention is thus characterized by a combination of features which assures its unique advantages and characteristics as compared to the prior art. These characteristics involve the configuration and material of the reflector, the manner in which the reflector is mounted, and the carrier for the source of radiation. The reflector has a curved reflector portion, the reflector surface of which resembles an axially bisected rotationally symmetrical paraboloid which is pushed apart transversely of its bisection, and a cylindrical portion which extends from the curved portion of the reflector. The source of energy is a tubular coil or double coil of resistance wire having a cylindrical configuration, and the focal points of the individual cylinder-paraboloid sectors of the reflector surface are located on a circle which coincides with the circumference of this coil. The reflector itself is produced of a material having a comparatively low coefficient of thermal conductivity, preferably a rust-resistant or rust-free steel which is known to have a comparatively low coefficient of thermal conductivity. The inner surface, that is the reflective surface of the reflector, is advantageously gold plated. The contact area between the reflector and the mount for the same is small so as to reduce the thermal-wedging effect, that is to reduce thermal conduction between them. The carrier for the heating coil is of a ceramic material, and has a portion located outside the reflector and another portion which extends through an opening in the reflector into the interior thereof; this second portion has a smaller cross-sectional area than the remainder of the carrier. The latter is advantageously clamped in place, so that it can be shifted axially, can be turned about its longitudinal axis and can be replaced whenever desired or necessary. The heating coil may be mounted on a metallic sleeve, preferably of rust-free steel, having axially spaced flanges between which the heating coil is located. This sleeve can then be pushed onto the smaller cross-sectional area portion of the carrier and can be secured thereon and placed in suitable manner, for example by means of an adhesive. The novel features which are considered as characteristic for the invention are set forth in particular in the appended claims. The invention itself, however, both as to its construction and its method of operation, together with additional objects and advantages thereof, will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
abstract | An X-ray cassette for computed radiography is provided, said cassette having a form of a hollow box comprising top and bottom, front and rear, and lateral sides, said top and bottom sides having width dimensions, between said lateral sides, and depth dimensions, between said front and rear sides, which are substantially greater than the dimensions of said front, rear and lateral sides, between said top and bottom sides, wherein said bottom side and said front, rear and lateral sides have a higher material stiffness than the top side and wherein said top side is a deformable carrier or support material, supporting a storage or stimulable phosphor sheet layer. |
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abstract | The invention relates to a method for preparing lead (212) for medical use. This method comprises the production of lead (212) by the decay of radium (224) in a generator comprising a solid medium to which the radium (224) is bound, followed by the extraction of the lead (212) from the generator in the form of an aqueous solution A1, characterized in that the lead (212) contained in the aqueous solution A1 is purified from the radiological and chemical impurities, also contained in said aqueous solution, by a liquid chromatography on a column. The invention also relates to an apparatus specially designed for automated implementation in a closed system of said method. It further relates to lead (212) produced by means of this method and this apparatus. Applications: manufacture of radiopharmaceuticals based on lead (212), useful in nuclear medicine for the treatment of cancers, particularly by a-radioimmunotherapy, or for medical imaging, in both humans and animals. |
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047740515 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a sintered nuclear fuel compact of UO.sub.2 or the mixed oxides (U, Pu)O.sub.2 and (U, Th)O.sub.2, with a neutron poison that is incorporated in the sintered matrix, as well as a method for producing this sintered nuclear fuel compact. 2. Description of the Prior Art A sintered nuclear fuel compact of this kind along with a method for its production is known from German Pat. No. 31 44 684 and related U.S. Pat. No. 4,512,939. The neutron poison comprises a rare earth element, in particular gadolinium. To produce this sintered nuclear fuel compact, UO.sub.2 starting powder is mixed with rare earth oxide powder (Gd.sub.2 O.sub.3) and the mixture is compressed into a compact, which is subjected to a heat treatment in the temperature range of from 1500.degree. C. to 1750.degree. C. in a sintering atmosphere having a reducing effect. The holding time for this temperature may be in the range from one hour to ten hours, while the heating speed of the compact may be in the range from 1.degree. C. per minute to 10.degree. C. per minute. Fuel rods of nuclear reactor fuel elements are filled with sintered nuclear fuel compacts of this kind. The rare earth elements and gadolinium in particular are neutron poisons that can be burned up in terms of neutron physics and that lose their neutron-poisoning property after a certain period of use of the nuclear fuel element in a nuclear reactor. A nuclear fuel element is for example used in the nuclear reactor over the course of three successive fuel element cycles, which are typically of equal length. At the end of one fuel element cycle, some of the fuel elements in a nuclear reactor are replaced with fresh unirradiated nuclear fuel elements. If the fuel element cycles are relatively long, the rare earth elements, and the gadolinium in particular, in the fuel elements freshly introduced into the nuclear reactor do not burn up completely by the end of their first fuel element cycle. A so-called residual poisoning remains, which causes undesirable losses in reactivity at the beginning of the next fuel element cycle in the nuclear reactor. SUMMARY OF THE INVENTION It is the object of the invention to disclose a sintered nuclear fuel compact with which such reactivity losses are avoided during relatively long fuel element cycles. With the foregoing and other objects in view, there is provided in accordance with the invention a sintered nuclear fuel compact with which reactivity losses in a nuclear reactor having long fuel element cycles are avoided, comprising, a nuclear fuel oxide selected from the group consisting of UO.sub.2, mixed oxide (U, Pu)O.sub.2 and mixed oxide (U, Th)O.sub.2 and mixtures thereof, having a neutron poison incorporated in the sintered matrix, said neutron poison selected from the group of the chemical compound form UB.sub.x with x=2; 4 and/or 12 and the chemical compound B.sub.4 C and mixtures of such chemical compounds. In accordance with the invention, there is provided a method of producing a sintered nuclear fuel compact with which reactivity losses in a nuclear reactor having long fuel element cycles are avoided, which comprises, forming a compact of a mixture of powders containing at least one nuclear fuel oxide selected from the group consisting of UO.sub.2, mixed oxide (U, Pu) O.sub.2 and mixed oxide (U, Th)O.sub.2, and at least one neutron poison selected from the group consisting of UB.sub.x, where x=2; 4 and/or 12 and B.sub.4 C, and sintering the compact of the mixture of powders. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in sintered nuclear fuel compact and method for its production, it is nevertheless not intended to be limited to the details shown, since various modifications may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The invention, however, together with additional objects and advantages thereof will be best understood from the following description. |
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description | The present application is based on and claims priority to the Applicant's U.S. Provisional Patent Application 61/095,887, entitled “Shutter Mechanism For Collimating X-Rays,” filed on Sep. 10, 2008. Field of the Invention. The present invention relates generally to the field of shutter mechanisms for collimating x-rays. More specifically, the present invention discloses a compact shutter mechanism allowing direct, independent control of two orthogonal sets of shutters. Shutter mechanisms for x-ray collimators involve a unique set of design requirements. The shutters are typically made of a radio-opaque material, such as lead. This results in shutters that have substantial weight if they are large. In addition, x-ray diagnostic devices used in medical and dental offices are subject to very tight space limitations. Therefore, the shutter mechanism should be compact as possible to meet these space and weight constraints. In addition, the safety of the patient and healthcare providers is always of paramount concern. This dictates that the shutter mechanism must be capable of a high degree of positional accuracy to ensure an accurate x-ray dosage to the patient through a well-defined aperture, and to prevent accidental exposure of others. The prior art in the field of shutter mechanisms for collimating x-rays includes the applicant's U.S. Pat. No. 5,396,534 (Thomas). A flexible band extends in sliding engagement about the periphery of an opening in a frame. FIG. 9 of this patent shows an embodiment using two orthogonal pairs of shutters to create an aperture that is adjustable in two dimensions. This invention provides a shutter mechanism for collimating x-rays having sets of longitudinal and transverse shutters. Two flexible bands moving around a generally rectangular path about the aperture independently control the positions of each set of shutters. Two drive shafts with cams or pulleys control the positions of the bands and shutters. These and other advantages, features, and objects of the present invention will be more readily understood in view of the following detailed description and the drawings. Turning to FIGS. 1 and 2, front isometric views are providing showing the present shutter mechanism with its shutters in the fully open and closed positions, respectively. FIG. 3 is a corresponding rear isometric view and FIG. 4 is a top view of the shutter mechanism with the shutters in the fully open position. In particular, the present invention employs a pair of longitudinal shutters 1 and an orthogonal pair of transverse shutters 2, which define a generally rectangular aperture 20 for the passage of x-rays. For example in FIG. 1, x-rays would pass through the aperture 20 in a generally vertical direction. It should be noted that the present invention is not limited to collimating x-rays. It could also be used for collimating other types of radiation or light. The shutters 1, 2 are generally planar. The longitudinal shutters 1 are mounted in a common plane, and the transverse shutter 2 are mounted in a common plane parallel to that of the longitudinal shutters 1, but slightly offset to allow the shutters 1, 2 to open and close without mechanical interference. The shutters 1, 2 can be made of plates of a radio-opaque material, such as lead. The shutters 1, 2 are supported by, and slide along two orthogonal pairs of shafts 8, 9 and 12, 13. Of these, one pair of parallel shafts that are spaced apart another are designated as the drive shafts 8 and 9. More specifically, the transverse shutter drive shaft 9 and the longitudinal shutter drive shaft 8 are used to drive the positions of the transverse and longitudinal shutters 2, 1, respectively, as will be discussed below. The other pair of parallel shafts 12, 13 that are space apart from one another are referred to as the guide shafts. The guide shafts 12, 13 are used only for guiding the longitudinal shutters 1. The orthogonal relationship of these pairs of shafts 8, 9 and 12, 13 is shown most clearly in the top view provided in FIG. 5. However, the plane of the guide shafts 12, 13 is slightly offset from the plane of the drive shafts 8, 9 to allow the shutters 1, 2 to operate without mechanical interference. The term “shaft” should be broadly construed to encompass any type of elongated member, regardless of its cross-sectional shape. The positions of the shutters 1, 2 are separately controlled by rotating the drive shafts 8, 9. Knobs 5 and 6 can be attached to the ends of the drive shafts 8, 9 to simplify manual adjustment of the drive shafts 8, 9 and to provide visual indicators of the shutter positions. Two flexible bands 10, 11 extend around the perimeter of the shutter mechanism along generally rectangular paths. Alternatively, the flexible bands 10, 11 could be configured to follow other path shapes. The bands 10, 11 are also spaced apart vertically (i.e., along the axis of the x-rays) from one another. In other words, these flexible bands 10, 11 are typically parallel to, but spaced apart from one another in planes that are orthogonal to the axis of x-rays passing through the collimator. Cables, wires, strings, belts or other types of flexible members could be used as the bands 10, 11. Four guide wheels or pulleys 7 on the outer corners of the shutter mechanism allow the bands 10, 11 to freely slide about their respective paths. Fixed guides or tracks could also be used at the outer corners of the shutter mechanisms in place of wheels or pulleys. In the embodiment shown in the accompanying drawings, the upper band 10 is employed to move the transverse shutters 2 (as will be described below), and can be referred to as the transverse band 10. Similarly, the lower band 11 moves the longitudinal shutters 1, and can be referred to as the longitudinal band 11. The rectangular path of each flexible band 10, 11 has two opposing edges adjacent and parallel to two of the shafts 8, 9 and 12, 13. More specifically, the opposing edges of the transverse band 10 are adjacent and parallel to the drive shafts 8, 9, while the opposing edges of the longitudinal band 11 are adjacent and parallel to the guide shafts 12, 13. The length and tension of the bands 10, 11 can be adjusted by screw adjustment mechanisms 14 or tensioned by springs. Drive cams or pulleys 3, 4 can be mounted on each of the drive shafts 8, 9 to drive the bands 11, 10. For the purposes of this application, the terms “cam” or “pulley” should be broadly construed to include any type of rotational device allowing a shaft to drive or control the position of a flexible band. In the embodiment shown in the figures, the transverse drive cam 4 is mounted on the rear end of the transverse drive shaft 9, so that the transverse drive cam 4 is aligned with the transverse band 10 along the rear edge of its rectangular path. As shown in FIG. 6, rotation of the transverse drive shaft 9 rotates the transverse drive cam 4, which in turn drives the transverse band 10 to translate about its rectangular path either in a clockwise or counter-clockwise direction, depending on the direction of rotation of the transverse drive shaft 9. In contrast, the longitudinal drive cam 3 is mounted at the front end of the longitudinal drive shaft 8, so that the longitudinal drive cam 3 is aligned with the longitudinal band 11 along the front edge of its rectangular path, as shown in FIG. 6. Rotation of the longitudinal drive shaft 8 rotates the longitudinal drive cam 3, which in turn drives the longitudinal band 11 to translate about its rectangular path. It should be noted that the direction of motion of the longitudinal band 11 will be opposite of that for the transverse band 10 because the longitudinal drive cam 3 and transverse drive cam 4 are on opposite sides of the rectangular paths of their respective bands 11 and 10. These drive cams 3, 4 could be placed on the same side of the shutter mechanism so that the bands 11, 10 would be driven in the same direction by rotation of their respective drive shafts 8, 9. Alternatively, the drive cams 3, 4 could be rotatably mounted to other portions of the frame or support structure of the shutter mechanism, other than the drive shafts 8, 9. The longitudinal shutters 1 are attached to longitudinal band 11 along the front and rear edges of its rectangular path (i.e., adjacent and parallel to the guide shafts 12, 13). Similarly, the transverse shutters 2 are attached to the transverse band 10 along the two lateral edges of the transverse band 10 (i.e., adjacent and parallel to the drive shafts 8, 9). More specifically, a shutter attachment 16 is secured to one lateral edge of each shutter 1, 2. This shutter attachment 16 includes a clip that is secured to one of the flexible bands 10 or 11 and causes the shutter attachment 16 and shutter 1, 2 to slide with movement of the band 10, 11. The shutter attachments 16 for the longitudinal shutters 1 are attached to the opposing front and rear edges of the rectangular path of the longitudinal band 11. This causes the longitudinal shutters 1 to move in opposing directions as the longitudinal band 11 moves, thereby either opening or closing the longitudinal shutters 1. Similarly, the shutter attachments 16 for the transverse shutters 2 are attached to the opposing lateral edges of the rectangular path of the transverse band 10. This also causes the transverse shutters 2 to move in opposing directions as the transverse band 10 moves, thereby either opening or closing the transverse shutters 2. Each shutter attachment 16 also has a horizontal channel or hole that slides along at least one of the drive shafts 3, 4 or guide shafts 12, 13 to maintain proper alignment of the shutter 1, 2 with respect to the remainder of the shutter mechanism. For the purposes of this application, the term “channel” should be broadly interpreted to include any type of channel, hole, track or sliding mechanism that allows a shutter to slide along a shaft. The channels in the shutter attachments 16 for the longitudinal shutters I slide along the guide shafts 12 and 13. The channels in the shutter attachments 16 for the transverse shutters 2 slide along the drive shafts 7 and 8. Each shutter 1, 2 is also equipped with a slider 17 on its other lateral edge opposite from the shutter attachment 16. This slider 17 has a U-shaped channel that receives one of the drive shafts 3, 4 or guide shafts 12, 13. Each shutter 1, 2 is supported between a shutter attachment 16 on one side, and a slider 17 on the other side, which both slide along a pair of parallel shafts 3, 4 or 12, 13. In other words, each shutter 1, 2 is suspended from, and slides along a pair of parallel shafts 3, 4 or 12, 13. The longitudinal shutters 1 slide along the guide shafts 12 and 13, and the transverse shutters 2 slide along the drive shafts 3 and 4. Finally, a frame or support structure would usually be necessary to support the components discussed above. FIGS. 7 and 8 show front and rear isometric views of the shutter mechanism including one possible type of support structure. It should be understood that other types of support structures could be readily substituted. The above disclosure sets forth a number of embodiments of the present invention described in detail with respect to the accompanying drawings. Those skilled in this art will appreciate that various changes, modifications, other structural arrangements, and other embodiments could be practiced under the teachings of the present invention without departing from the scope of this invention as set forth in the following claims. |
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047012986 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows a steel reinforced concrete subterranean cylindrical pressure vessel 1. The vessel encloses a cavity 2. The vessel 1 has a center opening 4, in a cover region 3 which is closed by a removable cover 5. The cavity 2 houses a nuclear reactor 6. The core of the reactor 6 is a stationary pile 7 of spherical fuel elements. The pile 7 has a diameter of approximately 1.2 to 1.5 m and a height of 1.5 to 2.5 m. The power density in the core is approximately 4 to 6 MW/m.sup.3 and has a total capacity yield of 10 to 20 MW. The fuel elements, which are prepared by a hot or cold press method, contain approximately 20 to 40 g heavy metal per pellet. The pile 7 is surrounded, on all sides, by a graphite reflector 8, having a bottom reflector 9, a side reflector 10 and a roof reflector 11. The graphite reflector 8 is approximately 0.75 to 1.0 m thick. The roof reflector 11 rests directly on the pile 7. A free space 12 is located between the roof reflector 11 and the cover 5. Another free space 13 is provided between the bottom reflector 9 and the bottom of the pressure vessel 1. A metal support installation 14 is located in the free space which supports the nuclear reactor 6 on the bottom of the pressure vessel. A blower 15 circulates cooling gas, preferably helium, which flows from top to bottom through the pile 7. The blower 15 is located in a vertical position centrally under the cover 5, its rotor protrudes into the free space 12. The blower drive motor 27 equipped with an external closure part 17 is installed in a passage 16 of the cover 5. A steel core vessel 18, closed at the bottom, laterally surrounds the pile 7 and houses part of the side reflector 10 and bottom reflector 9. The side and bottom reflectors are divided into inner 19, 22 and outer 20, 23 reflector portions respectively. A plurality of vertical channels 28 are provided in the inner side reflector 19. Absorber rods 21 for trimming and shutdown are located in a displaceable manner in the channels 28. Drives 29 for the absorber rods 21 are provided in passages 38 of the cover 5. The core vessel 18 together with the inner side reflector 19, the roof reflector 11, the fuel elements and the absorber rods 21, may be removed from above following removal of the cover 5. A shielding bell is used in the process. The absorber rods 21 assure and maintain a subcritical state of the pile 7 during the removal and installation process. The core vessel 18 is removed after the fuel elements are sufficiently burned off. A cooling system 24 is mounted over the entire inner side of the pressure vessel 1. The cooling system 24 is made up of pipes through which cooling water flows and is designed for safe removal of heat generated in the pile 7 during normal operation and during the removal of the decay heat. A gas tight jacket 25 is provided in the cavity 2 in front of the cooling system 24 to prevent entry of water into the primary loop. An annular space 26 is located between the jacket and the outer side reflector 20. A gas conduction jacket 30 is provided in the free space 12 separating the suction side 50 and compression side 51 of the blower 15. It is connected to the upper end of the core vessel 18. The output of the reactor is regulated by the rpm of the blower 15 and the secondary flow of the cooling system 24 alone, utilizing the negative temperature coefficient inherent in fuel pile reactors. The blower 15 suctions the cooling gas from the free space 12 and transports it into the pile 7. In normal operation the pressure is adjusted to approximately 8 to 10 bar. The temperature of the gas rises from 300.degree. C. to 500.degree. C. during its flow through the pile 7. The heated cooling gas passes into the free space 13 through openings in the core vessel and the bottom reflector 9. It is there distributed and moved into the annular space 26. From the annular space 26 the gas returns to the space 12. The pressure of the cooling gas is chosen so as to be higher than the pressure of the water in the cooling system 24. According to the invention a design as in nuclear reactor 6 advantageously avoids requirement of installations such as a charging device, a gas purification installation, a reactor protection system, and active regulating systems which are therefore not provided. The reactor thus has very low energy generation costs and a low necessary maintenance effort. FIG. 2 shows the entire nuclear reactor installation with the pressure vessel 1, located underground in a cavity 40, and resting on a foundation 31. A concrete shield 32 closes off the top of the cavity 40 in a light duty hall 33. The hall 33 has a gate 36 and is divided into a workshop and operating room 37 and a room 39 for the installation and removal of the core vessel 18. For installation and removal of the core a crane 34 running on a plurality of rails 35 is provided. The cavity 40 is lined with concrete. The intermediate space 41 between the wall of the cavity and the pressure vessel 1 is monitored for leakage and activity. A slight underpressure may be established in relation to the environment by a slight suction. Potential leakages are drained off discontinuously in a programmed manner. |
039416526 | abstract | For location of failed fuel cans in a sodium-cooled nuclear reactor, a gas flow under a pressure slightly higher than that of the sodium at the outlet of the fuel assemblies is supplied to the outlets of each of the assemblies in seriatim order. The resulting emulsion is raised by air-lift and collected in a tank located at a level higher than that of the outlet. The gas separates from the emulsion and a gas output from the tank is monitored to detect the presence of radioactive products therein. |
050769995 | summary | FIELD OF THE INVENTION This invention relates generally to safety systems for nuclear reactors and more particularly to passive systems for removing decay heat from water-cooled nuclear reactors. BACKGROUND OF THE INVENTION In a water-cooled nuclear reactor that has operated for a significant length of time, radioactive fission products build up in the reactor core. These fission products generate radioactive heat even when the reactor is shut down, with typical decay heat generation being one percent of the reactor full-power heat generation rate. If the reactor core is not cooled after shutdown, either by use of normal shutdown procedures or post-accident shutdown procedures, the reactor core may melt. Depending on details of reactor design and mode of reactor shutdown, decay heat removal may be accomplished by cooling the hot water reactor or allowing the reactor water to boil with steam exiting the reactor and adding makeup water to the reactor. In either case, active cooling systems are used to cool the reactor and prevent a reactor core meltdown. Active systems such as these can fail due to equipment failure or operator error as happened at Three Mile Island. What is needed is a passive decay heat removal system of high reliability that becomes automatically activated upon loss of coolant and which may have its operability verified during normal operations. Various approaches toward devising passive cooling systems for water-cooled reactors are disclosed in the following papers presented at the International Atomic Energy Agency Technical Committee Meeting on Passive Safety Features in Current and Future Water-Cooled Reactors in Moscow on Mar. 21-24, 1989: Application of Passive Systems in WWER-1000 Design Project of Increased Safety: Part I, V. I. Naletov, G. A. Tarakov, E. M. Damrin, and N. B. Trunov. PA1 Application of Passive Systems in WWER-1000 Design Project of Increased Safety: Part II, T. A. Brantova and N. S. Fil. PA1 Analysis of the Possibility to Increase WWER-440 Safety Level on the Base of Passive Systems, B. Dimitrov. The systems disclosed in these papers rely on opening of air doors to obtain cooling or depend upon flooding of the reactor by water located above the reactor containment level. Placement of an open-top box inside the pressure vessel in combination with a closed circuit heat exchanging loop as in the present invention is not disclosed in the prior art known to the applicant. SUMMARY OF THE INVENTION This invention is directed to a passive decay-heat removal system for water-cooled nuclear reactor in which an open-topped insulated box is located inside the reactor vessel and positioned below the water level that exists during normal operations of the reactor. A heat exchanger which evaporates fluid within a closed loop is disposed within the box, forming one end of a closed heat-transfer loop communicating with a condenser located outside of the pressure vessel and above the reactor water level. Heat removal by the heat-transfer loop during normal operations is limited owing to the immersion of the evaporator heat exchanger under water and to cooling of the water in the box by natural circulation heat transfer and thermal stratification in the box, which separates cold water therein from an upper layer of hot water, isolating coils of the evaporator from hot reactor water. Upon dropping of the level of the reactor coolant water, as might incur in an emergency situation, water drains from the box, exposing the heat exchanger therein to high temperature steam. The vapor produced in the evaporator then circulates through the loop to the condenser where it is condensed to liquid on the outside of the containment vessel, releasing heat to the open environment (atmosphere, lake, or soil) in the process. The resulting condensed liquid then flows by gravity back to the evaporator. This provides for continuous removal of decay heat by a system with no moving parts that might malfunction. Owing to the minimized flow, which occurs during normal reactor operation, loss of heat and the resulting loss of reactor efficiency is held to a very low level. It is, therefore, an object of this invention to provide a passive system for removal of decay heat from a water-cooled nuclear reactor. Another object is to provide such a system that becomes operable upon loss of coolant water without operator invention. Yet another object is to provide a passive heat removal system that has no moving parts. Still another object is to provide a system in which removal of heat from the reactor during normal operation is minimized. Other objects and advantages of the invention will be apparent from the following detailed description and the appended claims. |
description | This non-provisional utility application claims the benefit of prior filed provisional Application No. 61/322,870 filed Apr. 11, 2010. Application No. 61/322,870 is incorporated herein by reference. Not Applicable. Not Applicable. The present disclosure relates to the use of one or more ion beams to prepare materials for microscopic observation or spectroscopic analysis. Microscopic observational techniques include, but are not limited to, optical microscopy, scanning electron microscopy (SEM), transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), reflection electron microscopy (REM). Spectroscopic analysis techniques include, but are not limited to, x-ray micro-analysis, reflection electron energy-loss spectroscopy (REELS), electron back-scattered diffraction (EBSD), x-ray photoelectron spectroscopy (XPS), Auger electron spectroscopy (AES). Materials to be viewed under any microscopic technique may require processing to produce a sample suitable for microscopic examination. Ion beam milling of a material can produce samples that are well suited for microscopic examination. An ion beam irradiating device may generate, accelerate, and direct a beam of ions toward a sample. The impact of ions on the sample sputters material away from the area of ion impact. Furthermore the sample surface may be polished by the ion beam to a substantially smooth condition further enhancing observational properties of the sample. Regions of interest in the sample may be exposed and polished by the use of ion beams thus making a suitable observational sample from the material under investigation. Broad Ion Beam Slope-Cutting (BIBSC), also known as cross-section cutting using broad ion beam sources or cross-section polishing using broad ion beam sources, is a rapid method for removing sample material to expose a smooth and substantially artifact-free cross-sectional surface for ultimate analysis by various microscopies and spectroscopies. A notable advantage of the BIBSC technique is high rates of surface preparation that can exceed tens or hundreds or thousands of square microns per hour, often over sample milling times of tens or hundreds of minutes. Important considerations to users of the BIBSC technique include: reducing or minimizing the time and effort the user is occupied in processing the sample; reducing or minimizing the number of steps where delicate samples are directly handled and at risk for damage, such as during mounting to sample holders for processing or analysis; reducing or minimizing the time and effort the user is occupied transferring the sample into the ultimate analysis equipment (imaging or spectroscopy), and aligning the coordinates of the prepared sample region to the ultimate analysis equipment prior to analysis; ensuring high quality and high probability of success in processing and imaging the sample; reducing or minimizing the time that the BIBSC ion milling equipment and sample mounting equipment is occupied for each sample; and ensuring high-quality microscopy observation of the sample during sample mounting and ultimate analysis by reducing the working distance required between the sample and the objective or probe forming lens used for observation. In consideration of the foregoing points it is clear that embodiments of the present disclosure confer numerous advantages and are therefore highly desirable. The present disclosure is directed to ion beam sample preparation apparatus and methods for using the disclosed apparatus to prepare samples for later observation. Also disclosed are apparatus to quickly and repeatably retain and release both unprepared samples and prepared samples thereby facilitating preparation of samples in the ion beam apparatus and also facilitating the observation of the prepared samples in an observation apparatus. Features of the disclosure enable accurate and repeatable positioning of the sample both within the ion beam sample preparation apparatus and also within observation apparatus later used for observing prepared samples. An apparatus for preparing a sample using an ion beam according to an embodiment of the present disclosure comprises: an ion beam irradiating means disposed in a vacuum chamber and directing an ion beam toward a sample; a shield retention stage disposed in the vacuum chamber; said shield retention stage comprising: a first datum feature, a second datum feature, and, a shield retention means having at least a shield releasing position and a shield retaining position; a shield having at least a rigid planar portion, the shield removeably and replaceably held in said shield retention stage, said shield further comprising: a proximal sample surface for durably adhering the sample to the shield, a first shielding surface disposed in the path of the ion beam and positioned to shield a portion of the ion beam directed at the sample when said shield is held in the shield retaining position of the shield retention means, a third datum feature formed integrally with said shield, wherein said shield retention means in said shield retaining position urges said third datum feature to abut said first datum feature; and, a fourth datum feature formed integrally with said shield, wherein said shield retention means in said shield retaining position urges said fourth datum feature to abut said second datum feature. In a related embodiment of the ion beam sample preparation apparatus, the shield retention stage further comprises a fifth datum feature, and the shield further comprises a sixth datum feature formed integrally with the shield, wherein the shield retention means in said shield retaining position urges said sixth datum feature to abut said fifth datum feature. In a related embodiment of the ion beam sample preparation apparatus first shielding surface meets said proximal sample surface at an angle of less than about 90 degrees and more than about 80 degrees. In a related embodiment of the ion beam sample preparation apparatus the first shielding surface meets said proximal sample surface at an angle of less than about 87 degrees and more than about 83 degrees. In a related embodiment of the ion beam sample preparation apparatus the first shielding surface is made of non-magnetic material with low sputtering yield. In a related embodiment of the ion beam sample preparation apparatus at least a portion of the first shielding surface is made of tantalum or titanium. In a related embodiment of the ion beam sample preparation apparatus the third datum feature is a datum surface and at least a portion of said datum surface is coextensive with at least a portion of said proximal sample surface. In a related embodiment of the ion beam sample preparation apparatus the proximal sample surface has at least one recessed portion for the flowing of adhesive between the shield and the sample. In a related embodiment of the ion beam sample preparation apparatus the shield further comprises: a second shielding surface having a portion disposed in the path of a portion of the ion beam; a shield edge formed where the first shielding surface meets the proximal sample surface; and, a visible alignment mark on the second shielding surface, configured such that the location of said visible alignment mark is in a predetermined relationship to the region where the ion beam impinges on said shield edge when said shield is held in the shield retaining position of the shield retention means. In a related embodiment of the ion beam sample preparation apparatus the shield is made of a cladding material joined to a core material such that a portion of the cladding material forms at least a portion of the first shielding surface, and a portion of the core material forms the third and fourth datum features of the shield. An apparatus for preparing a sample using an ion beam according to another embodiment of the present disclosure comprises: an ion beam irradiating means disposed in a vacuum chamber and directing an ion beam toward a sample; a shield retention stage disposed in the vacuum chamber; said shield retention stage comprising: a first datum feature, a second datum feature, and a shield retention means having at least a shield releasing position and a shield retaining position; a shield having at least a rigid planar portion, removeably and replaceably held in said shield retention stage, said shield further comprising: a proximal sample surface, a sample clamping means coupled to said shield and configured to hold the sample against said proximal sample surface, a first shielding surface disposed in the path of the ion beam and positioned to shield a portion of the ion beam directed at the sample when said shield is held in the shield retaining position of the shield retention means, a third datum feature formed integrally with said shield, wherein said shield retention means in said shield retaining position urges said third datum feature to abut said first datum feature; and, a fourth datum feature formed integrally with said shield, wherein said shield retention means in said shield retaining position urges said fourth datum feature to abut said second datum feature. An apparatus for preparing a sample using an ion beam according to another embodiment of the present disclosure comprises: an ion beam irradiating means disposed in a vacuum chamber and directing an ion beam toward a sample; a rotating shield retention stage disposed in the vacuum chamber; said shield retention stage comprising: a first datum feature, a second datum feature, a shield retention means having at least a shield releasing position and a shield retaining position, a rotation axis located substantially in the plane of the first datum feature, and a rotation drive for rotating the shield retention stage around the rotation axis; a shield having at least a rigid planar portion, removeably and replaceably held in said shield retention stage, said shield further comprising: a third datum feature formed integrally with the shield, wherein said shield retention means in said shield retaining position urges said third datum feature to abut said first datum feature, a fourth datum feature formed integrally with the shield, wherein said shield retention means in said shield retaining position urges said fourth datum feature to abut said second datum featur, a first shielding surface disposed in the path of the ion beam and positioned to shield a portion of the ion beam directed at the sample when said shield is held in the shield retaining position of the shield retention means, a proximal sample surface for durably adhering the sample to the shield, and a shield edge formed where the first shielding surface meets the proximal sample surface, wherein said shield edge is held substantially perpendicular to said rotation axis when said shield is held in the shield retaining position of the shield retention means. A kit for preparing a sample in an ion beam sample preparation device according to another embodiment of the present disclosure comprises: a shield retention stage comprising: a first datum feature, a second datum feature, a shield retention means having at least a shield releasing position and a shield retaining position; a shield having at least a rigid planar portion, removeably and replaceably held in said shield retention stage, said shield further comprising: a third datum feature formed integrally with said shield, wherein said shield retention means in said shield retaining position urges said third datum feature to abut said first datum feature, a fourth datum feature formed integrally with said shield, wherein said shield retention means in said shield retaining position urges said fourth datum feature to abut said second datum feature, a first shielding surface, and a proximal sample surface for durably adhering the sample to the shield. A kit for observing in a microscope a sample prepared in an ion beam sample preparation device according to another embodiment of the present disclosure comprises: a shield retention stage comprising: a first datum feature, a second datum feature, a shield retention means having at least a shield releasing position and a shield retaining position; a shield having at least a rigid planar portion, removeably and replaceably held in said shield retention stage, said shield further comprising: a third datum feature formed integrally with the shield, wherein said shield retention means in said shield retaining position urges said third datum feature to abut said first datum feature, a fourth datum feature formed integrally with the shield, wherein said shield retention means in said shield retaining position urges said fourth datum feature to abut said second datum feature, a first shielding surface; and, a proximal sample surface for durably adhering the sample to the shield. 2—ion beam sample preparation apparatus 8—sample 10—vacuum chamber 18—chamber cover 20—ion beam irradiating means 22—central ion beam axis 40—shield retention stage 42—shield retention means 42a—shield retention means first member 42b—shield retention means second member 46—shield retaining position 48—shield releasing position 50—rotating shield retention stage 52—rotation drive 54—rotation axis 56—vacuum seal 60—shield 61—shielding surface 61a, 61b, etc.—first shielding surface, second shielding surface, etc. 62—proximal sample surface 63—shield edge 64—recessed portion 65—visible alignment mark 66—core material 67—cladding material 68—sample clamping means 70a, 70c, 70b, 70d, 70e, 70f—first datum feature, second datum feature, third datum feature, fourth datum feature, fifth datum feature, sixth datum feature. 72—datum surface 90—vacuum pump means 92—pumping manifold The Broad Ion Beam Slope-Cutting (BIBSC) sample preparation procedure can be described as a series of process steps, p1-p5: p1) Aligning the desired region of the sample to be processed to a usable portion of an ion shield; p2) Aligning the sample and shield in the BIBSC ion-milling system such that the desired region of the sample can be processed by the ion beam or beams; p3) Evacuating the ion-milling system to vacuum levels appropriate for ion beam milling; p4) Performing the ion-milling operation or operations, sometimes using a process monitoring step such as in situ light-microscopy imaging to verify sufficient cut depth and quality of the cross section; p5) Venting of the BIBSC ion-milling equipment and removal of the sample from the equipment. The analysis of prepared BIBSC sample can be described as a series of process steps, p6-p9: p6) Introduction of the sample to the ultimate analysis microscope and initializing the microscope so that analysis can commence; p7) Finding the location of the prepared cross-sectional surface by adjusting any number of the microscope's translation stages, tilt stages, and rotation stages so that the desired area can be imaged; p8) Performing the desired microscopic or spectroscopic analyses; p9) Removing the sample from the microscope; p10) After analyzing the sample, a decision may be made to reprocess the sample to change the cut depth, position, or angle—traditionally requiring a repeat of p1-p9. Embodiments of the present disclosure uniquely permit certain efficiencies and capabilities in the processing and subsequent observation and analysis of BIBSC produced samples. Beneficial features, functions, and aspects of the present disclosure include, but are not limited to: 1. Datum features on the shield, shield retention device in the sample-to-shield mounting apparatus, shield retention device in the BIBSC ion-mill, shield retention device in the ultimate analysis equipment allow significant time efficiencies in processing steps p1, p2 and p7; 2. The integral nature of the sample durably adhered to the shield, and to a lesser extent with the sample merely clamped to the shield, allows greater certainty in ensuring alignment of the shield to the sample remains consistent during p4 even over long time-scales and changes in temperature, whereas quality of the cross section cutting process is reduced if this precision alignment is not maintained; 3. The integral nature of the sample durably adhered to the shield in processing step p1 eliminates the requirement for expensive and sizable fixturing apparatus to maintain their spatial relationship together throughout the milling operation, and enables multiple samples to be prepared in advance of milling without multiple fixturing apparatus; 4. The integral nature of the sample durably adhered or clamped to the shield eliminates the requirement for dismounting the sample from the shield prior to observation in a microscope, even in cases where the smallest working distances between imaging objective and sample are employed. This permits reduction of both time and risk of damage to the sample during sample remounting in processing step p6; 5. In the case where reprocessing the sample as in step p10 is performed, the integral nature of the sample durably adhered or clamped to the shield can eliminate the need for steps p1 and p2 entirely, which significantly reduces processing time and risk of damage to the sample during sample remounting; and, 6. In the case where reprocessing the sample as in step p10 is performed, the integral nature of the sample durably adhered or clamped to the shield allows different cross-sectional planes to be cut very close to the originally cut cross-sectional plane by varying the angle of ion beam impinging on the sample and shield. Turning now to FIG. 1, an embodiment of an ion beam sample preparation apparatus 2 according to the present disclosure is shown comprising: a vacuum chamber 10 in which a sample 8 is prepared, chamber cover 18 which seals vacuum chamber 10 from the outside atmosphere, vacuum pump means 90 and pumping manifold 92 which together bring vacuum chamber 10 to vacuum levels appropriate for ion beam milling, ion beam irradiating means 20 which creates and directs an ion beam having a central ion beam axis 22 toward sample 8, a shield 60 which shields at least a portion of sample 8 from at least a portion of the ion beam, a shield retention stage 40 which holds and accurately positions shield 60 with respect to the direction and extent of the ion beam, and a shield retention means 42 which both retains shield 60 in shield retention stage 40 and also urges shield 60 to remain in a position whereby the ion beam may prepare sample 8. With continuing reference to FIG. 1, the ion beam preferably comprises noble gas ions. Elements used for the ion beam may include but are not limited to: Argon, Xenon, and Krypton. The ion beam may also comprise a mixture of ions and neutrals. Shield retention stage 40 is disposed in vacuum chamber 10 in a predetermined position and orientation with respect to central ion beam axis 22. After the sample has been prepared in the ion beam, chamber cover 18 may be removed, then the shield and sample may be removed for observation in a microscope. FIG. 3A shows a perspective view of shield retention stage 40 on which sample 8 has been durably adhered to shield 60 prior to placing the shield and sample combination in a shield retaining position 46 in shield retention stage 40. Shield 60 has a shielding surface 61 which is positioned in relation to sample 8 to shield at least a portion of said sample 8 from at least a portion of the ion beam. Also shown in FIG. 3A is a section line indicating the section view shown in FIG. 3B. FIG. 3B shows a section view illustrating the position and function of the shield retention means which is part of shield retention stage 40. FIG. 3B shows an embodiment of the shield retention means comprising a shield retention means first member 42a and a shield retention means second member 42b. Shield retention means first member 42a urges shield retention means second member 42b against shield 60. The action of shield retention means first member also urges shield 60 against shield retention stage 40 and thereby maintains the position of shield 60 within shield retention stage 40 while the sample is prepared by ion beam. An embodiment of the shield retention means may comprise a spring for shield retention means first member 42a and a solid member as shield retention means second member 42b configured to slide within a cavity in shield retention stage 40. FIG. 4 shows a view from the same sectional plane as in FIG. 3B, however in FIG. 4 the shield and sample have been removed to show a shield releasing position 48 of shield retention means 42. By means of the two positions provided by shield retention means 42, namely shield retaining position 46 and shield releasing position 48, a shield may be removably and replaceably secured in shield retention stage 40. A sample that has been durably adhered to shield 60 may be processed, removed, and then reprocessed by simply placing it in the shield retaining position and preparing the sample again in the ion beam. FIG. 5A shows a perspective view of shield retention stage 40 on which shield 60 is retained, wherein said shield has a shielding surface 61. FIG. 5A also shows a sectional plane used for FIG. 5B. FIG. 5B shows a sectional perspective view illustrating physical features of both shield 60 and shield retention stage 40 that facilitate accurate and repeatable positioning of the shield with respect to the shield retention stage. The positioning of shield 60 assures that shielding surface 61 and shield edge 63 are accurately positioned and accurately oriented with respect to the shield retention stage and are positioned with respect to central ion beam axis 22 to intercept at least a portion of the ion beam directed toward the sample. FIG. 6 shows a sectional perspective view as in FIG. 5B in which preferred embodiments of both shield 60 and shield retention stage 40 have a plurality of datum features 70a, 70b, 70c, 70d, 70e, and 70f. In the exploded view shown in FIG. 6, shield 60 has been removed from shield retention stage 40 and the shield is turned to expose a proximal sample surface 62 upon which a sample may be durably adhered prior to sample preparation by the ion beam. The plurality of datum features 70a, 70b, 70c, 70d, 70e, and 70f are provided on both shield 60 and shield retention stage 40 and they enable accurate and repeatable positioning of the shield 60 with respect to the shield retention stage 40. Datum features 70c, 70d, and 70f on the shield are shaped and positioned such that when they are caused to abut complementary datum features 70a, 70b, and 70e on the shield retention stage the shield may be held in a predetermined position and a predetermined orientation with respect to the central ion beam axis 22. Shield retention means 42 assures that datum features 70c, 70d, and 70f of shield 60 abut the corresponding datum features 70a, 70b, and 70e of the shield retention stage 40 when the shield is held in the shield retaining position. Shield edge 63, also visible in FIG. 6, is also caused to be in a predetermined position and predetermined orientation when the shield is held in the shield retaining position. Datum features are arranged in pairs such that a datum feature on the shield has a corresponding datum feature on the shield retention stage. In FIG. 6 one such pair of datum features is datum feature 70a on the shield retention stage and datum feature 70c on the shield. Another pair of datum features shown in FIG. 6 is datum feature 70b on the shield retention stage and datum feature 70d on the shield. Another pair of datum features shown in FIG. 6 is datum feature 70e on the shield retention stage and datum feature 70f on the shield. When the shield is in the shield retaining position, the shield retention means acts to urge the pairs of datum features to abut, thereby constraining the position of the shield with respect to the position of the shield retention stage. Datum features may be datum surfaces, as is shown in the preferred embodiment of FIG. 6, or they may be datum edges or datum vertices, or combinations of datum surfaces, datum edges, and datum vertices. Turning now to figures FIG. 7A, FIG. 7B, FIG. 8A, FIG. 8B, FIG. 9, and FIG. 10, shown are various features and embodiments of shield 60 consistent with the present disclosure. FIG. 7A is a perspective view of a shield showing a first shielding surface 61a, a second shielding surface 61b, and shield edge 63. Ions from the ion beam irradiating means that are blocked by the shield, and in particular the ions that are blocked by first shielding surface 61a are prevented from milling the sample. Ions not blocked by the shield may be used to prepare the sample for observation and analysis. When the ion beam is operating ions may or may not impact second shielding surface 61b. Whether ions do impact second shielding surface 61b depends on a number a factors including, but not limited to: the size of the ion beam; the angle at which the ion beam is directed; and, the position at which the ion beam is directed. It is a preferred embodiment of the shield that second shielding surface 61b be made of the same material as first shielding surface 61a. In preferred embodiments shield 60 is a generally planar rigid member, having one or more shielding surfaces that are smooth and may be polished, having a datum surface and at least an additional datum feature for facilitating accurate placement within the shield retention stage. Preferred materials for the shield are non-magnetic metals with low sputter yield including but not limited to tantalum or titanium. Lower cost embodiments of shield 60 may comprise a core material for the majority of the shield and a cladding material used for the shielding surfaces. Preferred core materials include but are not limited to copper. Preferred cladding materials include but are not limited to tantalum or titanium. Figures FIG. 14A and FIG. 14B illustrate two different embodiments of a shield 60, wherein each embodiment is shown comprising a combination of core material 66 and cladding material 67. FIG. 7B shows the same shield as shown in FIG. 7A but from a different angle thereby illustrating the position and nature of a plurality of datum features 70d and 70f. FIG. 8A shows the same shield as shown in FIG. 7A and FIG. 7B. FIG. 8A shows a perspective view of shield 60 from the side of the shield closest to the sample during ion beam sample preparation. Proximal sample surface 62 may be used to adhere the sample material to be prepared in the apparatus. Datum surface 72 is a datum feature that is a surface. In a preferred embodiment at least a portion of proximal sample surface 62 may be coextensive with at least a portion of datum surface 72. Shield edge 63 is formed by the intersection of first shielding surface 61a and proximal sample surface 62. The angle between first shielding surface 61a and proximal sample surface 62 has an impact on the quality of milling performed on the sample by the ion beam. A preferred embodiment is achieved when said first shielding surface 61a meets said proximal sample surface 62 at an angle of less than about 90 degrees and more than about 80 degrees. An even more preferred embodiment is achieved when said first shielding surface 61a meets said proximal sample surface 62 at an angle of less than about 87 degrees and more than about 83 degrees. FIG. 8B shows the same shield as shown in FIG. 8A but from a different angle thereby illustrating the position and nature of a plurality of datum features 70d and 70f, and datum surface 72, present on shield 60. FIG. 9 shows a perspective view of shield 60 having first shielding surface 61a, second shielding surface 61b, shield edge 63, and additionally comprising a visible alignment mark 65. When the shield is held in the shield retaining position the visible alignment mark is positioned so that it indicates the approximate location where a portion of the ion beam will pass over shield edge 63 and impact the sample when the shield edge is substantially perpendicular to the central ion beam axis. FIG. 10 shows a perspective view of shield 60 from the side of the shield closest to the sample during ion beam sample preparation. Proximal sample surface 62 may be used to adhere the sample material to the shield prior to ion beam sample preparation in the apparatus. Recessed portion 64 provides a recessed portion of proximal sample surface 62 useful for flowing adhesive under the sample thereby facilitating the durable adhering of sample to shield. Preferred materials used to adhere the sample to the shield include but are not limited to: UV cured glue, light cured glue, superglue, silver paint, and wax. Turning now to FIG. 11A, shown is a perspective view of shield 60, shielding surface 61, sample 8 durably adhered to the shield, and visible alignment mark 65. FIG. 11A depicts the sample prior to ion beam preparation. FIG. 11B is a perspective view of the same objects depicted in FIG. 11A, however, FIG. 11B represents the sample after ion beam sample preparation. Shielding surface 61 intercepts a portion of the ion beam, which travels along central ion beam axis 22. A portion of sample 8 is sputtered away by the ion beam during sample preparation thereby exposing a portion of the sample lying in the plane defined by shield edge 63 and central ion beam axis 22. A sample prepared in this way will be suitable for observation or analysis with a variety of microscopic or spectroscopic techniques, particularly those requiring a highly polished planar surface. FIG. 12A and 12B illustrate another embodiment of shield 60 in which a sample clamping means 68 is formed integrally with the shield on the proximal sample surface 62. FIG. 12A depicts this shield prior to clamping a sample while FIG. 12B depicts this shield after sample 8 has been secured to the shield by means of sample clamping means 68. In another embodiment, sample clamping means 68 may be formed separately and then coupled to the shield prior to clamping the sample. Adhesive may be applied between the sample clamping means and the sample to further ensure the sample does not move with respect to the shield. Use of the apparatus shown in FIG. 1 may proceed with reference to the following steps: outside of the vacuum chamber, a sample may be durably adhered to a shield; with the chamber cover removed, the sample and shield combination may be set in the shield retaining position of the shield retention stage; the chamber cover may then be replaced; with the chamber cover in place on the the vacuum chamber the vacuum pump means may be operated to evacuate the vacuum chamber through the pumping manifold thereby obtaining vacuum levels appropriate for ion beam milling; the ion beam irradiating means may then be operated to prepare the sample. When the sample is prepared to the extent desired by the user of the apparatus the ion beam irradiating means may be turned off, the vacuum chamber may be returned to atmospheric conditions, the chamber cover may be removed, and the prepared sample may be removed from the apparatus along with the shield to which it was previously adhered. A microscope may be fitted with a shield retention stage so that the prepared sample and shield may be retained and thereby the prepared region of the sample may be observed in the microscope. After observation the user may decide that additional sample preparation is needed. Since the sample is still durably adhered to the shield, it is a simple matter to return the sample and shield to the vacuum chamber for additional processing. The datum features on both the shield and the shield retention stage ensure that the shield may be retained in substantially the same position and orientation each time the sample is processed in the apparatus. A kit comprising a shield retention stage 40 with a plurality of datum features 70a, 70b, and 70e, shield retention means 42, and at least one shield 60 with a plurality of datum features 70c, 70d, and 70f may be supplied for fitting to a microscope. Such a kit facilitates the microscopic observation of samples prepared in the ion beam sample preparation apparatus 2. With reference now to FIG. 2, an embodiment of an ion beam sample preparation apparatus 2 according to the present disclosure is shown comprising: a vacuum chamber 10 in which a sample 8 is prepared, chamber cover 18 which seals vacuum chamber 10 from the outside atmosphere, vacuum pump means 90 and pumping manifold 92 which together bring vacuum chamber 10 to vacuum levels appropriate for ion beam milling, ion beam irradiating means 20 which creates and directs an ion beam having a central ion beam axis 22 toward sample 8, a shield 60 which shields at least a portion of sample 8 from at least a portion of the ion beam, a rotating shield retention stage 50 which holds and accurately positions shield 60 with respect to the direction and extent of the ion beam, a shield retention means 42 which both retains shield 60 in rotating shield retention stage 50 and also urges shield 60 to remain in a position whereby the ion beam may prepare sample 8, a rotation drive 52 which rotates the rotating shield retention stage 50 about rotation axis 54, and vacuum seal 56 which maintains the vacuum in vacuum chamber 10 while allowing rotating shield retention stage 50 to move about rotation axis 54. The ion beam preferably comprises noble gas ions. Elements used for the ion beam may include but are not limited to: Argon, Xenon, and Krypton. The ion beam may also comprise a mixture of ions and neutrals. Rotating shield retention stage 50 is disposed in vacuum chamber 10 in a predetermined position and orientation with respect to central ion beam axis 22. Rotating shield retention stage 50 may additionally comprise means for measuring the rotation angle of the stage. Rotation drive 52 may additionally comprise means to reach and maintain accurate angular position. Rotation drive 52 may additionally comprise means to reach and maintain accurate angular speed. Shield 60 of FIG. 2 has the same features, functions, and aspects possessed by shield 60 shown in figures FIG. 3A, FIG. 3B, FIG. 5A, FIG. 5B, FIG. 6, FIG. 7A, FIG. 7B, FIG. 8A, FIG. 8B, FIG. 13A, and FIG. 13B. FIG. 13A shows a perspective schematic view of rotating shield retention stage 50 on which sample 8 has been durably adhered to shield 60 prior to placing the shield and sample combination in a shield retaining position in rotating shield retention stage 50. Shield 60 has a shielding surface 61 which is positioned in relation to sample 8 to shield at least a portion of said sample 8 from at least a portion of the ion beam. Rotation drive 52 enables rotating shield retention stage 50 to rotate about rotation axis 54. In a preferred embodiment, shield edge 63 is disposed to be substantially perpendicular to rotation axis 54 when shield 60 is held in the shield retaining position. Also shown in FIG. 13A is a section line indicating the section view shown in FIG. 13B. FIG. 13B shows a section view illustrating the position and function of the shield retention means which is part of rotating shield retention stage 50. In FIG. 13B shield retention means first member 42a urges shield retention means second member 42b against shield 60. The action of shield retention means first member also urges shield 60 against rotating shield retention stage 50 and thereby maintains the position of shield 60 within rotating shield retention stage 50 while the sample is prepared by ion beam. An embodiment of the shield retention means may comprise a spring for shield retention means first member 42a and a solid member as shield retention means second member 42b configured to slide within a cavity in rotating shield retention stage 50. The shield retention means also has a shield releasing position in which the shield and sample are not held in the rotating shield retention stage 50. The shield releasing position may be identical to the shield releasing position illustrated in FIG. 4. In preferred embodiment, rotation axis 54 lies substantially in the plane defined by the abutment of the proximal sample surface 62 with the sample 8. The rotating shield retention stage 50 has the same plurality of datum features as shown on the non-rotating shield retention stage of FIG. 6. In addition, the rotating shield retention stage 50 datum features allow the interchangeable use of shield 60 previously described. The datum features of the shield retention stage 40 and the rotating shield retention stage 50 are substantially identical in design and thereby facilitate the easy interchange of shields between the stages. By means of the two positions provided by the shield retention means, namely shield retaining position 46 and shield releasing position 48, a shield may be removably and replaceably secured in rotating shield retention stage 50. A sample that has been durably adhered to shield 60 may be processed, removed, and then reprocessed by simply placing it in the shield retaining position and preparing the sample again in the ion beam. The datum features on both shield and retention stage assure that the shield may be positioned in a substantially identical position and orientation multiple times. In preferred embodiments that include rotating shield retention stage 50 central ion beam axis 22 passes substantially through rotation axis 54 in the region above shield edge 63. After the sample has been prepared in the ion beam, chamber cover 18 may be removed, then the shield and sample may be removed for observation in a microscope. Use of the apparatus of FIG. 2 may proceed according to all of the steps disclosed for the use of the apparatus of FIG. 1. However, the rotating shield retention stage shown in the embodiment of FIG. 2 gives the user additional capabilities. The user may rotate the rotating shield retention stage about the rotation axis to a desired angle before sample preparation, during sample preparation, or after sample preparation. This gives the user increased flexibility in exposing a region of interest in the sample for later microscopic observation. Although the present invention has been described in considerable detail with reference to certain preferred versions thereof, other versions are possible. It may be desirable to combine features shown in various embodiments into a single embodiment. A different number and configuration of datum features may be used to constrain the shield to the shield retention stage and is entirely within the spirit and scope of the present disclosure. Therefore, the spirit and scope of the appended claims should not be limited to the description of the preferred versions contained herein. Any element in a claim that does not explicitly state “means for” performing a specified function, or “step for” performing a specific function, is not to be interpreted as a “means” or “step” clause as specified in 35 U.S.C. Section 112, Paragraph 6. In particular, the use of “step of” in the claims herein is not intended to invoke the provisions of 35 U.S.C. Section 112, Paragraph 6. |
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claims | 1. A method for laser alignment within a charged particle beam system, comprising:positioning a laser beam so that it passes through a laser injection port window and into a vacuum chamber toward a workpiece;based on information from a first alignment detector, adjusting the laser beam so that it is pointed on a first designated point located before an objective lens and so that it passes through the objective lens;moving a second alignment detector to a designated position in the laser beam pathway;monitoring the position of the beam on the objective lens with second alignment detector;based on information from the second alignment detectors, adjusting the laser beam so that the laser beam is pointed to a second designated point aligned with the workpiece and so that the laser beam is aligned eucentrically to the workpiece. 2. The method of claim 1 further comprising, after adjusting the beam to the second point, based on information from the first alignment detector, monitoring and adjusting the position of the laser beam again so that it is pointed to the first point. 3. The method of claim 1 wherein the process of adjusting the laser beam on the first designated point and the second designated point is done repeatedly until the laser beam is aligned to both the first point and the second point. 4. The method of claim 1 further comprising machining a spot on the workpiece with the laser and comparing to a eucentric point. 5. The method of claim 4 wherein if the laser spot does not match with the eucentric point, repeating the procedure until they match. 6. The method of claim 1 wherein adjusting the laser beam to be eucentrically aligned with the workpiece further includes adjusting the laser beam to a eucentric point of a focused ion beam system directed toward the workpiece, the method further comprising milling the workpiece with the focused ion beam system directed toward the workpiece in the vacuum chamber. 7. The method of claim 6 further comprising imaging the workpiece with an electron beam system directed toward the workpiece in the vacuum chamber to monitor the ion beam milling. 8. The method of claim 1 further comprising milling the workpiece with the laser beam. 9. The method of claim 8 further comprising imaging the workpiece with an electron beam system directed toward the workpiece in the vacuum chamber to monitor the laser beam milling. 10. A multi-beam system, comprising:a vacuum chamber;a workpiece support for supporting a workpiece within the vacuum chamber;a focused ion beam system for generating a focused ion beam, said ion beam directed toward the workpiece and operable to remove material from the workpiece;a laser beam system for generating a laser beam for processing the workpiece in the vacuum chamber;an electron beam system for monitoring the material removal process;an objective lens;a laser beam alignment system operable to adjust the position laser beam through the center of the objective lens and direct it to a eucentric point of the workpiece. 11. The multi-beam system of claim 10 in which the laser beam is operable at a fluence greater than an ablation threshold of the workpiece. 12. The multi-beam system of claim 11 in which the laser beam is a nanosecond to femtosecond pulsed laser beam. 13. The multi-beam system of claim 10 in which the laser beam operable at a fluence that reacts with the workpiece without ablation. 14. The multi-beam system of claim 10 in which the laser beam is operable to cause thermally induced chemical desorption at the workpiece. 15. The multi-beam system of claim 10 in which the laser beam is operable to cause laser photochemistry reactions at the workpiece. 16. The multi-beam system of claim 10, wherein the laser beam alignment system includes first and second alignment detectors, and is operable to, based on information from a first alignment detector, adjust the laser beam so that it is pointed on a first designated point located before the objective lens and so that it passes through the objective lens; the laser beam alignment system further operable to monitor the position of the beam on the objective lens with second alignment detector and, based on information from the second alignment detectors, adjust the laser beam so that the laser beam is pointed to a second designated point aligned with the workpiece and so that the laser beam is aligned eucentrically to the workpiece. 17. The multi-beam system of claim 16 in which the laser beam alignment system is further operable to, after adjusting the beam to the second point, based on information from the first alignment detector, monitor and adjust the position of the laser beam again so that it is pointed to said first point. 18. The multi-beam system of claim 16 in which the second alignment detector is a quad cell detector. 19. The multi-beam system of claim 16 wherein the second alignment detector is located close enough to the output of the objective lens to precisely measure the laser beam location on the objective lens and to prevent damage to the second alignment detector induced by the focused laser beam. |
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043483399 | claims | 1. Method for manufacturing oxidic nuclear fuel bodies with oxygen-to-metal ratio of 2.0.+-.0.02 at treating temperatures between 1000.degree. and 1400.degree. C., which comprises adding up to 25% U.sub.3 O.sub.8 as a grain-growth-promoting sintering additive to a powder selected from the group consisting of UO.sub.2 powder, a powder mixture of UO.sub.2 with gadolinium oxide, and a powder mixture of UO.sub.2 with PuO.sub.2, mixing the additive and said powder and pressing the additive containing mixture into blanks, exposing said blanks in a furnace for a sintering time of up to 30 minutes to a first atmosphere of a mild oxidizing gas to cause the UO.sub.2 powder to come to a constant stoichiometric excess of O.sub.2 to U and sintering said blanks therein at a temperature within the range of 1000.degree.-1400.degree. C., and subsequently exposing the sintered blanks in said furnace to a second atmosphere of hydrogen as a reducing gas at a temperature within the range of 1000.degree.-1400.degree. C. 2. Method according to claim 1, wherein the oxidizing gas is technically pure carbon dioxide. 3. Method according to claim 1, wherein moisture is added to the reducing gas to reduce residual fluorine content of the blanks. 4. Method according to claim 1 wherein the hydrogen gas is diluted with an inert gas. 5. Method according to claim 4, wherein the inert gas as is nitrogen. 6. Method according to claim 1 or claim 2, wherein the sintering time is about 30 minutes at a temperature of 1000.degree. C. and 5 minutes at a temperature of 1400.degree. C. and longer sintering times are provided for adjusting the micro structure of the nuclear fuel bodies until a prescribed micro structure is achieved. 7. Method according to claim 1, wherein the UO.sub.2 -powder is a powder with the following properties: Specific surface: 5 to 6 m.sup.2 /g Bulk density: 2 g/cm.sup.3 Average particle size: 6 .mu.m. 8. Method according to claim 1, wherein said powder is UO.sub.2 powder. 9. Method according to claim 1, wherein said powder is a powder mixture of UO.sub.2 with gadolinium oxide. 10. Method according to claim 1, wherein said powder is a powder mixture of UO.sub.2 with PuO.sub.2. |
042723213 | summary | BACKGROUND OF THE INVENTION This invention relates to devices for handling heavy components within a nuclear reactor, and in particular for removing and reinserting reactor internals and control rods during refueling. Commercial nuclear power reactors have a plurality of fuel assemblies containing fissile fuel material wherein heat is generated and transferred to the working fluid. In a typical reactor, many of the fuel assemblies must be periodically replaced in order to maintain a sufficient level of core reactivity so that the nuclear fission process can be self-supporting. Refueling usually represents the largest time period during which the reactor is not producing power, and therefore every effort is made to minimize the refueling time. The refueling operation includes removal from the reactor of internal structures such as the upper guide structure, and possibly control rods, so that the fuel assemblies are accessible for easy removal or rearrangement. Refueling is particularly time-consuming in those reactors which utilize the kind of upper guide structure described in U.S. Pat. No. 3,849,257. In such reactors, both the upper guide structure (UGS) and the control rods must be removed from the reactor before access can be had to the fuel assemblies. Previously, the control rods were individually removed from the reactor and hung in a temporary storage area. Then an overhead crane would lift the UGS and place it on a support platform. It can be appreciated that the removal of individual control rods, the removal of the UGS, the reinsertion of the UGS, and the reinsertion of the individual control rods require many manual operations, during which several persons are exposed to radiation from the reactor core. SUMMARY OF THE INVENTION It is evident that a significant reduction in the reactor refueling time can be achieved if the control rods and UGS can be removed from the reactor in a single continuous operation, stored as a unit, and reinserted in a single continuous operation. According to the invention, a method and apparatus are provided for engaging all control rods and simultaneously lifting them out of the reactor core until the rods are fully withdrawn from the core and are within the upper guide structure, after which the lifting force is automatically transferred to the UGS so that it and the control rods may simultaneously be lifted out of the reactor vessel. The UGS with the rods protected therein can then be transferred as a single unit to a storage area. The invention further provides means by which the rods are prevented from moving within the UGS during the entire period that the UGS is disengaged from the reactor vessel. After the shuffling of the fuel assemblies is complete, the UGS and control rods are reinserted into the reactor vessel in reverse sequence. |
description | This application is a National Phase of PCT/EP2014/055360, filed Mar. 18, 2014, which claims priority from European Patent Application No. 13160589.1, filed Mar. 22, 2013. The contents of these applications are incorporated herein by reference in their entirety. The present invention relates to a contour collimator for radio therapy comprising a plurality of diaphragm elements provided in a guiding block and movably arranged with respect to one another to form a contour diaphragm for a radiation beam emitted by a radiation source towards the collimator, and at least one drive for moving the diaphragm elements, wherein a drive of its own is associated with each diaphragm element, the drives of a group of diaphragm elements are arranged substantially adjacent to one another, and a driving transmission of its own is provided between each drive and the associated diaphragm element. Usually, the diaphragm elements are shaped or formed as leaves. Various embodiments of multi-leaf collimators comprising diaphragm elements leaf modules each featuring a leaf unit and a leaf drive unit are known in the art. Multi-leaf collimators of such kind are preferably employed for controlling a shape of a high-energy radiation beam emanating from a radiation source and propagating in a direction of propagation. Multi-leaf collimators are commonly used in treatment devices for oncological radiation therapy. Said collimators delimit high-energy beams, in most cases high energy radiation of a linear accelerator, in such a way that the beams have exactly the same shape as the treatment object. Since such irradiation, e.g. of a tumor, occurs from various directions, it is possible to achieve a great irradiation intensity of the tumor and, at the same time, to stress the surrounding tissue only to a limited extent. The diaphragm elements or leaves of the multi-leaf collimator may also be called “shutter blades” or “lamellae”. The multi-leaf collimators may also be called contour collimators since due to the positioning of the leaves, contours of treatment objects, for example tumors, can be recreated for each beam application, each of which occurs from a certain solid angle. This is important in order to protect the adjacent healthy tissue to the greatest extent possible. In the case of organs at risk, such as spine or nerves, this is particularly necessary in order to preserve their functional capability. A general example of a multi-leaf collimator comprising leaf modules with a leaf unit and a leaf drive unit is obtainable from U.S. Pat. No. 4,794,629. In such multi-leaf collimators, each leaf unit must be moved into a certain position. Thus, in most cases, a leaf drive unit must be assigned to each leaf unit. According to the aforementioned publication, no separate motor is assigned to each leaf unit, which is why the leaf units are arranged in series by means of drive couplings and locking devices. However, it has also been known to assign an electric motor to each leaf unit that positions the leaf unit via a pinion and a gear rod-like drive engagement. However, the more precisely the shape of the treatment object, e.g. of a tumor, is to be recreated, the more and thinner leaves within the leaf units will be required. This means that a large number of electric motors and drive transmissions within the leaf drive units being connected to the leaf units must be housed in an extremely small space. Also, the leaf modules are arranged in such manner that they will be located within one irradiation head containing the radiation source and the collimator. During an irradiation treatment, the irradiation head usually has to be moved into various but defined angles with respect to the target volume, e.g. the tumor. Thus, it is desirable to design such a collimator as compact and lightweight as possible. Further, when designing leaf modules for a multi-leaf collimator, the accuracy of the adjusting, i.e. the displacement of the leaf units in the adjusting direction with relation to the leaf drive unit, is a decisive factor both for therapeutic success of the irradiation treatment and for possible adverse effects due to misguided irradiation. Several proposals have been made to employ drive mechanisms operating based on piezoelectric actuation within leaf drive units of leaf modules. General advantages of piezoelectric actuation, in particular by piezoelectric motors, may be identified as the high dynamics, the high positioning accuracy and the low impact of noise within the human acoustic range. Furthermore, piezoelectric motors may exert a clamping force on adjacent parts even when not being in operation. Also, piezoelectric motors do not emit a disruptive magnetic field as conventional electric motors do. This advantage particularly applies to the novel technique of combining imaging by magnetic resonance and radiation therapy with linear acceleration (MR-Linac). In this regard, the strong magnetic field established by the MRT will not disrupt conventional electric motors, as these are replaced by piezoelectric actuation. On the other hand, piezoelectric motors will not build a magnetic field disrupting the MR-imaging and/or the linear acceleration. US 2010/0278310 A1 discloses a multi-leaf collimator with rotatory electromechanical motor and operating method. The electromechanical motor may be a piezo motor. The piezo motor engages leaf units by means of a toothing in order to displace the leaf units. However, indirect transmission of the piezo motor actuation via the proposed toothing involves a backlash concerning the accuracy of displacement of the leaf unit. Furthermore, inevitable slackness within the toothed transmission represents a further major backlash. When the irradiation head is moved in the manner described above, variations in the position of the irradiation head result in changes to the adjusting position of the leaf units, because the weight of the leaf units interacting with the slackness in the toothing within the transmission and/or within the guidance of the leaf unit displacement will result in unwanted deviation of leaf unit adjustment. U.S. Pat. No. 7,242,750 B2 discloses a radiotherapy apparatus comprising a radiation source configured to radiate a radiation ray, a multi-leaf collimator, including a plurality of leaves, configured to limit a radiation range of the radiation ray and a drive unit configured to move at least one of the leaves with an ultrasonic wave. However, this publication does not explicitly address the disadvantageous effects of slackness within the transmission and/or guiding between leaf drive unit and leaf unit as observed in the course of the present invention and as described above. A linear drive and a method for displacing an object by a linear drive, wherein the linear drive comprises at least one piezoelectric actuator, has been proposed by US 2008/0191583 A1. U.S. Pat. No. 6,188,748 B1 discloses a contour collimator of the kind mentioned at the beginning of this specification. The contour collimator according to U.S. Pat. No. 6,188,748 B1 employs a linear drive, wherein a driving transmission is provided between each drive and the associated diaphragm element. The driving transmission has a flexible but tension resistant and pressure resistant power transmitting element one end of which is connected with the diaphragm element and the other end of which is connected with the associated drive. The drive comprises a gear which transmits a rotational movement of the drive to the transmitting element. However, the connection of the transmitting element with the diaphragm element and the associated drive requires coupling elements which increases the costs for manufacturing the collimator. Further, transmission gears are needed for driving the diaphragm elements. When moving the diaphragm elements, a mechanical backlash inevitably occurs. Thus, the dynamic behavior and the motion speed are rather low. Still further, an additional position sensor is required for detecting the exact position of the diaphragm elements as the mechanical friction may cause deviations from the target position. The frictional force sectionally applied by the transducer to the diaphragm element may also result in lateral displacement of the diaphragm element in a direction being oriented perpendicularly related to the adjusting direction. Further linear guiding of the diaphragm element has to be provided, resulting in further frictional force and thus in inaccuracy of adjustment. The invention is therefore based on the objective of designing a contour collimator of the kind mentioned at the beginning in such a way that with a design of the contour collimator being as compact as possible, comprising less mechanical transmission parts and provides a higher position accuracy. The term compact relates to a reduced assembly size of the contour collimator as well as the weight of the contour collimator. This objective is attained in accordance with the invention by the subject-matter disclosed in the independent claim. Preferred embodiments which may be realized in an isolated way or in combination with other preferred embodiments are disclosed subsequently and in the dependent claims. Thus, in a major aspect of the present invention, a contour collimator for radio therapy of the kind mentioned at the beginning is designed in such a way that the drive is a linear motor and each linear motor comprises a linearly movable rod directly fixed to each diaphragm element. The term “directly fixed” as used herein refers to any fixation which does not involve a backlash. Such a direct fixation may be realized by screwing or welding the rod to the diaphragm element. Alternatively, the rod and diaphragm element may be integrally formed, which also represents a direct fixation. The term “linear motor” as used herein refers to an electric motor that has had its stator and rotor unrolled so that instead of producing a torque or rotation, it directly produces a linear force along its length. Particularly, the linear drive of the present invention directly causes a translational movement of the rod which may be disposed with the drive coils. This translational movement is then directly transmitted to the diaphragm element due to the direct fixation of the rod to the diaphragm element. Linear motors may provide high acceleration values. The basic principle is in that the coils, which are arranged circularly in conventional rotative motors, are arranged along a linear distance, wherein the rotor, which rotates in conventional rotative motors, is pulled or attracted along the distance by the linearly moving magnetic field created by the coils. Each linear motor and the associated diaphragm element may be arranged substantially within a common plane. The term “substantially within a common plane” as used herein refers to an orientation which the skilled person would still consider as being within a common plane, i.e. an orientation with a deviation up to an amount of 1.0 mm compared to a strict orientation within a common plane. The linear motors of a group of diaphragm elements may be arranged parallel to each other. The term “parallel” as used herein refers to an orientation which the skilled person would consider as an orientation without any intersection. Each linear motor may comprise a housing, wherein the rod is extendable through the housing. Thus, the rod may be at leas partially disposed within the housing and the rod may be extended from and retracted into the housing. Each linear motor may comprise coils, wherein the rod may comprise magnets, wherein the rod is arranged so as to be linearly movable between the coils. Thus, a high accuracy for the movement control of the rod and the diaphragm element may be achieved as the provision of the rod with the magnets and the orientation of the coils allows an exact movement control. The rod may be welded or screwed to the associated diaphragm element. This is a preferred design of the direct fixation, which clearly prevents any mechanical backlash. Each diaphragm element may comprise a portion extended towards the associated linear motor, wherein the rod may be fixed to the extended portion. This facilitates the direct fixation of the rod to the diaphragm element. The extended portion may comprise an angled portion located at an end facing the linear motor, wherein the rod may be fixed to the angled portion. This design clearly prevents any deviation from the movement or adjusting direction of the diaphragm element. The linear motors of a group of diaphragm elements may be arranged in horizontal and vertical rows, wherein a vertical row of linear motors is associated with a sub-group of diaphragm elements adjacent to each other in a horizontal direction. This design is very compact. The angled portions of a sub-group of diaphragm elements may be arranged at the same position in a horizontal direction. The extended portions of a sub-group of diaphragm elements may be connected to the angled portions at positions shifted in a horizontal direction. The rod may protrude from the linear motor in a direction facing away from the diaphragm element in a retracted position of the rod. The linear drive may be designed to displace the associated diaphragm element in an adjusting direction and may additionally provide guidance to the diaphragm element with respect to any direction being oriented perpendicularly related to the adjusting direction. By combining the tasks to displace the diaphragm element in the adjusting direction and additionally provide aforesaid guidance to the diaphragm element within the linear drive, a collimator with compact design providing very precise and stable adjustability of the diaphragm element may be obtained. The term “guidance” as used herein refers to a technical effect achieved by appropriate means which properly impedes the diaphragm element from deviating from the axis represented by the adjusting direction by a non-tolerable value. The rod may extend substantially, i.e. with a deviation amounting up to ca.±10°, in the adjusting direction of the diaphragm element, i.e. the longitudinal axis of the guiding rod may substantially comply with the adjusting direction. Employing a guiding rod as proposed here yields several benefits. First of all, the diaphragm element which normally comprises heavy and expensive material may be of smaller length, as the guiding rod will provide sufficient range for the displacement of the diaphragm element. Consequently, the entire diaphragm element may be constructed lighter and more compact. Additionally, for the displacing and adjusting of the diaphragm element, a defined interaction of the linear drive with the guiding rod instead of the much larger diaphragm element may be envisaged, yielding a substantial advantage with regards to the precision of adjustment. The rod may comprise ceramic material and/or the guiding rod may be coated with ceramic material. Said ceramic material has been found to exhibit an advantageous interaction with linear drive, as magnets, which interact with coils of the linear drive, may be attached to the rod. The diaphragm element material may comprise a high density material, preferably tungsten. Tungsten has been found to have the capacity to very effectively shield beams from selected areas. A preferred embodiment of the diaphragm element material comprises sintered material comprising approximately 95% tungsten and further components, in particular iron and/or nickel and/or copper. Alternatively, diaphragm element materials with higher or lower fractions of tungsten compared to a preferred percentage of approximately 95% may be employed. Also, alternatively, a non-sintered material or even pure tungsten may be used as diaphragm element material. Generally, any material, in particular high density material, which has the capacity to shield beams, may be employed as diaphragm element material in order to implement the invention. Two groups of diaphragm elements may be provided, wherein each group comprises a plurality of diaphragm elements according to the invention, and wherein the diaphragm elements of each group face each other. Consequently, by adjusting the diaphragm elements facing each other in their respective adjusting direction, an area of arbitrary shape, in particular any contour of a treatment object, for example a tumor, can be recreated for beam application. Particularly, the present invention provides the following advantages. The collimator provides a reduction of the assembly size, weight and mechanical transmission parts. As the collimator uses a linear motor for displacing the diaphragm elements, an extremely high position accuracy is provided without mechanical loose effects. The linear motor allows a simplified control loop algorithm with no need to compensate looses and transmission. Further, the linear motor provides a reduction of dead time. The collimator achieves dynamic movements of the diaphragm elements up to 4 m/s in the patient plane. The direct fixation of the rod to the associated diaphragm element represents a simplified mechanical assembly and easier serviceability. Further, the direct fixation of the rod to the associated diaphragm element provides a reduction of friction loose and allows to omit transmission parts such as transmission gears. Still further, the direct fixation of the rod to the associated diaphragm element prevents mechanical backlash. A position sensor is already implemented in the linear drive such that an additional position sensor for primary position detection is not necessary. Particularly, the principle of the linear motor already implements an encoder system for the position detection. The collimator also allows the omission of a dead time compensation for control loop when driving sense changes. Contrary to other drive principle known in the prior art, no transmission of a rotary movement into a linear movement is required. The collimator also provides a simplified approach for establishing new treatment methods for tracking and gating, i.e. a compensation of tumor movement in realtime. More particularly, the collimator allows the establishment of new dynamic treatment requirements and a direct and dynamic control of movements to follow tumor movement without compensation of mechanical backlash. Still further, the linear drive provides independence of gravitation influences. Furthermore, the collimator provides a reduction of effort for fine adjustment and calibration needs for a sensor such as the position sensor. The linear drive used in the present invention allows a reduced noise in comparison to standard DC drives. The collimator also provides a high force/weight ratio. Summarizing, the following embodiments are preferred embodiments of the present invention: A contour collimator for radiotherapy, comprising a plurality of plate-shaped diaphragm elements provided in a guiding block and movably arranged with respect to one another to form a contour diaphragm for a radiation beam emitted by a radiation source towards the collimator, and at least one drive for moving the diaphragm elements, wherein a drive of its own is associated with each diaphragm element, the drives of a group of diaphragm elements are arranged substantially adjacent to one another, and the drive is a linear motor, characterized in that each linear motor comprises a linearly movable rod directly fixed to each diaphragm element. The contour collimator according to the preceding embodiment, characterized in that each linear motor and the associated diaphragm element are arranged substantially within a common plane. The contour collimator according to any preceding embodiment, characterized in that the linear motors of a group of diaphragm elements are arranged parallel to each other. The contour collimator according to any preceding embodiment, characterized in that each linear motor comprises a housing, wherein the rod is extendable through the housing. The contour collimator according to the preceding embodiment, characterized in that each linear motor comprises coils, wherein the rod comprises magnets, wherein the rod is arranged so as to be linearly movable between the coils. The contour collimator according to any preceding embodiment, characterized in that the rod is welded or screwed to the associated diaphragm element. The contour collimator according to any preceding embodiment, characterized in that each diaphragm element comprises a portion extended towards the associated linear motor, wherein the rod is fixed to the extended portion. The contour collimator according to the preceding embodiment, characterized in that the extended portion comprises an angled portion located at an end facing the linear motor, wherein the rod is fixed to the angled portion. The contour collimator according to the preceding embodiment, characterized in that the linear motors of a group of diaphragm elements are arranged in horizontal and vertical rows, wherein a vertical row of linear motors is associated with a sub-group of diaphragm elements adjacent to each other in a horizontal direction. The contour collimator according to the preceding embodiment, characterized in that the angled portions of a sub-group of diaphragm elements are arranged at the same position in a horizontal direction. The contour collimator according to the preceding embodiment, characterized in that extended portions of a sub-group of diaphragm elements are connected to the angled portions at positions shifted in a horizontal direction. The contour collimator according to any preceding embodiment, characterized in that the rod protrudes from the linear motor in a direction facing away from the diaphragm element in a retracted position of the rod. FIG. 1 is perspective view of a contour collimator 10 according to the present invention, whose core is formed by a guiding block 12. The guiding block 12 comprises a lower guide plate 14, an upper guide plate 16 as well as two side walls 18 and 20. The side walls 18, 20 connect the lower guide plate 14 and the upper guide plate 16 to one another. Thus, the guiding block 12 substantially comprises the shape of a rectangular frame. Needless to say, the collimator 10 may comprise further guiding blocks 12 or the guiding block 12 may be divided in several rectangular frame elements. The guiding block 12 may be covered at the upper side and the lower side by cover plates which respectively comprise an opening, which may be rectangular and provided centrally in the cover plates. The top side of the lower guide plate 16 is provided with a plurality of lower guide grooves extending in a longitudinal direction of the guiding block 12 and formed parallel to one another at equal lateral distance. The upper and lower guide plates 14, 16 are made preferably of brass, bronze or ceramics or another radiation-resistant material having good sliding properties. In the same way, upper guide grooves are provided on the bottom side of the upper guide plate 16. Since the guiding block 12 is made symmetrically with respect to a middle plane 22 extending perpendicular with respect to the longitudinal direction of guiding block 12 and through the center of openings in the upper and lower cover plates, only the design of the guiding block 12 on one side with respect to the middle plane 22 is described for the purpose of simplification. The design on the other side is formed analogously thereto. Further, the contour collimator 10 comprises a plurality of plate-shaped diaphragm elements 24 which are provided in the guiding block 12 and movably arranged with respect to one another to form a contour diaphragm for a radiation beam emitted by a radiation source (not shown) towards the contour collimator 10. For this purpose, each of the diaphragm elements 24 is inserted in movably translatory fashion in each pairing of the vertically superposed guide grooves. The width of the individual guide grooves corresponds to about half the thickness of a diaphragm element 24. The thickness of a plate-shaped diaphragm element 24 may be about 1 mm. The diaphragm elements 24 inserted in the guide grooves are slightly spaced laterally, so that they do not come into contact. In order to allow the shielding of beams emitted by the radiation source, the diaphragm elements 24 comprise tungsten material. For example, the collimator 10 may comprise 40 diaphragm elements 24 on each side of the middle plane 22. Particularly, the diaphragm elements 24 on each side of the middle plane 22 respectively form a group 26, 28 of diaphragm elements 24. FIG. 2 shows a front view of the contour collimator 10. Further, the contour collimator 10 comprises at least one drive 30 for moving the diaphragm elements 24. A drive 30 of its own is associated with each diaphragm element 24. The drives 30 of a group 26, 28 of diaphragm elements 24 are arranged substantially adjacent to one another. The drive 30 is a linear motor. More particularly, each linear motor 30 and the associated diaphragm element 24 are arranged substantially within a common plane. The linear motor may be a linear motor of the series LM1247 or LM2070 available from the company Dr. Fritz Faulhaber GmbH & Co. KG, Daimlerstraβe 23/25, 71101 Schönaich, Germany. FIG. 3 shows a left side view of the contour collimator 10. As is shown, the linear motors 30 of a group 26, 28 of diaphragm elements 24 are arranged parallel to each other. More particularly, the linear motors 30 of a group 26, 28 of diaphragm elements 24 are arranged in horizontal and vertical rows 32, 34, wherein a vertical row 34 of linear motors 30 is associated with a sub-group 36 of diaphragm elements 24 adjacent to each other in a horizontal direction. The sub-group 36 of diaphragm elements 24 comprises 8 diaphragm elements 24 arranged directly adjacent to one another in the horizontal direction. Accordingly, on each side of the middle plane 22, there are 5 sub-groups 36 of diaphragm elements 24 The linear motor 30 comprises a housing 38. The linear motor 30 further comprises coils (not shown). The coils are disposed within the housing 38. Each linear motor 30 also comprises a linear movable rod 40 directly fixed to the associated diaphragm element 24 as will be explained in more detail below. The rod 40 is extendable through the housing 38. For example, the rod 40 is welded or screwed to the associated diaphragm element 24. More particularly, each diaphragm element 24 comprises a portion 42 extended towards the associated linear motor 30. The rod 40 is fixed to the extended portion 42. For this purpose, the extended portion 42 comprises an angled portion 44 located at an end 46 facing the linear motor 30, wherein the rod 40 is fixed to the angled portion 44. The angled portions 44 of a sub-group 36 of diaphragm elements 24 are arranged at the same position in a horizontal direction. The extended portions 42 of a sub-group 36 of diaphragm elements 24 are connected to the angled portions 44 at positions shifted in a horizontal direction. The rod 40 protrudes from the linear motor 30 in a direction facing away from the diaphragm element 24 in a retracted position of the rod 40 as will be explained in more detail below. The rod 40 comprises magnets (not shown). The magnets may be attached to the rod 40 on an outer surface thereof or may alternatively be disposed within the rod 40. The rod 40 is arranged so as to be linearly movable between the coils in the housing 38. Within the guiding block 12, position sensors (not shown) may be arranged for detecting the position of the diaphragm elements 24. A position sensor of its own may be associated with each diaphragm element 24. Each position sensor may be disposed so as to face the extended portion 42 of the associated diaphragm element 24. The position sensors are adapted to determine the position of the diaphragm elements 24 by measuring a magnetic field. For example, the position sensors may be magnetic sensors, in particular Hall effect sensors. In this case, such a position sensor comprises a magnetic element adapted to provide a magnetic field, and a detector device capable of detecting said magnetic field, wherein the diaphragm element 24 comprises one of the magnetic element and the detector device, and the other of the magnetic element and the detector device is attached to a supporting device such as a part of the housing 38 relative to which the diaphragm element 24 is movable. In particular, the diaphragm element 24 may comprise at least part of the measurement device. Preferably, the part of the measurement device may be integrated into the diaphragm element 24, in particular milled into the diaphragm element 24. This allows to provide the diaphragm element 24 with the measurement device with minimal impact on the operability of the collimator. Alternatively, the part of the measurement device may be attached to the diaphragm element 24, in particular mounted onto a surface of the diaphragm element 24 or glued to the diaphragm element 24 such as the extended portion 42. Since the magnetic sensor is small, it may be attached to the outer surface of the diaphragm element 24 without affecting the operability of the collimator system. Further, magnetic sensors also allow for contact-free measurements, and therefore measurement results are not affected by friction or abrasive wear. Alternatively, instead of an analog Hall effect sensor, sine/cosine type Hall effect sensors may be used with the present invention. Such position sensors match user's positioning needs without requiring an external encoder. Additional sensors such as potentiometers may be equipped within the design. Examples for apparatus and methods for detecting the position of the diaphragm elements 24 are described in WO2012/123093 A1, the contents of which relating to methods and apparatus for position sensoring such as magnetic encoders are incorporated herein by reference. Further, adjacent the linear motors 30, boards 48 may be disposed for supporting circuit boards and other electrical connections (not shown) necessary for activating the linear motors 30. At the outer side of the guiding block 12, control means 50 such as electronic control units for controlling the linear motors 30 are provided. FIG. 4 shows a top view of the contour collimator 10. Particularly, this drawing is very explicit in showing that within contour collimator 10, adjacent diaphragm elements 24 may be stacked in such a way that a shielding of beams from a selected area to form a contour diaphragm can be implemented. FIG. 5 shows a perspective view of a left portion of the contour collimator 10. As mentioned above, the angled portions 44 of a sub-group 36 of diaphragm elements 24 are arranged at the same position in a horizontal direction. The extended portions 42 of a subgroup 36 of diaphragm elements 24 are connected to the angled portions 44 at positions shifted in a horizontal direction. FIG. 5 exemplarily shows that the extended portion 42 of the uppermost diaphragm element 24 of a sub-group 36 is connected to the opposite end of the angled portion 44 than the extended portion 42 of the lowermost diaphragm element 24 of the same sub-group 36. In the example of FIG. 5, the extended portion 42 of the uppermost diaphragm element 24 of a sub-group 36 is connected to the left end 52 of the angled portion 44 if seen from the left to the right in FIG. 5, whereas the extended portion 42 of the lowermost diaphragm element 24 of the same sub-group 36 is connected to the right end 54 of the angled portion 44. The extended portions 42 of diaphragm elements 24 of the same sub-group 36 are connected to the angled portions 44 at intermediate positions between the left end and right end shifted in the horizontal direction. This shifting in the positions of the extended portions 42 in the horizontal direction allows to arrange the diaphragm elements 24 in a very compact manner as the angled portions 44 do not obstruct each other. FIG. 5 further shows that the linear motors 30 may be individually activated so as to displace the associated diaphragm element 24 in order to form the contour diaphragm. FIG. 6 shows a front view of the left portion of the contour collimator 10. Particularly, this drawing is very explicit in showing that within contour collimator 10, the rods 40 protrude from the linear motors 30 in a direction facing away from the diaphragm elements 24 in a retracted position of the rods 40. FIG. 7 shows a left side view of the left portion of the contour collimator 10. Particularly, this drawing is very explicit in showing that within contour collimator 10, the linear motors 30 are arranged in horizontal rows 32 and vertical rows 34. In the example shown in FIG. 7, there are 8 horizontal rows 32 and 5 vertical rows 34. FIG. 8 shows a top view of the left portion of the contour collimator 10. Particularly, this drawing is very explicit in showing that within contour collimator 10, the linear motors 30 may be individually activated so as to displace the associated diaphragm element 24 in order to form the contour diaphragm. Further, FIG. 8 shows the arrangement of the diaphragm elements 24 in a very compact manner as the angled portions 44 do not obstruct each other. FIG. 9 shows a perspective view of a sub-group 36 of diaphragm elements 24 to which rods 40 of the linear motors 30 are directly fixed. As explained with reference to FIG. 5, the angled portions 44 of the sub-group 36 of diaphragm elements 24 are arranged at the same position in a horizontal direction. The extended portions 42 of a sub-group 36 of diaphragm elements 24 are connected to the angled portions 44 at positions shifted in a horizontal direction. FIG. 5 exemplarily shows that the extended portion 42 of the uppermost diaphragm element 24 of a sub-group 36 is connected to the opposite end of the angled portion 44 than the extended portion 42 of the lowermost diaphragm element 24 of the same sub-group 36. FIG. 9 also shows that the extended portions 42 of the sub-group 36 of diaphragm elements 24 are arranged at the shifted positions in a vertical direction. In the example shown in FIG. 9, the rod 40 of uppermost linear motor 30 is connected to the uppermost extended portion 42 of the left diaphragm element 24 of the sub-group 36 and the rod 40 of lowermost linear motor 30 is connected to lowermost extended portion 42 of the right diaphragm element 24 of the sub-group 36 if seen from the left to the right in FIG. 9. FIG. 10 shows a front view of the sub-group 36 of diaphragm elements 24. Particularly, this drawing is very explicit in showing that within contour collimator 10, the rods 40 protrude from the linear motors 30 in a direction facing away from the diaphragm elements 24 in a retracted position of the rods 40. FIG. 11 shows a left side view of the sub-group 36 of diaphragm elements 24. Particularly, this drawing is very explicit in showing that within contour collimator 10, the linear motors 30 of the sub-group 36 are arranged in a vertical row 34. In the example shown in FIG. 11, there are 8 linear motors 30 in the vertical row 34. FIG. 12 shows a top view of the sub-group 36 of diaphragm elements 24. Particularly, this drawing is very explicit in showing that within contour collimator 10, the linear motors 30 may be arranged in a compact manner as the linear motors 30 of the sub-group 36 may be arranged in a vertical row 34. Further, FIG. 12 shows the arrangement of the diaphragm elements 24 in a very compact manner as the angled portions 44 do not obstruct each other. As is shown in FIG. 12, the sub-group 36 comprises 8 diaphragm elements 24 arranged parallel and adjacent to one another in the horizontal direction. As mentioned above, in order to displace a diaphragm element 24 in an adjusting direction for forming the contour diaphragm, the linear motor 30 associated with the diaphragm element 24. Particularly, the linear motor 24 is supplied with electric current or electric voltage. The electric current or electric voltage creates a magnetic field around the coil within the housing 38. The magnetic field causes the rod 40 of the linear motor 30 to extend or retract through the housing 38 depending on the direction of the magnetic field. This extension or retraction of the rod 40 causes the diaphragm element 24 to which the rod 40 is directly fixed to extend or retract. By individually activating the linear motors 30, the individual diaphragm elements 24 are displaced in the adjusting direction to form the contour diaphragm corresponding to the treatment object to be irradiated. The exact position of the diaphragm element 24 is detected by the position sensor. As the rod 40 of the linear motor 30 is directly fixed to the diaphragm element 24, an additional position sensor, i.e. a secondary position sensor, for compensating a backlash as with conventional electric drives is not necessary as the direct fixation prevents any backlash. 10 contour collimator 12 guiding block 14 lower guide plate 16 upper guide plate 18 side wall 20 side wall 22 middle plane 24 diaphragm elements 26 group of diaphragm elements 28 group of diaphragm elements 30 drive/linear motor 32 horizontal row 34 vertical row 36 sub-group 38 housing 40 rod 42 extended portion 44 angled portion 46 end 48 board 50 control means 52 left end 54 right end |
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summary | ||
abstract | A nuclear reactor cooling system with passive cooling capabilities operable during a reactor shutdown event without available electric power. In one embodiment, the system includes a reactor vessel with nuclear fuel core and a steam generator fluidly coupled thereto. Primary coolant circulates in a flow loop between the reactor vessel and steam generator to heat secondary coolant in the steam generator producing steam. The steam flows to a heat exchanger containing an inventory of cooling water in which a submerged tube bundle is immersed. The steam is condensed in the heat exchanger and returned to the steam generator forming a closed flow loop in which the secondary coolant flow is driven by natural gravity via changes in density from the heating and cooling cycles. In other embodiments, the cooling system is configured to extract and cool the primary coolant directly using the submerged tube bundle heat exchanger. |
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055027547 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The core plate wedge assembly in accordance with the invention may be used in conjunction with a shroud restraint assembly of the type shown in FIG. 2. This shroud restraint assembly comprises a tie rod 54 having a circular cross section. A lower end of tie rod 54 is anchored in a threaded bore formed in the end of a spring arm 56a of a lower spring 56. Tie rod 54 extends from the end of spring arm 56a to a position adjacent the outer circumferential surface of the top guide support ring 18c. The upper end of tie rod 54 has a threaded portion. The lower spring 56 is anchored to a gusset plate 58 attached to the shroud support plate 52. The gusset is part of the original construction in some reactors, or is otherwise bolted in place as part of the repair. The bottom spring has a slotted end (see FIG. 9) which fits over the gusset plate 58 and forms a pair of clevis hooks 56c. The clevis hooks 56c hook under opposite ends of a clevis pin 60 inserted through a hole machined in the gusset plate 58. Engagement of the slotted end with the gusset plate 58 maintains alignment of lower spring 56 under the action of seismic motion of the shroud, which may be oblique to the spring's radial orientation. The tie rod 54 is supported at its top end by an upper support assembly which hangs from the top edge of the shroud. In accordance with the installation method of the present invention, a pair of notches or slots are machined in the shroud head ring 28a of shroud head 28. The notches are positioned in alignment with a pair of bolted upper support plate segments 62 of the upper support assembly when the shroud head 28 is properly seated on the top surface of shroud flange 18a. These notches facilitate the coupling of the tie rod assembly to the shroud flange in accordance with the invention, as described in detail hereinafter. The preferred machining technique is electrical discharge machining, although any other suitable machining technique can be used. Machining of these notches may be performed with the shroud head removed from the reactor, avoiding delay of in-reactor outage work. The pair of notches at each tie rod azimuthal position receive respective hook portions 62a of the upper support plates 62. Each hook 62a conforms to the shape of the top surface of shroud flange 18a and the shape of the steam dam 29. The distal end of hook 62a hooks on the inner circumference of shroud dam 29. The upper support plates 62 are connected in parallel by a top support bracket (not shown) and a support block 66 which forms the anchor point for the top of the tie rod. Support block 66 has an unthreaded bore, tapered at both ends, which receives the upper end of tie rod 54. After the upper end of tie rod 54 has been passed through the support block 66, a threaded tensioning nut 70 is screwed onto the threaded portion of tie rod 54. The assembly comprised of support plates 62 with hooks 62a, support block 66, tie rod 54, lower spring 56, clevis pin 60 and gusset plate 58 form a vertical load path by which the shroud flange 18a can be anchored to the shroud support plate 52. In the tensioned state, upper support plates 62 exert a restraining force on the top surface of the shroud flange 18a which opposes separation of the shroud at any assumed failed circumferential weld location. The upper restraint spring 72 is a double cantilever "wishbone" design, to react the lateral seismic loads without adding bending load on the top support. The spring 72 is rotatable relative to the upper support assembly. The end of the outer arm of spring 72 has an upper contact spacer 74 mounted thereon. Upper contact spacer 74 is designed to bear against the inner surface of the reactor pressure vessel wall. The upper spring assembly is installed with enough elastic preload to prevent mechanical wear of its parts due to reactor vibration. In the event of seismic loading at some oblique angle, the spring 72 can rotate on its axle mounting to absorb the azimuthal motion component, without transmitting oblique loading into the support block 66. Cantilever torsion arms on each side of the upper spring 72 restore the rotational alignment of the spring after seismic deflection. The upper contact spacer 74 which bears against the vessel 10 reacts the restraint load from the vessel and pivots to follow the spring rotation. The spring arm 56a of lower spring 56 laterally supports the shroud 18 at the core plate 18e, against the vessel 10, via a lower contact spacer 76. The lower spring assembly is installed with a controlled preload, obtained by machining lower contact spacer 76 to match the measured assembly fitup. The top end of spring arm 56a has a threaded bore to provide the attachment for the bottom of the tie rod 54. The member 56d connecting the upper wishbone spring 56a, 56b to the clevis hook 56c is offset from the line of action between the lower end of tie rod 54 and the clevis pin 60. Axial loads in the tie rod therefore cause bending of the lower connecting member 56d and associated pivoting of the clevis hook 56c about the clevis pin 60. The specific configuration is designed to add the desired axial flexibility to the assembly to minimize thermal expansion loads. This flexibility is adjusted along with that of the upper and lower lateral springs to tune the dynamic frequency response of the reactor internal structure to minimize lateral seismic loads. The lateral seismic loads from the main mass of the reactor core are reacted (for a cracked shroud) by the upper and lower springs at the top guide 18c and core plate 18e, respectively. Gross motion limit stops are also mounted on the stabilizers to limit lateral displacement of other segments of the shroud due to circumferential weld failures. Each upper support block 66 has a limit stop which blocks gross lateral deflection of the middle shroud wall 18d relative to the top guide support ring 18c in the event that the welds between top guide support ring 18c and middle shroud wall 18d become severed. If left unchecked, gross lateral deflection of middle shroud wall 18d could damage peripheral fuel assemblies in the fuel core. A middle support 78 may be used to provide a limit stop, as shown in FIG. 4, if the middle shroud wall 18d has a circumferential weld in its middle. To facilitate mounting of the middle support 78, a mid-support ring 80 is secured to the tie rod 54. The middle support 78 has a section of an annular recess which form fits on collar 80, thereby preventing lateral shifting of middle support 78 relative to tie rod 54. The middle support 78 is preloaded against the vessel wall at assembly by radial interference which bends the tie rod 54. Thus it provides both a limit stop for the middle shroud wall 18d, and a mid-span support for the tie rod, improving its resistance to vibratory excitation failure. Further, each lower spring 56 has a limit stop 82 which blocks gross lateral deflection of the lower shroud wall 18f relative to the core plate support ring 18e in the event that the welds between core plate support ring 18e and lower shroud wall 18f become severed. Gross lateral deflection of lower shroud wall 18f and shroud support 51 welded thereto could, if not checked, cause damage to the control rod guide tubes located underneath the core. Lateral displacement is limited by the radial clearance between arm 56a of lower spring 56 and limit stop 82 mounted on arm 56b of lower spring 56. The gusset 58 limits displacement of the shroud support 50. The present invention is an apparatus for restraining a core plate, and the fuel assemblies seated thereon, against lateral deflection relative to the core shroud. This repair is preferably used in conjunction with the above-described repairs which restrain the core shroud against lateral and vertical deflection relative to the reactor pressure vessel. In accordance with the preferred embodiment of the invention depicted in FIG. 3, the core plate 21 is restrained against lateral deflection relative to the middle shroud wall 18d by a core plate wedge assembly 84. The method involves the placement of a plurality of core plate wedge assemblies (e.g., four) in the gap between the circular outer peripheral edge of the core plate and the circular cylindrical inner surface of the middle shroud wall 18d at respective azimuthal positions. These core plate wedge assemblies are wedged into place to maintain the spacing between the core plate and the shroud, thereby maintaining the alignment of the fuel assemblies. The core plate wedge assemblies are preferably located in azimuthal alignment with the shroud restraint assembly. Thus, the core plate wedge assembly in conjunction with the corresponding lateral shroud restraint assembly form a direct path for transmission to the reactor pressure vessel of a load exerted laterally by the fuel core. Referring to FIGS. 4A and 4B, the core plate wedge assembly 84 comprises a core plate wedge 86 and a core plate wedge clip 88. The core plate wedge 86 has a planar mating surface 86a and another planar surface 86c parallel thereto. The core plate wedge clip 88 has a planar mating surface 88a and another planar surface 88d which is at an oblique angle (preferably 6.degree.) relative to surface 88a. In the untrimmed state, the surface 86d of core plate wedge 86 is parallel to the 86a. Prior to assembly of core plate wedge 86 and core plate wedge clip 88, the gap G between the outer circumferential surface of core plate 21 and the inner circumferential surface of the middle shroud wall 18d (see FIG. 3) is measured. Then the core plate wedge 86 is trimmed as shown in FIG. 4A to form a new surface 86d which, when core plate wedge 86 is coupled to core plate wedge clip 88 in its final wedged position, will be parallel to surface 88d and separated therefrom by a distance which increases to approximately G as the bolt is tightened and parts 86 and 88 slide against each other. To assemble the core plate wedge assembly 84, the mating surface 88a of core plate wedge clip 88 is placed flush against the mating surface 86a of core plate wedge 86 and then a wedge bolt 90 is used to couple the core plate wedge 86 and the core plate wedge clip 88. The core plate wedge 86 has a first unthreaded circular cylindrical bore of first diameter for receiving the circular cylindrical head of wedge bolt 90 and a second unthreaded circular cylindrical bore of second diameter for receiving the threaded shaft 90a of wedge bolt 90. The bore of first diameter communicates with the bore of second diameter at a shoulder 86b. Longitudinal displacement of wedge bolt 90 relative to core plate wedge 86 is prevented by shoulder 86b, while allowing wedge bolt 90 to rotate freely relative to core plate wedge 86. The core plate wedge clip 88 has a threaded bore 88b which threadably engages the threaded shaft 90a as the wedge bolt 90 is screwed in. As the wedge bolt 90 is rotated in the direction of tightening, the core plate wedge clip 88 slides relative to core plate wedge 86 along the bolt axis. To facilitate mutual sliding of mating surfaces 86a and 88a relative to each other, a lubricant may be applied to one or both mating surfaces prior to assembly. In the initial configuration of the core plate wedge assembly 84, the core plate wedge clip 88 engages only a small portion of the threaded shaft 90a of wedge bolt 90 and occupies an initial axial position relative to core plate wedge 86. In this initial axial position, the distance separating surfaces 86d and 88d is less than the gap G by an amount sufficient to allow the latching projection 88c on core plate wedge clip 88 to pass through the gap. The core plate wedge assembly 84 is then held at an elevation such that when the wedge bolt 90 is tightened, the latching projection 88c will hook underneath the core plate 21, as seen in FIG. 3, as the core plate wedge clip 88 travels axially toward the bolt head. At the same time, the distance separating surfaces 86d and 88d increases. When the distance separating surfaces 86d and 88d equals the gap G, the surface 86d contacts the middle shroud wall 18 and the surface 88d contacts the core plate 21. The wedge bolt 90 can then be tightened until the desired preload is attained. Once the core plate wedge 86 has been trimmed, the amount of preload is a function of the distance which core plate wedge clip 88 travels relative to the core plate wedge. In this state the core plate wedge assembly 84 maintains the spacing G between the core plate and the shroud and transmits a load from the core plate to the shroud. This mounting allows simple installation and subsequent removal, if required for reactor servicing access. When the desired amount of preload has been attained, the wedge bolt 90 is locked against further rotation relative to the core plate wedge 86 by engagement of a pair of wishbone spring latches 92. As best seen in FIG. 4B, each spring latch 92 has a short leg with a projection 92a that interlocks with the core plate wedge 86 and a long leg with a key 92b that interlocks with one of a multiplicity of longitudinal slots 90b (see FIGS. 4A and 5) formed on the outer circumference of the head of wedge bolt 90. The tip of the long leg of spring latch 92 has a surface 92c which is oblique relative to the leg axis. This oblique surface is contacted by a portion of a tool (not shown) which is used to screw and unscrew the wedge bolt 80. The tool surface bears against the oblique surface, thereby camming the key on the long leg to a position where it will not interfere with the head of the wedge bolt during wedge bolt rotation. As seen in FIG. 5, the head of wedge bolt 90 has an axial recess 90c of hexagonal cross section for receiving a form-fitting portion of the tool to rotate the bolt and transmit preload torque. Only the insertion of the torquing tool is required to disengage the locking latch so that wedge bolt 90 is free to turn. In addition, the short leg of the spring latch 92 has a projection 92d which can be gripped by a remotely manipulated tool in the event that spring latch 92 must be removed. To remove the spring latch, the short leg must be displaced toward the long leg by an amount sufficient to enable projection 92a to clear the interfering portion 86e of the core plate wedge 86. Two views of the spring latch are presented in FIGS. 6A and 6B. In accordance with the preferred embodiment, the core plate wedge and core plate wedge clip are made of austenitic stainless steel (e.g., Type 316). The wedge bolt and associated spring latch are made of Ni-Cr-Fe alloy X-750. Both are specified and fabricated with controls to assure maximum corrosion resistance in the BWR environment. In accordance with the preferred arrangement, four core plate wedge assemblies are installed at respective azimuthal positions distributed at angular intervals around the core plate circumference. However, the concept of the invention is directed to the installation of three or more core plate wedge assemblies and is expressly not limited to an arrangement of four. The preferred embodiment of the core plate wedge assembly in accordance with the invention has been disclosed for the purpose of illustration. Variations and modifications of the disclosed structure which fall within the concept of this invention will be readily apparent to persons skilled in the art of mechanical engineering in the boiling water reactor environment. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter. |
claims | 1. A method of fabricating a nuclear fuel comprising:a) providing a fissile material;b) providing a plurality of hollow microballoons each microballoon comprising carbon;c) providing a phenolic resin; andd) combining said fissile material, said phenolic resin, and said microballoons; ande) heating said combination sufficiently to form at least some fissile material carbides creating a nuclear fuel particle comprising fission product collection spaces formed from the hollow core of each said microballoon. 2. The method of fabricating a nuclear fuel of claim 1 whereby:a) said step of combining said fissile material, said phenolic resin, and said microballoons comprises:i. forming said fissile material into microspheres;ii. heating said fissile material microspheres sufficiently to form at least some fissile material carbides of said microspheres; andiii. overcoating said fissile material microspheres with a microballoon mixture of said phenolic resin, and said microballoons. 3. The method of fabricating a nuclear fuel of claim 1 whereby:a) said step of combining said fissile material, said phenolic resin, and said microballoons comprises:i. spheronizing said fissile material, said phenolic resin, and said microballoons into a spheroidal particle. 4. The method of fabricating a nuclear fuel of claim 1 whereby:a) said step of combining said fissile material, said phenolic resin and said microballoons comprises:i. overcoating said microballoons with a fissile material mixture of said phenolic resin and said fissile material. 5. The method of fabricating a nuclear fuel of claim 1 further comprising:a) mixing said nuclear fuel particle with a metal creating a nuclear fuel particle with a metal matrix; andb) compacting and extruding a plurality of said nuclear fuel particle with a metal matrix forming one or more fuel rods. 6. The method of fabricating a nuclear fuel of claim 1 further comprising:a) overcoating said nuclear fuel particle with a metal mixture of a metal and a phenolic resin creating a nuclear fuel particle with a metal matrix precursor;b) heating said nuclear fuel particle with a metal matrix precursor sufficiently to cure and carbonize said phenolic resin and form metal carbide creating a nuclear fuel carbide particle with a metal matrix carbide shell; andc) compacting and extruding said nuclear fuel carbide particle with a metal matrix carbide shell forming one or more fuel rods. 7. The method of fabricating a nuclear fuel of claim 6 further comprising:a) heating said one or more fuel rods sufficiently to react metal in said nuclear fuel particle in a matrix with the fissile material in said nuclear fuel particle with a metal matrix forming at least some fissile material carbides creating a nuclear carbide fuel with a metal matrix. 8. The method of fabricating a nuclear fuel of claim 6 further comprising:a) said metal selected from the group of metal consisting of: Zr, Ti, Nb and combinations thereof. 9. The method of fabricating a nuclear fuel of claim 1 further comprising:a) said hollow microballoons comprising a representative diameter between 1 and 60 microns. 10. The method of fabricating a nuclear fuel of claim 1 further comprising:a) said hollow microballoons comprising a representative diameter between 1 and 10 microns. 11. The method of fabricating a nuclear fuel of claim 1 further comprising:a) said fissile material comprising a representative diameter between 2 to 500 microns. 12. The method of fabricating a nuclear fuel of claim 1 whereby:a) said step of combining said fissile material, said phenolic resin, and said microballoons comprises:i. forming said fissile material into microspheres;ii. heating said fissile material microspheres sufficiently to form at least some fissile material carbides of said microspheres; andiii. overcoating said heated fissile material microspheres with a microballoon mixture of said phenolic resin, and said microballoons. 13. The method of fabricating a nuclear fuel of claim 12 further comprising:a) overcoating said nuclear fuel particle with a metal mixture of a metal and a phenolic resin creating a nuclear fuel particle with a metal matrix precursor;b) heating said nuclear fuel particle with a metal matrix precursor sufficiently to cure and carbonize said phenolic resin and form metal carbide creating a nuclear fuel carbide particle with a metal matrix carbide shell;c) compacting and extruding one or more of said nuclear fuel particle with a metal matrix forming one or more fuel rods at a temperature at least approximately 1,100° C. and less than approximately 1,400° C.;d) cooling said one or more fuel rods;e) heating said cooled one or more fuel rods;f) said metal selected from the group of metal consisting of: Zr, Ti, Nb or a combination thereof;g) said hollow microballoons comprising a representative diameter between 1 and 10 microns;h) said fissile material comprising a representative diameter between 2 to 500 microns; andi) each said one or more fission product collection space at least at a partial vacuum. 14. The method of fabricating a nuclear fuel of claim 13 further comprising:a) mixing said nuclear fuel particle with a metal before said step of compacting and extruding said nuclear fuel particle. 15. The method of fabricating a nuclear fuel of claim 1 whereby:a) said step of combining said fissile material, said phenolic resin, and said microballoons comprises:i. spheronizing said fissile material, said phenolic resin, and said microballoons into a spheroidal particle. 16. The method of fabricating a nuclear fuel of claim 15 further comprising:a) overcoating said nuclear fuel particle with a metal mixture of a metal and a phenolic resin creating a nuclear fuel particle with a metal matrix precursor;b) heating said nuclear fuel particle with a metal matrix precursor sufficiently to cure and carbonize said phenolic resin and form metal carbide creating a nuclear fuel carbide particle with a metal matrix carbide shell;c) compacting and extruding one or more of said nuclear fuel particle with a metal matrix forming one or more fuel rods at a temperature at least approximately 1,100° C. and less than approximately 1,400° C.;d) cooling said one or more fuel rods;e) heating said cooled one or more fuel rods;f) said metal selected from the group of metal consisting of: Zr, Ti, Nb or a combination thereof;g) said hollow microballoons comprising a representative diameter between 1 and 10 microns;h) said fissile material comprising a representative diameter between 2 to 500 microns; andi) each said one or more fission product collection space at least at a partial vacuum. 17. The method of fabricating a nuclear fuel of claim 16 further comprising:a) mixing said nuclear fuel particle with a metal before said step of compacting and extruding said nuclear fuel particle. 18. The method of fabricating a nuclear fuel of claim 1 whereby:a) said step of combining said fissile material, said phenolic resin and said microballoons comprises:i. overcoating said microballoons with a fissile material mixture of said phenolic resin and said fissile material. 19. The method of fabricating a nuclear fuel of claim 18 further comprising:a) overcoating said nuclear fuel particle with a metal mixture of a metal and a phenolic resin creating a nuclear fuel particle with a metal matrix precursor;b) heating said nuclear fuel particle with a metal matrix precursor sufficiently to cure and carbonize said phenolic resin and form metal carbide creating a nuclear fuel carbide particle with a metal matrix carbide shell;c) compacting and extruding one or more of said nuclear fuel particle with a metal matrix forming one or more fuel rods at a temperature at least approximately 1,100° C. and less than approximately 1,400° C.;d) cooling said one or more fuel rods;e) heating said cooled one or more fuel rods;f) said metal selected from the group of metal consisting of: Zr, Ti, Nb or a combination thereof;g) said hollow microballoons comprising a representative diameter between 1 and 10 microns;h) said fissile material comprising a representative diameter between 2 to 500 microns; andi) each said one or more fission product collection space at least at a partial vacuum. 20. The method of fabricating a nuclear fuel of claim 19 further comprising:a) mixing said nuclear fuel particle with a metal before said step of compacting and extruding said nuclear fuel particle. |
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063339617 | abstract | Reflection masks are disclosed for use in microlithography (especially soft-X-ray microlithography) in which the linewidth of a circuit pattern defined by such a mask is transferred onto a wafer with reduced adverse effects from reflectance non-uniformities of the mask. A representative reflection mask includes a multilayer mirror that reflects soft X-rays of a prescribed wavelength and a soft-X-ray-absorber layer defining pattern elements. The multilayer mirror is configured to have a thickness period that varies through the thickness dimension of the multilayer mirror. |
abstract | A system for non-contact cleaning of particulate contamination of surfaces includes one or more sources that create a charge imbalance between a surface and particles that contaminate the surface, and a power supply that creates a pulsed electrical bias on the surface. This imbalance produces an electrostatic force that propels the particles off the surface. The cleaning process can be associated, for example, with microelectronic lithography and manufacturing. |
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claims | 1. A system for electron pattern imaging comprising:a device for converting electron patterns into visible light provided to receive an electron backscatter diffraction (EBSD) pattern from a sample and convert the EBSD pattern to a corresponding light pattern;a first optical system positioned downstream from the device for converting electron patterns into visible light for focusing the light pattern produced by the device for converting electron patterns into visible light;a camera positioned downstream from the first optical system for obtaining an image of the light pattern;an image intensifier positioned between the device for converting electron patterns into visible light and the camera for amplifying the light pattern produced by the device for converting electron patterns into visible light; anda device positioned within the system for protecting the image intensifier from harmful light, the device selected from the group consisting of: a near infrared (NIR) cut-off filter positioned upstream from the image intensifier for preventing NIR light from reaching the image intensifier, a light sensitive sensor in communication with a controller for detecting when a predetermined level of harmful light has been reached, and a short wavelength photo-cathode material positioned within the system to prevent excitation by NIR light. 2. The system of claim 1, wherein the image intensifier is positioned between the device for converting electron patterns into visible light and the first optical system. 3. The system of claim 1, wherein the image intensifier is positioned within the first optical system. 4. The system of claim 1, wherein the image intensifier is between the first optical system and the camera. 5. The system of claim 4, further comprising a second optical system positioned between the image intensifier and the camera. 6. The system of claim 5, wherein the second optical system comprises a focusing lens, a relay lens, a fiber optic relay, or any combination thereof. 7. The system of claim 1, further comprising an alignment mechanism for aligning at least one of the camera and the image intensifier with the device for converting electron patterns into visible light. 8. A method of electron pattern imaging comprising:positioning a device for converting electron patterns into visible light adjacent to a sample;positioning a first optical system downstream from the device for converting electron patterns into visible light;positioning a camera downstream from the first optical system;positioning an image intensifier between the device for converting electron patterns into visible light and the camera;positioning a device within the system for protecting the image intensifier from harmful light;receiving an electron backscatter diffraction (EBSD) pattern from a sample with the device for converting electron patterns into visible light such that the device for converting electron patterns into visible light converts the EBSD pattern to a corresponding light pattern;focusing the light pattern produced by the device for converting electron patterns into visible light with the first optical system;amplifying the light pattern produced by the device for converting electron patterns into visible light with the image intensifier; andobtaining an image of the light pattern with the camera,wherein the device for protecting the image intensifier from harmful light is selected from the group consisting of: a near infrared (NIR) cut-off filter positioned upstream from the image intensifier for preventing NIR light from reaching the image intensifier, a light sensitive sensor in communication with a controller for detecting when a predetermined level of harmful light has been reached, and a short wavelength photo-cathode material positioned within the system to prevent excitation by NIR light. 9. The method of claim 8, wherein the image intensifier is positioned between the device for converting electron patterns into visible light and the first optical system. 10. The method of claim 8, wherein the image intensifier is positioned within the first optical system. 11. The method of claim 8, wherein the image intensifier is between the first optical system and the camera. 12. The method of claim 11, further comprising a second optical system positioned between the image intensifier and the camera. 13. The method of claim 12, wherein the second optical system comprises a focusing lens, a relay lens, a fiber optic relay, or any combination thereof. 14. The method of claim 8, further comprising an alignment mechanism for aligning at least one of the camera and the image intensifier with the device for converting electron patterns into visible light. 15. A system for electron pattern imaging comprising:a device for converting electron patterns into visible light provided to receive an electron backscatter diffraction (EBSD) pattern from a sample and convert the EBSD pattern to a corresponding light pattern;a camera positioned downstream from the device for converting electron patterns into visible light for obtaining an image of the light pattern;an image intensifier positioned between the device for converting electron patterns into visible light and the camera for amplifying the light pattern produced by the device for converting electron patterns into visible light; anda device positioned within the system for protecting the image intensifier from near infrared (NIR) light. 16. The system of claim 15, wherein the device for protecting the image intensifier from NIR light comprises an NIR cut-off filter positioned upstream from the image intensifier for preventing NIR light from reaching the image intensifier. 17. The system of claim 15, wherein the device for protecting the image intensifier from NIR light comprises a light sensitive sensor in communication with a controller for detecting when a predetermined level of harmful light has been reached,wherein the controller shuts down the image intensifier when the predetermined level of harmful light is detected by the light sensitive sensor. 18. The system of claim 15, wherein the device for protecting the image intensifier from NIR light comprises a short wavelength photo-cathode material positioned within the system to prevent excitation by near infrared (NIR) light. 19. The system of claim 15, further comprising a first optical system positioned downstream from the device for converting electron patterns into visible light for focusing the light pattern produced by the device for converting electron patterns into visible light. 20. The system of claim 19, wherein the image intensifier is positioned at one of between the device for converting electron patterns into visible light and the first optical system, within the first optical system, or between the first optical system and the camera. |
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summary | ||
040630986 | abstract | A system for deflecting a beam of particles having different momenta, preferably through a 90.degree. angle, so as to cause the beam to impinge upon a moving target and to scan across the target. The system includes a means responsive to a beam from a suitable source for causing the beam to periodically scan in a scanning plane and further means for deflecting the periodically scanned beam through the desired angle in a deflection plane so that the deflected beam impinges on the target. Means are included in the system for reducing the momentum dispersion at the target in both the deflection and the scanning planes and for spatially focussing the beam so as to produce a desired beam diameter at the target. |
description | This application is a continuation-in-part of, and claims the benefit under 35 U.S.C. § 120 from, nonprovisional U.S. patent application Ser. No. 15/406,737, entitled “Scalable, electro-optically induced force system and method”, filed Jan. 15, 2017, now U.S. Pat. No. 9,984,782. The subject matter of each of the foregoing documents is expressly incorporated herein by reference. The present invention relates generally to techniques for applying a force or inducing a change in momentum to objects using interacting electromagnetic fields and, particularly, to techniques for applying a force or inducing a change in momentum to objects using a laser beam as one of the sources of the electromagnetic fields. The manipulation of samples or devices through the use of electromagnetic (EM) radiation has a number of advantages over mechanical or “tactile” manipulation. Typically, EM manipulation is less mechanically destructive and can be accomplished through mechanical barriers where other more traditional means are not effective. EM manipulation has become more prevalent as technology has advanced and is now accomplished through both constant field applications (as in the case of superconductor facilitated magnetically induced levitation) and oscillating field applications (e.g. laser assisted cooling and trapping). The manipulation of mass through the use of laser light has found many applications as laser technology has evolved. Not simply laser photolysis or spectroscopy, but coherent control of chemical reactions is becoming possible (see P. Brumer and M. Shapiro, Sci. Am., pg. 56, March 1995). Laser atom or molecule trapping has seen a great deal of activity (see S. Chu, Science, pg. 861, 23 Aug. 1991; C. N. Cohen-Tannoudji and W. D. Phillips, Phys. Today, pg. 33, October 1990) and has lead to the observation of Bose-Einstein condensation and the improvement of atomic clocks. Control of larger mass samples with laser energy has also been demonstrated. “Optical tweezers” have been used to stretch single strands of DNA and manipulate chromosomes inside cell nuclei and move entire cellular organelles without destroying the cell wall (see S. Chu, Sci. Am., pg. 71, February 1992). Standing wave laser radiation has also been used to deflect atomic beams in flight (see P. E. Moskowitz, P. L. Gould, and D. E. Pritchard, J. Opt. Soc. Am. B., 2, 11, 1784, 1985). All of these techniques allow for control of small samples with laser light, but none of these is practically applicable to larger samples or efficiently uses the laser light to accomplish the manipulation. One of the difficulties is that many of the current techniques operate by inducing an electric charge polarization in the sample. The force which can be induced by the laser beam is directly related to the degree to which a sample can be polarized before it is damaged. The laser peak intensity must be controlled or the sample can be overheated, ionized or destroyed. This limits the achievable manipulation force. Also, these techniques commonly require the laser to be focused on the target sample, limiting the length of interaction and thus the efficiency with which the laser energy is coupled into translation. Other techniques rely on the transfer of photon momentum in the optical scattering process, but this is extremely inefficient as photons at commonly accessible wavelengths have very little mass. Therefore, it is the object of the present invention to provide a system that 1) employs laser light to apply a force to objects that 2) is scalable, that 3) maximizes the efficiency with which the laser light is utilized for said force, that 4) the intensity of laser light employed by the system should not be limited by the risk of damage to the object upon which the force is induced. The essence of electro-optically induced force and/or momentum is to mimic the repulsion manifested when two like charged wires come into proximity with one another. Wires carrying like charges repel one another due to the mutual opposition of the electric fields generated by the charges on the wires. An electro-optically induced force is realized when one of the wires is replaced with a substitute that maintains an electric field in opposition to the field generated by the first wire (e.g. a photon or radiation field). The principles employed by the present invention to electro-optically induce a force to manipulate objects are now described. FIG. 1 shows a schematic of 1) the electric field around an electrically charged wire parallel to the Z-axis (i.e. perpendicular to the page of the schematic) (100), 2) the projection of said electric field on the Y-axis (vertical in the plane of the page), said projection is described by a sine function (102), 3) the electric field of a laser beam propagating in space perpendicular to the wire, 1) above (104), 4) the electric field of a laser beam retro reflected by a mirror with a surface perpendicular to the laser beam path (106). FIG. 2 shows a schematic of the field interaction, in space at a given instant, between the electric field from a charged wire (200) and the electric field of a laser beam travelling proximate and perpendicular to the charged wire (202) (i.e. for those fields illustrated in FIG. 1). It should be noted that, in the case of a freely propagating laser, a continuous sine wave, and the Y-axis component of the electric field shown, a second sine wave, the net field interaction will always be zero, a function of the orthogonality of sine functions. But, in the case of a retro reflected laser beam, by positioning the retro reflecting mirror judiciously, the spatial relationship between the standing electric field of the wire (200) and the electric field of the laser (202) and can be chosen so the net field interaction between the electric field of the laser and the electric field near the wire, at the given moment illustrated, is non-zero. The net field interaction (204) is shown schematically in FIG. 2. It should be noted that the spatial relationship between the electric field of the laser and the standing electric field near the wire can be altered, and a non-zero net field interaction achieved, using transmissive optics as well. A schematic is shown in FIG. 3. In this case, the change in refractive index, n, between the original propagation medium (e.g. air, for which n˜1) and the medium of the optic (e.g. glass, for which n˜1.5) gives rise to a phase change between the electric field near the wire along the propagation direction of the laser (300) and the electric field of the laser (302). This changes the integral of the interaction between the two electric fields (304) to a, potentially, non-zero value. A final case is shown schematically in FIG. 4. The spatial relationship between the electric field near the wire (400) and the electric field of the laser (402) is altered by directing the laser through an aperture of an electrically conductive material. In such a case the conductive material screens the electric field near the wire (400) from interaction with the electric field of the laser (402) internal to the conductive material. This changes the integral of the interaction between the two electric fields (404) to a, potentially, non-zero value. The above describes methods for achieving a non-zero net field interaction at a given moment in time, meaning a given phase of the laser light, but the electric field of a laser oscillates at frequency given by f=c/λ, where c is the speed of light and λ is the wavelength of the laser. This frequency is typically hundreds of terahertz, much faster than practical electronic signals can be generated, and so any momentary interaction between the laser and the standing electric field of the wire will quickly integrate to zero. In order to sustain and extend a net field interaction over time, it is necessary to vary the electric field near the wire. FIG. 5 shows a schematic of the interactions, in time, between the electric field near a wire driven by a square waveform (500) and the electric field of a laser beam traveling proximate and perpendicular to the wire (502). The accumulated field interaction (504) over one period of the waveform driving the charge on the wire is given byF=∫ sin(ω1(t))*fsg(t,θpm)∂t eq. 1 and is shown schematically in FIG. 5. Where, ω1 is the frequency of the laser beam, fsg is the time-variant electric field near the wire, and θpm is the value of the time-variant optical phase modulation induced to the electric field near the wire. It should be noted that, 1) depending on the relative phase of the given electric fields, the above integral (eq. 1) can be either positive or negative, corresponding to the case of generating a repulsive or attractive force, respectively and 2) although the waveform shown in FIG. 5 is a square wave, that need not be the case. Since the signal driving the electric field on the wire cantilever is electronically generated, fsg can be configured as any function that is determined to facilitate the desired effect. Note is given to the case of a phase modulated sine wave. Consider a specific case: a signal generator operating at 5.64 GHz and common laser wavelength, 532 nm or ˜5.64×1014 Hz. In this case, the laser is operating at a frequency that is one hundred thousand times faster than the signal generator. In this example, the electric field of the laser and the electric field induced near the wire may only be in opposition, in time, space, and intensity, a fraction measured in parts per million (ppm) depending on the specific wave shape and/or modulation induced to the electric field near the wire. But, as described below, a field carrier array may be constructed with very high density, with feature spacing in the sub-micron range. In such a case, one could achieve one million, or more, field interaction locations per meter, and so, even if the electric field of the laser and the electric field near the charge carrier are in opposition a small fraction of the time per oscillation at each location, integrating one million locations over the length of a one meter array, comprised of one million elements, can afford substantial accumulated force between the electric field of the laser and the electric field(s) near the array. The teachings of the present invention are operable in systems where the electric fields are neither free-space nor produced by a laser. The present invention is generally directed to any electro-optical system including a charge carrier configured to carry a charge distribution that gives rise to a first electromagnetic field, and a radiation source configured to generate a second electromagnetic field that interacts with the first electromagnetic field so as to produce a net force on the charge carrier. The present invention allows for the application of laser light to achieve a scalable electromagnetically induced force to samples of greater mass than previous techniques and for more efficient use of the applied laser light. The sample upon which the force is realized must be of specific construction to support an electric charge distribution and variation to achieve a net repulsion or attraction with the laser light. Using this approach the laser need not be focused on the sample nor induce an electric charge polarization on the sample. This allows the force to be integrated over a longer distance, greatly improving the efficiency with which the laser energy is used and increasing the mass amenable to manipulation by the techniques of the present invention. Several recent advancements have driven this invention. First, the laser power available from smaller, less expensive devices is increasing continually. Laser devices are now being constructed that produce watts of laser energy from laser heads that are millimeters or even micrometers in size. As for the magnitude of the accumulated force, let us consider the example of a 10 W laser. Currently such lasers can be fabricated at relatively low cost, and weigh on the order of grams. Ten Watts (10 N*m/s) of optical power, if converted entirely to work (N*m) could lift 1 kg of mass 1 meter per second against earth's gravity (9.8 m/s^2). An array of one thousand 10 W lasers, or the case of a single 10 kW laser, could potentially lift tons. It is the object of this invention to be scalable and to enable such applications. Second, laser energy efficiency is improving as technology advances. As an example, 35-60% of the energy spent to drive some laser diodes is converted directly to laser light energy. Third, circuit features as smaller than 0.25 micrometers can now be fabricated and this technology continues to advance to fabricate features of even smaller size. This feature size is less than half the wavelength of light produced by many high power laser packages. The provides an excellent opportunity to construct a periodic charge distribution giving rise to a periodic electric field that can directly counter the periodic electric field of a laser beam. This is an ideal arrangement for electromagnetic repulsion/attraction (i.e., electro-optically induced force). Signal Generator System FIG. 6 shows schematic of a circuit used to process a commercially available square wave into a signal sufficient to induce a measureable force between a continuous wave single frequency laser and a conductive wire cantilever. A square wave signal generator (600) is set to generate a drive signal at a subharmonic of the laser frequency (e.g. one hundred thousandth of the laser frequency). The output of the signal generator is directed through an adjustable delay line (602). It is advantageous to drive the signal at higher voltage since the field interactions sought scale with the field intensity and thus voltage applied. The signal can be amplified using a broadband amplifier (604) to increase the net force induced. Commercially available signal sources (e.g. square wave clock generators) may have signal jitter in the 100 to 300 femtosecond range, but the laser electric field oscillation for commercially available single frequency lasers (e.g. a 532 nm laser) is approximately 2 femtoseconds. In such a case, the jitter of the square wave will preclude any efficient induction of force on the wire cantilever as it will tend to average out any net repulsion/attraction. It is necessary to improve the square wave signal jitter to provide for stable and efficient field interactions. To accomplish this, the processed square wave signal, described above, is directed into the forward port of a broadband directional coupler (606). The output of the directional coupler is sent to a third adjustable delay line (608) and back to the input port of the broadband coupler (606), creating a loop. The delay line (608) is adjusted to phase lock the resultant signal loop. In effect the delay line (608) is adjusted to ensure that the signal loop is an integer multiple of the square wave length. In such an arrangement, the jitter of the input signal is reduced by the square root of the number of round trips the signal makes in the loop. In the case of a 40 db power coupler, the jitter is improved by a factor of 10. To improve the signal jitter further, the signal from the reference port of the broadband coupler (606) is directed to the forward port a second broadband coupler (610). The output of the second broadband coupler (610) is directed through a fourth delay line (612) and back to the input port of the second broadband directional coupler (610). The delay line (612) is adjusted to phase lock the second resultant signal loop. In effect the delay line (612) is adjusted to ensure that the second signal loop is also an integer multiple of the square wave length. Again, the jitter of the input signal is reduced by the square root of the number of round trips the signal makes in the second loop. In the case of a 40 db power coupler, the jitter is improved by a further factor of 10. In the above description, two successive directional coupler stages should be sufficient to reduce the (e.g. 200 fs) jitter of the square wave clock generator source to a value comparable to the oscillation frequency of the single frequency laser source (approximately 2 fs), but successive stages of broadband couplers and paired delay lines can be employed to reduce the jitter to desired levels. In this embodiment, the wire cantilever configured to achieve a repulsion with the single frequency laser, below, is within the second coupler (610)/delay line (612) loop described above. The interaction site (614) is shown schematically in FIG. 6. The signal from the reference port of the second directional coupler (610) is directed to a monitor oscilloscope (616). FIG. 7, shows a schematic of a preferred embodiment. A continuous wave (CW) single frequency laser (700) produces a beam (702) that is directed to a retro-reflecting mirror (704) to provide the optical electromagnetic field which will induce a force at the wire cantilever site (614) described above and shown schematically in FIG. 6. The square wave clock generator (706), signal processing circuit (708), and monitor oscilloscope (710) are the same as those described in FIG. 6, (600), (602-612), and (616), respectively. The wire cantilever (712) consists of a length from the second directional coupler (610)/delay line (612), above, configured to be in close proximity (e.g. less than 1 mm) to the laser beam (702) and retro-reflecting mirror (704). The system is configured such that the linear polarization of the laser beam (702) is perpendicular to the wire cantilever (712). Interferometric Force Detection System A fiber optic force detection system similar to that used in atomic force microscope studies (D. Rugar, H. J. Mamin, and P. Guethner, Appl. Phys. Lett. 55, 25, (1989) 2588) is employed here to measure force induced between a wire cantilever loop (712) and a laser beam (702). The output of a single-mode fiber-coupled probe laser (714) is directed into a single mode 2×2 fiber coupler (716). One optically cleaved end of the output fiber (718) of the fiber coupler (716) is positioned in close proximity (e.g. single microns) and perpendicular to the wire cantilever (712). A piezo electric actuator, PZT (720) is attached to the optical fiber (718) and driven by a signal generator (722). The PZT (720) drives the motion of the optical fiber (718) perpendicular to the wire cantilever (712). The single frequency laser light exiting the output of the fiber coupler (718) is made incident upon the wire cantilever (712). A portion of the light exiting the optical fiber (718) and incident on the wire cantilever (712) is reflected back into the optical fiber (718) and co-propagates with the light reflected from the internal surface of the optically cleaved fiber (718). The return signal, the optical interferometric signal between the counter propagating reflection of the optically cleaved end of the optical fiber (718) and the reflective surface of the wire cantilever (712), travels back through the optical fiber (718), back through the fiber coupler (716), and is directed into a detector photodiode (724). The output of the detector photodiode (724) is measured using a lock-in amplifier (726), using the reference signal from the signal generator (722) driving the PZT (720). The signal from the lock-in amplifier is observed on a monitor oscilloscope (728). A monitor photodiode (730) can be employed to measure the second output of the fiber coupler (716) to ensure system stability. The constructive and destructive interference between the light reflected from the internal surface of the optically cleaved fiber (718) and the light reflected from the wire cantilever (712), driven and detected at the frequency of the lock-in amplifier (726), allows for a very sensitive detection of movement of the wire cantilever (712) (e.g. nanometers). When the wire cantilever (712) is chosen with a small force constant, this allows for very sensitive force detection (e.g. nanoNewtons). Laser/Carrier Field Interaction The choice of frequency of the square wave generator (706) is critical for manifestation of the desired effect. FIG. 5 shows a condition where the laser frequency (502) is an odd multiple of the cantilever driver frequency (500). In such a case, the positive, rising, half cycle of the square wave is integral over an extra (e.g. positive) half cycle of the laser oscillation and the negative, falling, half cycle of the square wave is integral over a complimentary (e.g. negative) half cycle of the laser oscillation. This frequency, and phase, relationship gives rise to cumulative, non-zero, field interaction (504) between the electric field of the laser (502) and the electric field near the wire cantilever (500) over time. It has been noted, above, that depending on the relative phase of the given electric fields, the above integral (eq. 1) can be either positive or negative, corresponding to the case of generating a repulsive or attractive force, respectively, on the wire cantilever. In the present case, the relative phase between the oscillating electric field of the laser (700) and the oscillating electric field of the wire cantilever (712) can be easily achieved by adjusting the delay line of the signal processing circuit (602). Carrier Array FIG. 8 shows a schematic where a set of field carriers (e.g. a set of wire cantilevers, or alternatively a series of traces on a printed circuit board) (832) is arrayed perpendicular to and along a laser's (800) propagation path (802) and an equivalent set of transmissive optics (804) is paired with each field carrier to facilitate the process, as shown schematically in FIG. 3. The field carriers shown (834) share a common platform, so a single interferometric detection measurement (814-830), as shown in FIG. 7, is sufficient to measure the cumulative force on the ensemble of field carriers. As the circuit path for each carrier is unique, each field carrier signal wave must have separate phase control to facilitate the desired effect. The phases are adjusted to maximize the force detected via the interferometric detector signal described above. While the preferred embodiments, above, employ a single frequency laser, the techniques of the present invention are applicable to multi-mode lasers. Multi-Mode Laser Embodiment Typically, lasers that provide high output power operate in several longitudinal modes of the laser resonator cavity. The wavelengths of the cavity modes of any laser are given by the expression: n I=2L, where, n is an integer, I is the wavelength of the laser light, and L is the length of the laser resonator cavity. As a result, when a multi-mode laser is used several different wavelengths of laser light are produced simultaneously. Therefore, the present invention also includes an embodiment that uses the laser light from multi-mode lasers by providing a path to match each of the laser's active laser cavity modes. Free Space Laser Propagation Embodiment It should be noted that the electric field near a set of charge carriers can be configured such that a non-zero net interaction with the electric field of the laser can be achieved, in the case where the laser is directed to travel proximate and perpendicular to the charge carriers in free space, with no optics in the laser beam path. A schematic is shown in FIG. 9. In this case, the charges on a set of separate carriers (e.g. a set of parallel traces on a PCB) are configured to achieve a non-sinusoidal electric field variation in the direction of laser travel (900). The electric field of the laser beam (902) is shown schematically over the same distance. In such a case, the space integral (904) of the electric field near the charge carrier array (900) and the electric field of the laser (902) can be a non-zero value. FIG. 10 shows a schematic where a set of field carriers (e.g. a set of wire cantilevers, or alternatively a series of traces on a printed circuit board) (1032) is arrayed perpendicular to and along a laser's (1000) propagation path (1004). The field carriers shown share a common platform (1034), so a single interferometric detection measurement (1014-1030) is sufficient to measure the cumulative force on the ensemble of field carriers. As the circuit path for each carrier (1032) is unique, each field carrier signal wave must have separate phase control to facilitate the desired effect. The phases are adjusted to maximize the force detected via the interferometric detector signal described above. Jitter Reduction Circuit with Combiner Embodiment FIG. 11 shows a schematic of an alternate embodiment of a circuit to reduce jitter in an electronic signal, the first embodiment having been presented in FIG. 6. In FIG. 11, a sinusoidal signal generator (e.g. 1 GHz, 300 fs jitter) (1100) is directed into a first input port of a first signal combiner (1102). The first signal combiner (1102) provides an isolation (e.g. 40 db) between the two input ports of the combiner (1102). The output of the first combiner (1102) is directed into the first input port of a, similar, second combiner (1104). The output of the second combiner (1104) is directed to the input of a first adjustable delay line (1106) and the output of the first delay line (1106) is directed into the second input port of the first combiner (1102), thereby creating a feedback loop that propagates through the first delay line (1106). When the delay line (1106) is adjusted such that the length of one round trip through the circuit (1102, 1104, and 1106) is an integral multiple of the wavelength of the signal from the signal generator (1100), the feedback loop serves to reduce both the jitter and amplitude noise of the input signal by the square root of the number of round trips through the circuit, determined by the isolation values of the combiners (1102 and 1104). As an example, if the combiners serve to transmit ninety nine percent of the signal from the first combiner input port to the combiner output port and one percent of the signal exits through the second input port of the combiner, a result of imperfect isolation, the number of round trips through the feedback loop would be approximately ninety eight. In such a case the noise in the above circuit would be reduced by the square root of approximately ninety eight, or nearly ten-fold. In FIG. 11, a second combiner set is shown (1108 and 1110). The purpose of the second combiner set (1108 and 1110) is to reduce the signal noise further. The signal from the second input of the second combiner (1104), the leakage from the second combiner, is directed to the first input port of a third signal combiner (1108). The output of the third combiner (1108) is directed to the first input port of a fourth combiner (1110) and the output of the fourth combiner is directed to the input of a second adjustable delay line (1112). The output of the second delay line (1112) is directed to the second input of the third signal combiner (1108), thereby creating a second feedback loop. The length of the second feedback loop is adjusted to be an integral multiple of the wavelength of the signal from the signal generator (1100). Thus the second set of combiners (1108 and 1110) and the second delay line (1112) serve to further reduce the signal noise from the output of the second signal combiner (1104) (e.g. by a further factor of 10) The signal, leakage, from the second input port of the fourth signal combiner (1110) is directed to a commercial circuit (1114) that converts a sinusoidal signal to a square wave signal, thus providing the signal upon the charge carriers in FIGS. 7, 8 and 10 (712, 812/832, and 1012/1032, respectively) sufficient to induce a measurable force between a continuous wave single frequency laser and a conductive charge carrier or series of charge carriers. While the present invention has been described with reference to a few specific embodiments, the description is illustrative of the invention and is not to be construed as limiting the invention. Various modifications may occur to those skilled in the art without departing from the true spirit and scope of the invention as defined by the appended claims. |
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claims | 1. A microscope system comprising:a scanning electron microscope for generating an electron beam operating with electron energy greater than an absorption-edge energy of an element of interest and irradiating a test object to induce secondary fluorescence X-ray radiation from the element of interest;an image detector comprising a detector array with a two dimensional array of elements; anda zone plate lens for selectively imaging the secondary fluorescence X-ray radiation, generated by the electron beam of the scanning electron microscope, from the element of interest onto the image detector. 2. A microscope system as claimed in claim 1, wherein the lens is a chromatic lens. 3. A microscope system as claimed in claim 1, further comprising a spectral filter. 4. A microscope system as claimed in claim 3, wherein the spectral filter is a multilayer optic. 5. A microscope system as claimed in claim 3, wherein the spectral filter is a crystal. 6. A microscope system as claimed in claim 1, further comprising a pupil aperture for improving preferential imaging of the secondary radiation. 7. A microscope system as claimed in claim 1, wherein the zone plate lens has a high diffraction efficiency for an energy of a fluorescence line of the element of interest. 8. A microscope system as claimed in claim 1, wherein the lens is made of a compound including the element of interest. 9. A microscope system as claimed in claim 1, wherein the element of interest is aluminum and the lens comprises aluminum. 10. A microscope system as claimed in claim 1, wherein the element of interest is copper and the lens comprises copper. 11. A microscope system as claimed in claim 1, wherein the element of interest is tantalum and the lens comprises tantalum. 12. A microscope system as claimed in claim 1, wherein the element of interest is phosphorus and the lens comprises a phosphorus compound. 13. A microscope system as claimed in claim 1, wherein the element of interest is boron and the lens comprises a boron compound. 14. A microscopic imaging method, comprising:irradiating a test object with an electron beam of a scanning electron microscope;selectively imaging induced secondary fluorescence X-ray radiation from the test object, being generated by the electron beam of the scanning electron microscope, with a chromatic lens; anddetecting the secondary fluorescence X-ray radiation imaged by the chromatic lens using an image detector with a two dimensional array of elements. 15. A method as claimed in claim 14, further comprising aperturing the secondary fluorescence X-ray radiation prior to detecting the secondary fluorescence X-ray radiation to improve preferential imaging of the secondary radiation. 16. A method as claimed in claim 14, wherein the chromatic lens is a zone plate lens. 17. A method as claimed in claim 16, wherein the zone plate lens has a high diffraction efficiency for an energy of a fluorescence line of an element of interest. |
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048851248 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a portion of a reactor containment structure 2 in the region which includes a seal table room. Structure 2 is formed to have a recess 4 which houses a seal table 6 permanently connected to the reactor pressure vessel and carrying a plurality of high pressure fittings 8 each supporting a respective thimble (not shown) while maintaining a pressure seal between the interior of the reactor and the surrounding region, which includes recess 4 and which is at atmospheric pressure. Each fitting 8 receives a respective tube, or casing, 10 defining a passage which communicates with the interior of the associated thimble. A plurality of tubes 10, one for each thimble, is supported by a movable frame 12 which also carries several multi-path transfer devices 14. Each transfer device 14 is provided at its bottom with a plurality of, for example ten, fittings each connected to a respective tube 10 and at its top with a single fitting connected to a further tube, or casing, 18. Internally, transfer device 14 contains a mechanism which is rotatable to establish a transfer passage between tube 18 and any selected one of the tubes 10 connected to that transfer device 14. Additional transfer devices identical to device 14 are each connected to a separate set of tubes 10 and to a separate further tube similar to tube 18. The tube 18 of each transfer device 14 is coupled to one port of a respective further multi-path transfer device 20 having, at one side, a plurality of, for example five, ports each provided with a respective fitting 22 and, at its side opposite from fittings 22, a single port provided with a further fitting 26. Fitting 26 is coupled to a conduit 28 which is mounted to a neutron detector drive unit 30. Device 20 and drive unit 30 rest on a floor 24 which is connected to structure 2. Similarly to each transfer device 14, each transfer device 20 is provided internally with a mechanism which can be rotated to establish a transfer passage between fitting 26 and any selected one of fittings 22. A complete system includes a plurality of transfer devices 20 each having an associated drive unit 30 and each associated with a respective transfer device 14. However, it is common practice to provide further tubes via which other fittings 22 of a transfer device 20 can be connected to other transfer devices 14 for emergency operation. In addition, one fitting 22 of each transfer device 20 is coupled to a common thimble for calibration purposes and one fitting 22 of each transfer device 20 is connected to an associated neutron flux detector storage device. Drive unit 30 contains a cable which is equipped at its end with a neutron flux detector. With transfer devices 14 and 20 set to appropriate positions, drive unit 30 is operated to cause a flux detector to traverse tube 18 and a selected one of tubes 10 and to be advanced along the corresponding thimble in order to obtain neutron flux readings at selected points along the length of the thimble. All of the structure described thus far is known in the art. Heretofore, to effect temporary removal of an obstruction which has developed in a thimble, it was the practice to disconnect tubes 10 from fittings 8 and move frame 12 away from seal table 6. This action would not, by itself, disturb the pressure seals provided by fittings 8, the interior of each thimble being, in any event, at atmospheric pressure. Then, a cleaning tool was inserted directly into the affected thimble via the associated fitting 8. When this was done, the cleaning action frequently caused undue movements of the thimble in the region of the associated pressure fitting 8, resulting in loosening of the fitting components and thereby reducing the ability of the fitting, which may be a Swagelok.TM. device, to hold the thimble in place against the high pressure existing within the reactor pressure vessel. Therefore, it has occurred that the high pressure existing within the reactor pressure vessel would result in a forceful ejection of the thimble, followed by expulsion of radioactive coolant in the form of steam. According to the invention, this undesirable result is prevented by effecting an obstruction removal operation while moveable frame 12 remains in place and tubes 10 remain connected to their associated fittings 8. The presence of tubes 10 helps to reduce stresses on fittings 8 during a cleaning operation, while if a fitting should be loosened, the presence of frame 12 will block complete ejection of the associated thimble. In order to effect obstruction removal, there is provided, according to the present invention, a manual drive unit 32 containing a cable, which can be identical to the neutron detector cables generally used in such systems, the cable being wound on a reel within unit 32 and the reel being supported by a hub 34 provided with a slot for receiving a tool which will effect manual rotation of the reel. The cable extends through a pipe 36 having, at its free end, an adapter fitting 38, which may be a Swagelok.TM. device constructed to be coupled to that end of tube 18 which is normally connected to an associated fitting 22. In order to effect removal of an obstruction, tube 18 is removed from its associated fitting 22 on transfer device 20 and reconnected to adapter fitting 38, tube 18 normally having a sufficient length and flexibility to permit its end to be displaced by the required amount. If a flux detector is present in tube 18, the above step must be preceded by retraction of the flux detector into device 20. Transfer device 14 is actuated, if necessary, to connect tube 18 to the appropriate tube 10, after which hub 34 is rotated to advance the cable disposed in unit 32, with a suitable cleaning brush attached to the free end of the cable, through tube 18, transfer device 14, the connected tube 10, and the associated thimble until the obstruction has been reached and cleared. The cable in unit 32 is given a sufficient length to cause the attached cleaning tool to travel as far as necessary along a thimble. If obstructions must be removed from several thimbles connected to the same transfer device 14, hub 34 is rotated to retract the cleaning tool at least into tube 18, after which transfer device 14 is rotated to establish connection with the tube 10 associated with the next thimble to be cleaned and the cleaning operation described above is repeated. If it is necessary to remove an obstruction from a thimble which is associated with another one of the transfer devices 14 whose tube 18 is connected to another port of the illustrated transfer device 20, the tube corresponding to tube 18 which is connected to the other fitting 22 of transfer device 20 is disconnected from its associated fitting 22 and connected to adapter fitting 38, in the manner described above. If it is desired to limit the travel of the cleaning tool along a thimble, the associated cable can be provided with a suitable stop in a manner known in the art. After an obstruction has been removed, the cable housed in unit 32 can be fully retracted until the cleaning tool comes into contact with fitting 38. Then, tube 18 can be removed from fitting 38 and the cleaning tool can be replaced with a "dummy" neutron detector, which is a simple metal piece having the same external shape and dimensions as a neutron detector. Then, tube 18 can be reconnected to fitting 38 and the cable housed in unit 32 can be reintroduced into the thimble to advance the dummy detector along the length of the thimble and thus assure that the obstruction has been satisfactorily cleared. For clearing many types of obstructions, it is sufficient to employ the same type of cable as that used to displace a detector, but with no detector attached, in which case obstruction removal is effected by the exposed cable end. However, in order to improve the obstruction removal operation, it is preferred to mount at the free end of the cable a cleaning tool in the form of a brush similar to those employed for cleaning rifle barrels. An example of such a cleaning tool is illustrated in FIG. 2 and is composed of a pair of twisted wires 40 carrying bristles 42, only a few of which are shown, extending in different angular directions to form a cylindrical brushing surface. Wires 40 are secured, as by welding, to a connector 44 provided with an external screw thread. A connector 46 secured to the free end of a detector cable for attachment of the cleaning tool of FIG. 2 is shown in FIG. 3. Connector 46 is provided at its right-hand end and with an internally threaded bore 48 in which connector 44 of FIG. 2 can threadedly engage. The left-hand end of connector 46 is provided with a bore which receives the solid core 50 of a detector drive cable. Core 50 is surrounded by a helically wound wire 52 via which displacement forces can be imposed on the cable. Core 50 may be welded or otherwise bonded to connector 46, which can alternatively be used to carry a dummy flux detector. While the description above shows particular embodiments of the present invention, it will be understood that many modifications may be made without departing from the spirit thereof. The pending claims are intended to cover such modifications as would fall within the true scope and spirit of the present invention. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive, the scope of the invention being indicated by the appended claims, rather than the foregoing description, and all changes which come within the meaning and range of equivalency of the claims are therefore intended to be embraced therein. |
051679124 | abstract | An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. |
062193996 | claims | 1. A maintenance method in a nuclear power plant including a reactor primary containment vessel provided with a suppression pool of a suppression chamber, comprising the steps of: improving clearness of a water in the suppression pool and decontaminating an interior of the suppression pool; and inspecting a coated film applied on an inner surface of a wall portion of the suppression pool, said improving, decontaminating and inspecting steps being performed while maintaining a water level in the suppression pool. 2. A maintenance method in a nuclear power plant according to claim 1, further comprising the step of carrying out a repair coating to the coated film through the underwater operation in the suppression pool after inspecting the coated film applied on the inner surface of the suppression pool wall. 3. A maintenance method in a nuclear power plant according to claim 1, further comprising the steps of preparing a film thickness measuring device into the suppression pool and measuring a thickness of the coated film on the inner surface of the suppression pool wall by using the film thickness measuring device. 4. A maintenance method in a nuclear power plant according to claim 1, further comprising the step of preparing an underwater plate thickness measuring device into the suppression pool and measuring a plate thickness of a plate constituting the suppression pool wall by using the plate thickness measuring device. 5. A maintenance method in a nuclear power plant according to claim 1, further comprising the step of closing a strainer provided on the inner surface of the suppression pool wall in an underwater operation and inspecting a valve of a piping communicating with the outside of the suppression pool through the strainer. 6. A maintenance method in a nuclear power plant according to claim 1, further comprising the step of welding defect portions and portions to be repaired of the suppression pool wall, inner structure of the suppression pool, ducts, machineries and duct supports in an underwater operation in the suppression pool. 7. A maintenance method in a nuclear power plant according to claim 1, further comprising the steps of welding defect portions and portions to be repaired of the suppression pool wall, inner structure of the suppression pool, ducts, machineries and duct supports in an underwater operation in the suppression pool and carrying out non-destructive test to said welded portions in an underwater operation in the suppression pool for quality inspection. 8. A maintenance method in a nuclear power plant according to claim 1, further comprising the step of carrying out a cutting working for repairing an inner structure of the suppression pool, ducts, machineries and duct supports in an underwater operation in the suppression pool. 9. A maintenance method in a nuclear power plant according to claim 1, wherein the clearness and the decontamination of the water in the suppression pool is performed by collecting substances floating in the water of the suppression pool and removing and collecting sludge substance deposited on an inner bottom surface of the suppression pool wall. 10. A maintenance method in a nuclear power plant according to claim 9, wherein the substance floating in the water of the suppression pool is sucked up together with the surrounding water out of the suppression pool by suction means movable in or above the water. 11. A maintenance method in a nuclear power plant according to claim 10, wherein the substances and the surrounding water sucked up in the suppression pool are subjected to solid-liquid separation on an outside the suppression pool. 12. A maintenance method in a nuclear power plant according to claim 9, wherein said suction means comprises a rotating brush and suction port arranged around the rotating brush so as to suck up the sludge substance on the inner bottom surface of the suppression pool together with the surrounding water therein. 13. A maintenance method in a nuclear power plant according to claim 1, wherein the decontamination of the interior of the suppression pool includes removing of sludges and deteriorated matters on the inner surface of the suppression pool wall. 14. A maintenance method in a nuclear power plant according to claim 13, wherein the sludges on the inner surface of the suppression pool wall are removed through a sucking step by using a suction means comprising a rotating brush and a suction port arranged around the rotating brush so as to suck up the sludges and deteriorated matters together with the surrounding water in the suppression pool. 15. A maintenance method in a nuclear power plant according to claim 14, wherein the sludges and the deteriorated matters sucked up with the surrounding water in the suppression pool are subjected to solid-liquid separation on an outside of the suppression pool. 16. A maintenance method in a nuclear power plant according to claim 1, wherein the inspecting step of the coated film applied on the inner surface of the suppression pool wall is performed by visually observing an surface condition of the coated film by using an underwater camera unit. 17. A maintenance method in a nuclear power plant according to claim 16, wherein the underwater camera unit includes at least one of a fixed camera disposed in the suppression chamber and a camera movable in the water of the suppression pool. 18. A maintenance method in a nuclear power plant according to claim 17, wherein the underwater camera unit includes a fixed camera disposed in the suppression chamber and a camera movable in the water of the suppression pool, said fixed camera being used to set a general inspection position in the suppression pool and said movable camera including a first movable camera used to set a fine position approaching the inspection point and a second movable camera used to observe a state of the coated film while approaching the inspection position more closely than the first movable camera. 19. A maintenance method in a nuclear power plant according to one of claim 17, wherein information obtained by the underwater camera unit is displayed on a monitor television disposed outside the suppression pool to thereby allow observation in the air. 20. A maintenance method in a nuclear power plant according to claim 1, further comprising the step of carrying out a repair coating to the coated film through the underwater operation in the suppression pool after inspecting the coated film applied on the inner surface of the suppression pool wall. 21. A maintenance method in a nuclear power plant according to claim 20, wherein said repair coating step to the coated film on the inner surface of the suppression pool wall is carried out by peeling off a deteriorated or deformed coated film at a portion to be repaired by using one of a disc sander and a grinder provided with a suction means and a substrate treatment is carried out by sucking up the surrounding water out of the suppression pool. 22. A maintenance method in a nuclear power plant according to claim 20, wherein said repair coating step of the coated film on the inner surface of the suppression pool wall is carried out by applying an underwater coating to the inner surface of the suppression pool wall by using one of a brush having suction means arranged around the brush, a roller and other coating means and a coating splashed during the underwater coating step is sucked up outside the suppression pool together with the surrounding water. |
summary | ||
050376020 | description | DETAILED DESCRIPTION OF THE INVENTION The following description is of the best presently contemplated mode of carrying out the invention. This description is not to be taken in a limiting sense, but is made merely for the purpose of describing the general principles of the invention. The scope of the invention should be determined with reference to the appended claims. In making reference to the drawings, like numerals will be used to refer to like parts throughout. At the outset, it is noted that the following detailed description is based on an RFQ accelerator which is commercially available from Science Applications International Corporation of San Diego, Calif. A good description of this RFQ may be found in Appendix B. submitted herewith. Appendix B comprises a paper presented at The First European Accelerator Technology Conference, held in Rome, Italy, in June of 1988. The paper is entitled "A Compact 1 MeV Deuteron RFQ Linac." The authors of the paper are D. A. Swenson and P. E. Young. Further, the target system is based on the eight position target handling system which is commercially available from Scanditronix of Uppsala, Sweden. Some information relative to the target system is provided in Appendix A. It is noted that the information presented in Appendix A does not necessarily relate to the Scanditronix-based target system. Rather, much of the information is background information related to target systems in general. At least some portions of Appendix A, e.g., describing the "windowless target system" present a novel approach, never before utilized (to Applicants' knowledge), that offers significant advantages over other types of target systems. Referring first to FIG. 1, a block diagram of a system 12 for producing radionuclides for application to PET is shown. Essentially, this system includes an accelerator subsystem 14, a targetry subsystem 16, a control subsystem 18, and an accelerator support subsystem 20. (Hereafter, these subsystems may be referred to by their identifying name without including the term "subsystem" therewith, e.g., the targetry 16. Moreover, the terms "subsystem" and "system" may be used interchangeably.) It is the function of the accelerator 14 to accelerate a beam of .sup.3 He.sup.++ ions to an energy level of approximately 8 MeV. It is the function of the targetry 16 to receive this accelerated beam, expose a target material thereto, and generate selected precursors from the resulting radionuclides (created by irradiating the target material with the accelerated beam). In turn, these precursors are presented to an automated pharmaceutical system 22 that is programmed to produce one or more desired radiopharmaceuticals used by a patient 24 undergoing PET. The control subsystem 18 provides the control signals for automatically operating the accelerator 14 and the targetry 16, as initiated by a technician 26. Similarly, the accelerator support system 20 provides the necessary support functions associated with the operation of the accelerator, e.g., vacuum pumps, cooling mechanisms, and the like. Operation of these support functions is monitored and controlled (as required) by the technician 26 through the control subsystem 18. The accelerator 14 includes an ion source 30 for generating (or otherwise producing) the .sup.3 He.sup.++ ions used by the system. This source may be conventional, such as a duoplasmatron ion source. Advantageously, .sup.3 He is commercially available at a modest cost. The ions from the source 30 have a low energy associated therewith, on the order of 0.05 MeV. The low energy ions from the source 30 are presented to a Low Energy Beam Transport (LEBT) apparatus 32 where they are focused and otherwise tailored for injection into a Radio Frequency Quadrupole (RFQ) linear accelerator ("linac") 34. The RFQ linac 34 accelerates the beam to an energy of 8.0 MeV. A High Energy Beam Transport (HEBT) apparatus 36 then directs or presents the beam to the targetry 16. The HEBT 36 may be any suitable apparatus as is known in the art, e.g., a series of magnets or simply a beam pipe through which the high energy beam drifts. The accelerated beam may be selectively directed to a beam dump apparatus 38, e.g. a block of lead, in the event portions of the accelerator 14 are being tested and it is not desired to direct the beam to the targetry 18. Advantageously, the RFQ-based accelerator system 14 has no beam activation problems as are common with prior art proton/deuteron beam systems. There is very little beam loss within the RFQ and there is no beam loss associated with the extraction process. Further, no shielding is required around the RFQ 34, thereby significantly reducing the quantity of shielding required. Moreover, accelerator maintenance is not complicated by shielding enclosures or activation problems. The accelerated beam, after drifting a short distance through the HEBT 36, passes through a vacuum isolation valve into the isotope-production targetry system 16. The beam is allowed to expand during this drift to reduce the power density on the thin foils separating the accelerator vacuum from the target material (usually a gas) in the targetry system. The targetry system 16 includes at least one target material 40 and a plurality of precursor units 42. When the target 40 is bombarded with the high energy beam from the accelerator 14, various reactions occur (known to those skilled in the art) resulting in the creation of certain radionuclides. Further details concerning preferred target materials, the reactions that occur, and the resulting precursors obtained, are presented in Appendix A. As has been indicated, one of the advantages of the present invention is that the targetry 16 may be realized using commercially available target systems, modified only to accommodate .sup.3 He.sup.++ targets. An example of such a system is the target handling system manufactured by Scanditronix of Sweden. Such commercially available targetry subsystems may include, either as an integral part thereof or as an option, a suitable automated pharmaceutical system that programmably utilizes the precursors to produce a desired radiopharmaceutical. Because the targetry system 16 and the automated pharmaceutical system 22 are generally known in the art, further details associated with the systems will not generally be presented herein. Of particular interest, and unlike most reactions for proton and deuteron-based systems which involve neutrons in the final state, most of the .sup.3 He-based reactions involve a charged particle in the final state. Such particles can be easily shielded by sheets of aluminum or the target casing itself. Accordingly, the .sup.3 He-based reactions of the present invention significantly reduce the neutron production in the targets relative to that in the proton and deuteron targets. For example, if the radioisotope produced by the present invention is .sup.11 C, the ratio of neutrons produced to radionucleus produced is 0.5. If the radioisotope produced by the present invention is .sup.18 F, the ratio is 0.08. Since .sup.18 F is by far the most widely used PET isotope, the present invention is thus ideal for its production because of this low ratio of neutrons/radionucleus. This low neutron production significantly reduces the shielding requirements of the system. Still referring to FIG. 1, it is seen that the accelerator support 20 includes a vacuum subsystem 44, a thermal control subsystem 46, an RF power subsystem 48, and an instrumentation subsystem 50. These subsystems are described more fully below in connection with the descriptions of FIGS. 3 and 8-10. Referring next to FIG. 2, a pictorial diagram of the system 12 of the present invention is shown. This figure is presented primarily to illustrate the relative sizes of the various components of a preferred embodiment of the system 12. As shown in FIG. 2, the control subsystem 18, as well as portions of the accelerator support subsystem 20, are generally included in standard size electronic equipment racks 52 placed adjacent the accelerator 14. Other portions of the accelerator support subsystem 20, such as pumps 54 and 56, and associated tubing or plumbing, as well as suitable mechanical support structure 60 (e.g., a rigid table upon which the RFQ 34 is mounted) are positioned at convenient locations around (e.g., under) the accelerator 14. In this preferred embodiment, the RFQ linac 34 is only 3.4 meters long and is enclosed in a 0.3 meter diameter vacuum tank. Thus, the length of the linac 34 is approximately ten feet, while the ion source 30 and LEBT 32 are only about two feet in length, making the overall length of the accelerator system only about twelve feet. The rf (radio frequency) power requirement for the RFQ structure and beam is about 400 kw peak or 8 kw average assuming a 2% duty cycle. This power is provided by 16 small power amplifier tubes (FIGS. 5D, 5E), mounted inside the RFQ vacuum tank and close coupled to the linac structure. The linac structure and power amplifiers are cooled by two separate water cooling systems, described more fully below in connection with FIG. 8. The RFQ tank is evacuated by two turbomolecular pumps to an operating pressure of about 1.times.10.sup.-6 Torr. The entire vacuum system is described more fully below in connection with FIG. 7. The performance and operational parameters of the RFQ linac 34 are summarized below in Table 1. TABLE 1 ______________________________________ RFQ Linac Parameters ______________________________________ Particle He3++ Frequency 425 MHz Charge 2 proton units Structure length 3.40 m Injector voltage 25 kV Input energy 50 keV Output energy 8.0 MeV Ion source current 30 mA Output current electrical 15 mA particle 7.5 mA Output emittance .005 cm-mrad Pulse repetition rate 120 Hz Pulse length 166 us Pulse duty factor 2.0 % Average current electrical 300 uA particle 150 uA Radial aperture 0.15 cm RF power cavity (peak) 280 kW beam (peak) 120 kW total (peak) 400 kW total (average) 8 kW Weight (RFQ) 300 kg ______________________________________ Still referring to FIG. 2, it is noted that the racks 52 of electronic equipment are roughly eight feet in length, two or three feet in width, and typically no more than six or seven feet in height. Hence, the accelerator 14, including its support subsystems 18 and 20, can be placed in an extremely compact space compared to the cyclotron-based systems of the prior art (which systems typically occupy at least three times the floor space as do the equivalent components of the present invention). Moreover, the concrete shielding 62 placed around the targetry 16 need only be two feet in width, compared to the minimum of four feet in width that is used by equivalent target systems employed in a proton/deuteron-based system. Referring next to FIG. 3, a more detailed block diagram of the radionuclide production system of the present invention is shown, with emphasis on the control features and elements thereof. This diagram will be explained by discussing the control and operation of the main components thereof, i.e., the ion source 30, the low energy beam transport 32, the RFQ 34, and the targetry subsystem 16. Referring first to the ion source 30, this source is preferably a conventional duoplasmatron operating at 25 kV. Such an apparatus produces energies of 50 keV for the doubly charged helium ions. The duoplasmatron comprises two major assemblies: a plasma generator and an extraction electrode assembly. Helium-3 gas, which is readily commercially available from numerous sources, is injected into the plasma generator and is ionized through an arc discharge with electrons emitted from a heated filament. A focussing magnetic field is placed at the aperture of the source to enhance the ionization efficiency of the ion source. The generated plasma flows out of a small aperture in the anode and becomes the source of ions that are extracted through the extraction electrode. A suitable duoplasmatron that can be used as the ion source 30 is the model Ionex 740A, manufactured by General Ionex Corporation. This device provides an output current (ion flow) of 30 mA. This is more than sufficient for proper operation of the RFQ 34, and the additional capacity provides a margin of performance, thereby insuring that sufficient current is always available at the input to the RFQ. The gas flow rate from the ion source 30 is preferably maintained at less than 0.01 Torr-liter/sec. This is achieved by maintaining the ion source at operating pressure of 10.sup.-5 Torr with the vacuum system 44. The source of helium-3 gas is stored in a small bottle located in one of the equipment racks 52 (FIG. 2) and transported to the ion source 30 by flexible tubing. Advantageously, helium-3 gas is commercially available at a cost of around $160/liter. The estimated cost for a .sup.3 He RFQ facility is only about $2,700/year, thereby contributing to the low operating cost of the system. The ion source 30 is mounted on one end of the accelerator assembly 14 in a metal enclosure. This enclosure further serves as a grounded shield around the plasma generator, which is at a potential of 25 kV. The plasma generator is about 17 cm in diameter, 21 cm long, and is isolated by a vacuum tight, electrically insulating cylinder. Because the plasma generator operates at a relatively low voltage, atmospheric air is used for electrical insulation in the ion source housing. Four Ion Source power supplies 64 provide the various dc voltages and currents required to operate the ion source 30. Three of these supplies (arc, filament and magnet) are at the plasma generator potential and are isolated by 20 kV from ground. In the preferred embodiment, the Arc supply is adjustable to 150 V dc, and provides a pulsed output current of up to 10 amps. The rise time of the arc current is carefully controlled by a transistorized modulator so as to provide a beam current rise time of a few microseconds. The repetition rate is also adjustable over a range of 100 Hz to 1.2 kHz through the control system. The power supply operates from a single 120 V, single phase, 60 Hz isolated ac power source. The filament power supply, used to supply a current to the filament of the plasma generator, is adjustable from zero to 8 V dc, and supplies a current of up to 80 A. Power is derived from the isolated 120 V, single phase, 60 Hz ac power source. The magnet power supply, used to power the focussing magnets of the ion source, is adjustable from zero to 75 V dc, and provides up to 4 A of current. It also operates from the 120 V, single phase, 60 Hz isolated ac power source. The extraction power supply is adjustable up to 30 kV dc and provides currents of up to 50 mA pulsed and 0.5 mA continuous. This power supply also operates from the 120 V, single phase, 60 Hz ac power source, and is referenced to ground potential. All of the power supplies 64 contain internal regulators to stabilize the output voltage and/or current to within 1% of the required value due to variations in line voltage (.+-.5%) and load impedance (.+-.10%). The voltage ripple at the dc output of the power supplies should be kept at less than 1% to ensure proper operation of the ion source 30. The power supplies 64 are controlled, and their status monitored, through the computer based control system 18. Those power supplies referenced to the ion source potential (20 kV) also have a fiber optic control interface so that the critical control components will be at ground potential. High speed analog voltage and current waveforms are transmitted to the control system through fiber-optic coupled Voltage-to-Frequency convertors. The ion source power supplies 64 are preferably located in free standing, grounded metal enclosures that are part of the equipment racks 52, and are conveniently positioned near the accelerator. A high voltage insulated power cable assembly couples the three isolated power supplies and up to eight channels of instrumentation and control signals to the elements of the ion source 30. The exterior of this power cable is a flexible metal tubing which is grounded for personnel safety and protection. All of the power supplies 64 may be obtained from commercially available sources. Turning now to the Low Energy Beam Transport (LEBT) system 32, the function thereof is two fold, namely: (1) to accept the charged particle beam from the ion source 30 and to focus it into a strongly converging beam for injection into the RFQ 34; and (2) to provide a high-conductance vacuum port for pumping the gas load that emanates from the ion source. Conventional apparatus, known to those skilled in the art, is used to achieve these two functions. The beam entering the LEBT 32 is focused using an rf conventional beam lens configuration. This beam lens configuration, based on rf electric fields, has a strong focal action for low energy particle beams. Further this particular lens configuration may be used at a substantially lower frequency than the RFQ frequency. Rf power for the lens is produced by an LEBT rf power source 66. As is known to those skilled in the art, the rf beam lens has distinct advantages over electrostatic quadrupole lens combinations in that no high voltage insulators are required to support the resonant electric fields, and the temporary alternation of polarity of the fields provides the alternating gradient feature required by the particle beam dynamics. Moreover, the beam maintains a near circular cross section throughout the lens which has important consequences in preserving the emittance of space-charge dominated beams. Further, the lens has the same focal length in both transverse planes and is tunable in both planes simultaneously by a single knob--the rf field amplitude. Advantageously, the lens has no frequency or phase constraint relative to the RFQ linac, and is thus easily activated by simply energizing the rf power source 66. Still referring to FIG. 3, and also to FIGS. 4A and 4B, the RFQ linac 34 will now be described. As has been indicated, the preferred RFQ linac 34 for use in the system 12 is a commercially available RFQ device available from Science Applications International Corporation of San Diego, Calif. The description of the device herein is presented is intended only to clearly show how this commercially available device is integrated into the radioisotope production facility of the present invention. Essentially the RFQ 34 is a cylindrical pipe 80, loaded with four scalloped vanes 82. The vanes are installed in a high vacuum enclosure, and excited with rf power. The vacuum system 44 provides the requisite vacuum, and the RFQ rf power system 48 provides the requisite rf power. The vane tips define a tiny aperture 84 along the axis of the cylinder through which a particle beam passes. The rf power excites an rf cavity mode that has a strong quadrupole electric field pattern in this aperture that focuses the particle beam, keeping it small and away from the vane tips. Ripples on the vane tips introduce a longitudinal component of electric field along the axis that accelerates the particle beam. The pipe or tube 80 is the main structural element of the RFQ. This tube and the four vanes 82 are made from aluminum. The vanes are mounted inside the tube on a number of concentric push/pull screw assemblies 86. These assemblies 86 hold the vanes 82 in position and provide for their precise alignment using conventional means such as micrometer threads, precision alignment surfaces, and a locking plate. The majority of the external surfaces are copper plated for electrical conductivity. The vacuum requirement is enormously simplified by surrounding the entire RFQ assembly 34 with a simple vacuum manifold, thereby eliminating hundreds of vacuum seals that would otherwise be required. Advantageously, the RFQ design provides low fabrication costs, lightweight structure, easy assembly and disassembly, removable vanes, design flexibility, rigidity, superb alignment capabilities, and excellent vacuum properties. The cross section of the preferred RFQ cavity is shown in FIGS. 4A and 4B. The RFQ resonates at 425 MHz and has an inside diameter of 6.200 inches (15.748 cm), a radial aperture of 1.5 mm, and constant vane-tip radius of 1.28 mm. As has been indicated, the mechanical design is based on the use of a heavy-walled aluminum tube 80 (8" OD, 6" ID) as the main structural element of the assembly. After all welding on the assembly is completed, the assembly is stress relieved before final machining. The latter includes boring the inside of the cylinder to the precise diameter of 6.20 inches, and machining four precision flats 88 on the outer surface of the cylinder. Extreme care must be taken to insure that these flats are parallel to and equidistant from the axis of the interior surface and parallel or perpendicular to each other. The preferred RFQ is 3.4 meters long and is configured as two 1.7 m long RFQ's connected in tandem. Fabrication and operational advantages result from this end-to-end configuration over a single-long-tank configuration. The four RFQ vanes 82 are mounted inside the heavy-walled aluminum tube (the vane housing) as shown in FIGS. 4A and 4B. Electrical contact between the vanes and the vane housing is based on flexed fins at the base of the vanes, which are designed to produce a force of 100 pounds/inch or greater against the vane housing. The range of fin flexure is designed to allow mechanical alignment of the vanes with a tolerable effect on this contact force. Each vane 82 is held in position by 14 pairs of concentric push/pull screw assemblies 86 as shown in FIG. 5B. The pushing screws have a micrometer thread to the vane housing and form the vane-base alignment surfaces. The pulling screws serve to pull the vane bases against these alignment surfaces. The locking plates load the alignment screw threads to prevent accidental movement. The RFQ vanes 82 are designed in conventional manner with the vane tips extending close to the end plates of the RFQ cavity with a cutout between the vane tips and the vane bases to allow the rf magnetic fields to wrap around the ends of the vanes. A profile, end and side views, of the vane termination is shown in FIG. 5A. The gap between the vane tip and the end plate is 0.500 cm. the cutout has an area of about 13.2 cm.sup.2. The vane base makes electrical contact with the end plate through a segment of a spring ring in a groove in the end of vane base. Preferably, the vanes 82 are fabricated from the aluminum alloy 7075, which has the best spring properties for the flexed fins. The vane material is purchased as rectangular bars with gun-drilled cooling channels through their long dimensions. The bars, bolted to a rigid machining fixture, are machined to the desired cross section by conventional CNC milling machines. At this stage, the vane tip is still in the form of a rectangular blade 0.256 cm thick. The ends of the vanes are cut off and contoured by a computer-controlled wire electrical discharge machining (EDM) process. The last step in the machining of the vanes is to put the delicate contours on the vane tips. The longitudinal vane-tip profile involves a numerical solution of the idealized RFQ potential function. Computer Aided Machining (CAM) processes translate most cutting processes into straight line segments and circular arcs. Using these segments, the standard vane-tip profile between a peak and an adjacent valley is translated into three segments, namely a circular arc, a straight line, and a circular arc, in such a way as to preserve the height and location of the peak, the depth and location of the valley, the slope at the midpoint between the peak and valley, and a smooth interface between all segments. At the input end of the RFQ 34, the radial matching section is blended smoothly into the radial cut forming the end of the vane tip. At the output end of the RFQ, a circular arc, of one-centimeter radius, is appended to each vane, blending smoothly with the radial cut forming the end of the vane tip. The constant vane-tip-radius design allows the use of a special shaped cutter for contouring the vane tips, which greatly reduces the cost of the vane-tip machining. As is known to those skilled in RFQ design, the radius of this cutter must come from the geometrical details of the vane-tip profile itself. The constraint is simply that the tool radius must be smaller than the minimum concaved radius of the vane-tip profile. The interior surface of the vane housing and the majority of the vane surfaces are copper plated (UBAC-R1 process) for electrical conductivity. The vane tips are left unplated as a precaution against possible problems with copper plating in the region of high field and critical geometry. The exterior of the vane housing and flanges are anodized black to provide a smooth stable surface for precision alignment measurements. The RFQ assembly process starts with the installation of the 48 micrometer-thread pushing screws of the assemblies 86 that form the alignment surfaces and the 24 locking plates that restrict their motion. The pushing screws are initially set to their nominal position relative to the flats on the exterior surface of the vane housing. The vanes 82 are installed to their nominal positions, one at a time, in any order. They may be aligned as they are installed or the alignment may be postponed until several or all have been installed. After the vanes are installed, the position of the vanes is adjusted by moving the pushing and pulling screws to achieve the desired gap spacing. The counteracting forces from the pushing and pulling screws keeps the vane position under positive control and contributes to the alignment accuracy achievable from this design. Advantageously, all of the measurements required to align a vane, or to check its alignment, can be made at any time without regard to the status of the other vanes. The primary reference for all alignment measurements are the four flat surfaces 88 accurately machined on the outer surface of the vane housing. The vane alignment is based on depth-micrometer measurements from these flats through holes in the housing and the vanes, to selected flat portions of the vanes. Referring for a moment back to FIG. 3, the rf power system 48 provides the power that accelerates the .sup.3 He.sup.++ beam to the desired energy level. As indicated above, the RFQ is configured as two 1.7-m-long sections in tandem. Each of these sections requires 200 kw of rf power (peak). The power for each section is supplied by 8 small planar triodes 81 mounted directly on the RFQ cavity wall inside the RFQ vacuum enclosure. The 8 tubes are mounted in pairs on each of the four quadrants of the structure as shown in FIGS. 5B and 5C. Each pair is driven in parallel by one input cavity resonator 83. This close-coupled scheme offers many advantages over conventional rf power systems. For example, the close-coupled scheme: (1) eliminates the need for separate rf output cavities for each power source; (2) eliminates the need for transmission lines between each power source and the linac; (3) eliminates the need for high-power rf windows for each transmission line; (4) replaces the conventional rf drive loop with an integrated drive loop for each power source or cluster of power sources; and (5) provides a convenient, rigid, mechanical support for each power source. Suitable planar triodes are commercially available from, for example, Eimac Corporation of Salt Lake City, Utah. The Eimac planar triodes (Models Y-690, YU-141, YU176) produce 30 kW of rf power with a 2% rf duty factor and an efficiency of 60%. They are small in size and relatively low in cost. Further advantages provided by powering the linacs with a multiplicity of smaller power units exist. For example, it is relatively easy to survive the failure of any one unit by calling on some reserve power from the remaining units. Also, the system hardware, being small in size and large in number, results in favorable design and fabrication costs. As is known to those skilled in the art, the planar triode operates well in a "grounded grid" configuration. This implies that the anode and the loop operate at an elevated potential (6-8 kV) and should have considerable capacitance to ground (200 pf or more). Using the required electrical insulation as the dielectric of the required rf bypass capacitor results in a compact and rigid configuration. The anode cooling water enters the anode bypass capacitor ring, passes through the loop to the anode cap, and then back through the loop and capacitor ring on the way out. Each cluster of triodes requires a grid/cathode circuit, typically involving a resonant input cavity. The configuration shown in FIGS. 5D and 5E involves a three-quarter wavelength coaxial cavity with the outer conductor grounded, a tuning stub at the far end, and the open end of the center conductor connected to the cathode. The four input cavity resonators on each section are driven in-phase through a four-way power splitter and equal-length lines. In summary, close-coupled, loop-drive, rf power sources, using the linac resonator itself as their output resonator and power combiner, offer substantial savings in the cost, complexity, weight and efficiency of rf power sources for linac applications. All problems associated with the extraction of the rf power from the power source, transmission of the rf power to the linac, and the injection of the rf power into the linac are solved, in the simplest way, by the close-coupled configuration. The system control is further simplified by eliminating concerns over reflected power and standing waves in the non-existent transmission lines. Turning now to the control aspects of the present invention, and referring back to FIG. 3 momentarily, it is seen that the control system 18 includes a control processor 78 and a plurality of Programmed Logic Controllers (PLC's) 68 that interface with a conventional keyboard 70, a CRT 72, and a printer 74. (In FIG. 3, the keyboard, CRT, and printer are shown as interfacing with the PLC 68. However, it is to be understood that these devices may interface directly with the processor 78.) Essentially, the PLC's 68 include a programmed microprocessor, or equivalent device, that is programmed in a specified manner so as to perform a desired function. From an operator point-of-view, for example, the accelerator system has three states: "standby", "ready", and "run". Transitions between these states is essentially a push-button operation. The transition from "standby" to "ready" involves approximately a five minute delay for component warm-up. The other transitions are essentially instantaneous. From a system point-of-view, however, the control system handles all of the automated tasks of closed loop and logic control. A system timer 76 augments the operation of the PLC 68 by generating the controlled time signals that are used in the pulsed RFQ system. The system timer 76 is discussed in more detail below in connection with FIG. 6. In general, the control system provides the following automated functions: system startup, with proper warm-up periods (5 minutes from a cold start), and component monitoring; run programming, including target selection, duration of irradiation, and logging with hard copy printout; continuous monitoring of RFQ operating parameters, with appropriate protective interlocks or warnings; color CRT display of operating parameter, interlock status, and irradiation parameters; and fault finding guides to locate malfunctions rapidly and simply. The computer or processor 78 provides the system 12 with all the control instructions and also monitors the important parameters for the processing of the precursors. The software and hardware for controlling the targetry system 16, including the precursor units 42, is provided with the commercially available targetry systems. Other software for controlling the accelerator 14 can be readily incorporated into this commercially available equipment by those skilled in the art in order to provide a user friendly, hospital-proven control system for a clinical environment. Because the RFQ-based accelerator is a pulsed system, a synchronizing clock signal must be distributed to all pulsed subsystems. To this end, a system timer 76 is used to generate the appropriate synchronized signals. A block diagram of the system timer 76 is shown in FIG. 6. The basic pulse rate of the accelerator is 120 Hz and is phase locked to the incoming AC power at trigger generator 102. The resulting beam pulse is 83 microseconds long. Pulses to the individual support subsystems are delayed up to 1000 .mu.sec as required for timing of the support subsystems using variable delay circuits 104-109. Pulse gates 110-115, also variable up to 1000 .mu.sec, are connected in tandem to the variable delay circuits 104-109, and drive the individual subsystems. The subsystems that require these timing pulses are the ion source 30, the low energy beam transport rf system 66, the RFQ rf system 48, and the simultaneous four target option system (FIG. 4). An oscilloscope, used to measure the system pulsed parameters, including the beam current, also receives timing pulses. One or more sample and hold circuits (not shown) may also receive these timing pulses. Such sample and hold circuits are used primarily to facilitate the measuring of other pulsed signals, especially when the results of the measurement are to be displayed on a suitable display device included in the console. The delays and widths associated with the timing pulses are set by the operator through the control system. The delay circuits 104-109 and the gates 110-115 are easily implemented by those skilled in the art using analog and/or digital commercially available components. Referring next to FIG. 7, an elementary diagram of the vacuum system 44 is shown. Vacuum systems are, of course, known in the art. The description that follows is presented simply to illustrate the best mode in which known vacuum system components could be combined to serve the purposes of the present invention. Vacuum pumping is accomplished by two turbomolecular vacuum pumps 120 and 122, each connected to the vacuum enclosure. One pump is in the Ion Source/LEBT end of the enclosure and the other is in the RFQ end. The required pressure in the LEBT region is 10.sup.-5 Torr, or less during operation. In the RFQ area, the required pressure is 10.sup.-6 Torr, or less. These pressures are met with the two turbomolecular vacuum pumps 120, 122 each with a capacity of 450 liter/sec (385 liter/sec in hydrogen). The two turbomolecular pumps and the vacuum enclosure are roughed by a single rotary-vane mechanical pump 124. Advantageously, the turbo pumps provide long term, reliable operation, requiring little maintenance. Cryogenic pumps may also be used, but it is believed that they would not offer the maintenance free operation provided by the turbo pumps. The pumps are controlled and monitored through the control system 18. The pressure in the vacuum enclosure is also measured with both thermocouple and ion gauges. The details of operating and maintaining the vacuum system 44 are conventional, and are known to those skilled in the art. Referring next to FIG. 8, an elementary diagram of the thermal system 46 is shown. Like the vacuum system, thermal systems are also known in the art. The description that follows is presented simply to illustrate the best mode of such a thermal system used with the present invention. A thermal system is required because several subsystems of the accelerator produce heat which must be removed. The function of the thermal system is to circulate low conductivity water through the components and remove the heat from the water by a water-to-air heat exchanger 128. To this end, the thermal system includes a primary pump 130 that pumps water from a storage tank 128 (at a rate of about 6 gallons per minute) through the water-to-air heat exchanger 132, through a filter 134, through one of three parallel paths (the ion source path, the vacuum system path, or the RFQ path), and back to the tank 128. The RFQ path is most critical because the temperature rise of the vanes 82 must be tightly controlled. To keep the distortion of the vanes to a minimum, including the vane-to-vane spacing, the allowable temperature rise and variation of the coolant in the vanes should not exceed one degree Centigrade. To this end water flows through the four vanes 87 (parallel connected) and returns through copper tubes 136 that have been thermally bonded to each quadrant of the vane housing. Because of the direct contact of the water with the vanes, the temperature of the water is an accurate indication of the vane temperature. The temperature is stabilized by a temperature controlled feedback loop that includes a secondary pump 138 for recirculating the water back through the vanes 82. This loop further includes a temperature controller 140 coupled to a solenoid valve 142 which allows water from the heat exchanger 132 to be mixed with the RFQ water so as to maintain a constant temperature. In the ion source path, it is estimated that 1100 W of power is dissipated in the ion source 30. To keep the temperature rise to less than two degrees Centigrade, about 3 gpm (gallons per minute) of cooling water is required. The vacuum system path, on the other hand, requires much less cooling, and only about 0.1 gpm of water is required. The thermal system pump 130 is designed to produce a differential pressure of 40 psi (pounds per square inch) at a flow rate of approximately 6.1 gpm. The heated water from the pump, including the heat from the loads, passes through the water-to-air heat exchanger where a blower 144 moves 400 CFM (cubic feet per minute) of ambient air through the heat exchanger fins, thereby removing the heat from the water. Referring next to FIG. 9, a basic flow chart illustrating the method of obtaining suitable radiopharmaceuticals for PET applications in accordance with the present invention is depicted. This method is preferably carried out automatically by the control system 18; but it could also be carried out one step at a time, with each step being initialized manually. The method includes the steps of: (1) obtaining low energy .sup.3 He.sup.++ ions from a suitable source (block 150); (2) focusing these low energy ions into a beam and transporting this beam to the input port of an RFQ linac (block 160); (3) accelerating the beam using the RFQ linac to an energy of around 8.0 MeV (block 170); (4) transporting or otherwise directing the high energy beam into a target system (block 180); (5) irradiating a suitable target material with the high energy beam to produce radionuclides of interest (block 190); (6) preparing suitable precursors from the radionuclides (block 200) that can be used in (10) preparing desired radiopharmaceuticals (block 210) that have application to PET. Should it be desired to test or calibrate the system without directing the high energy beam to a target material (block 172), then the beam is directed to a suitable beam dump (block 174), and the desired measurements or calibration steps are performed (block 176). The irradiating step includes moving the proper target into position using the target handling system (block 178), and then directing the high energy beam to the target (block 180). Advantageously, the step of preparing precursors having application to PET (block 200) may include automatically and programmably collecting the radionuclides resulting from irradiation of the target(s) (block 202), and automatically processing the same to produce the precursors of interest (block 204). A major advantage of the .sup.3 He.sup.++ RFQ utilized by the present invention is that it is extremely light weight in comparison to a cyclotron (<0.5 tons compared to approximately 20 tons), yet the RFQ-based system can nevertheless produce the radioisotopes of interest (.sup.18 F, .sup.13 N, .sup.15 O, and .sup.11 C) in more than adequate quantities. The radioisotope .sup.18 F is produced particularly copiously. Moreover, the .sup.3 He.sup.++ target reactions have the property that fewer neutrons are produced per isotope nucleus than with low energy proton or deuteron based systems. This fact, coupled with the fact that helium-3 causes almost no neutron production in collisions with the accelerating structure, results in the elimination of the radiation shielding for the accelerator and a factor of nine reduction in total facility shielding weight (including the vault) compared to a proton/deuteron cyclotron facility. Moreover, the natural exit of the beam from the linear structure of the RFQ, as opposed to the forced extraction from the circular cyclotron, also provides the additional advantage that component activation is minimized. Further, no enriched target materials are required. A single beam particle type can be used to produce all four isotopes, therefore avoiding particle switching. The entire system can further operate using approximately 20 kW of power, only about 20% of the power consumption for present cyclotron facilities. Finally, the RFQ beam cross section is circular, instead of the strongly elliptical shape from a cyclotron, thereby leading to better beam utilization in cylindrical targets. Advantageously, the order of magnitude reduction in facility weight, the virtual elimination of the accelerator weight, and the relative lack of activated components, gives rise to the possibility of a transportable radiopharmaceutical production system. Such a transportable system is illustrated in FIG. 10, wherein the entire radiopharmaceutical production facility 12 is installed in a tailer 222 of a conventional 18-wheel truck transport 220. Other suitable forms of transport, of course, could also be used, such as a railway car, or ship. A transportable system such as is shown in FIG. 10 makes the PET technique far more accessible geographically and financially than has heretofore been the case, thus representing a true advance in the PET technology art. While the invention herein disclosed has been described by means of specific embodiments and applications thereof, numerous modifications and variations could be made thereto by those skilled in the art without departing from the spirit and scope thereof. Accordingly, it is therefore to be understood that within the scope of the appended claims, the invention may be practiced otherwise than as specifically described herein. ##SPC1## |
058698418 | description | DETAILED DESCRIPTION OF THE INVENTION The present invention provides a method and a device for imaging sources of x-ray and gamma-ray radiation. The device, designated generally as numeral 10 in FIG. 1, incorporates a plurality of lens/detector assemblies 17 to first focus and then detect radiation emanating from a radioactive source 15, such as a tumor in a patient 12 that has incorporated some radioactivity as it grows. Each lens/detector assembly 17 comprises a plurality of high efficiency and high resolution crystal diffraction lenses 18 that focus onto detectors 19 only the radiation of a desired energy and origin. As disclosed infra, and with reference to FIG. 6, each lens 18 comprises a plurality of concentric rings 45, which in turn are comprised of very accurately mounted diffracting crystals. These crystals are oriented so that only radiation having a predetermined energy is focused onto the detector 19. The detectors 19 of the device are shielded from unwanted radiation. The device 10 is designed to accommodate the detection of radiation from a myriad of sources. For clarity, the radiation source 15 in the exemplary embodiment shown in FIG. 1 is a tumor that has absorbed a radio-isotope in vivo, whereby the tumor emits radiation of a predetermined wave length .lambda.. However, other radiation sources are also appropriate, including radioisotope-impregnated fissures in a mineral or in a manufactured object, an x-ray or gamma-ray beam scattering from a target, x-rays or gamma rays produced by particle-beam bombardment of a target, and high metabolic rate regions in a living organism wherein a radioisotope has been incorporated. After emanating from the source 15, the radiation is subjected to a means for focussing the radiation, such as the lens, 18. The lens 18 directs the radiation to a detection device 19. The output of the detector is analyzed by a computer. The exemplary device 10 is a plane circular array of lens/detector assemblies 17 with the source 15 situated at the center 13 of the array, the detectors 19 positioned along the periphery of the array, and the focussing means 18 positioned approximately medially between the source 15 and the detectors 19. As noted supra, the detectors 19 comprise the periphery of the plane circular array and therefore are distally placed relative to the center 13 of the circular array and the focussing means 18. A three-dimensional scan of the source 15 can be accomplished with two lens/detector assemblies 17. FIG. 2 is an exemplary embodiment of a three-dimensional imaging system comprising two intersecting and concentric orthogonal arrays 10 of the lens/detector assemblies. The radiation source 15, resting on a movable platform 16, is located at the intersection of the two arrays at their common center 13 at the time of imaging. Prior to high resolution imaging operations, conventional scintillation counters 20 are provided for quick scan capabilities of the radiating area to approximately locate the source's position. For the sake of additional clarity, FIG. 1 is an elevational view of FIG. 2 taken along lines 1--1. If the present invention is used as a medical imaging system, then the source 15 is a patient in whom a radioisotope has been injected. A reference source 14 of the same isotope is positioned at a suitable point on the patient's body and the location of the patient's tumor is measured with respect to the reference source 14. Imaging of an extended source is best accomplished by moving the movable platform 16 across the center 13 of the intersecting arrays 10. Alternatively, one could move the lens system relative to the source if means have been provided therefor. The positions of the lenses, detectors, and a platform 16 containing the source 15 and the reference source 14 are monitored by conventional electronic sensors (not shown) and recorded and analyzed by a computer (not shown). Lens/Detector Assembly Detail Each lens/detector assembly 17 incorporates a plurality of movable focussing means (such as lenses 18) and detectors 19. The positions of the lenses, detectors and a platform 16 containing the source 15 are monitored by conventional electronic sensors (not shown) and recorded and analyzed by computer (not shown). FIG. 3 is a cross sectional view of FIG. 1 taken along lines 3--3 and presents a detailed depiction of the lens/detector assembly 17. Each lens/detector assembly 17 incorporates a plurality of movable focussing means (such as the lenses 18), detectors 19, and shielding around the detectors 19. Shielding is also placed along the longitudinal axis 23 of the assembly and longitudinally along the outside of the assembly. The axis and outside radiation shields 29, 30 respectively, are cone-shaped and mounted between the lens 18 and the source 15 and the lens and the detector 19. Generally, the axis and outside shields can be any convenient configuration such as cone- or cylindrically-shaped. Lead, iron, and brass are suitable shielding materials. S and D denote the lens-source, and the lens-detector distances respectively. Lenses and detectors are mounted on tracks 22 equipped with electronic sensors. The tracks allow for independent axial movement of either or both the lens 18 and detector 19: Typically, the detector is moved in the same direction but twice as far as the lens. Generally, the detector 19 comprises a sodium iodide crystal, a zinc sulfide crystal, anthracene, or some other substance or combination of substances that scintillates when contacted with ionizing radiation. Also comprising the detector is a photomultiplier tube to monitor the scintillations as they occur. Optionally, a 2.times.2 or 3.times.3 detector array is also suitable to enable a determination as to whether the source being imaged is on the axis of the lens or off the axis of the lens and if off-axis, to determine in which direction it is off-axis. In the case of the 2.times.2 array, the source is on axis when the counting rate in all four segments is equal. In the 3.times.3 array, is source is on axis when most of the radiation interacts with the central detector and the other detectors have equally weak count rates. The 3.times.3 array can also be used to obtain the lowest background possible. If the center detector is large enough to intercept all of the focused radiation when the source is on axis, then one needs to consider only the background in the center detector. Furthermore, an energy sum coincidence can be made between the center detector and the outside detectors that can increase the efficiency for detecting the full energy of the gamma ray, thus increasing the full energy count rate without increasing the background count rate. Thus, one has the efficiency of a large detector for detecting the full energy of the gamma ray, while retaining the low background counting rate of only the central detector. The detector could also be a solid state detector made of silicon or germanium or a composite material such as CdTe. When radiation is absorbed by these detectors, positive and negative charges are generated that can be collected and measured with suitable electronics. These detectors have much better energy resolution and thus lower background counting rates. This would allow one to detect weaker sources. These detectors are, however, much more expensive. In order to focus x-ray and gamma radiation, the present invention utilizes the phenomenon of crystal diffraction which is illustrated in FIG. 4. FIG. 4a depicts the phenomenon known as Laue diffraction. The incident radiation beam 31 enters through one surface of a diffracting crystal. After interacting with a specific array of parallel atomic layers 34, the radiation beam is split into two beams, a transmitted beam 32, and a diffracted beam 33, with both beams exiting through a surface opposite to the one through which the radiation entered. Both the transmitted and the diffracted beams are produced by a coherent superposition of scatterings by atoms in the parallel crystal layers. The angle 35 between the radiation beam and the crystal layers is designated as p. The maximum fraction diffracted by crystals with some mosaic structure using Laue diffraction is 50%, with the remaining fraction being transmitted without deflection. Typically between 10.sup.4 and 10.sup.7 atomic layers are suitable to approach 50% diffraction. The actual number of layers depends on the wavelength of the gamma rays and the width of the mosaic structure of the crystal. In practice, the maximum diffracted beam is less than 50% because some absorption of the beam occurs as it passes through the crystal. FIG. 4b depicts the phenomenon known as Bragg diffraction acting upon an incident beam 131. After multiple scatterings with the atoms comprising a specific array of parallel atomic layers 134, the net outcome is the emergence of a "diffracted" beam 133, which contains nearly all of the incident energy. Some absorption of the radiation occurs during this process which continues until either the radiation is diffracted out of the crystal or is absorbed in the crystal. The angle 135 between the radiation beam and the crystal layers is designated as p. The diffracted beam exits through the same surface as the one through which the radiation entered. Again, the beam is produced by a coherent superposition of scatterings by atoms in the parallel crystal layers. Bragg diffraction is most effective for energies below 200 keV and the fraction diffracted can reach 90%. For both Laue and Bragg diffraction, diffraction occurs only when the Bragg condition is obeyed, (equation 1): EQU .lambda.=2d.sub.hkl sin p (1) where .lambda. is the radiation wavelength, d.sub.hkl the spacing between the atomic layers indicated by the Miller indices h,k,l and p the angle between the direction of the radiation beam and the atomic layers (one can convert energy E in keV to wavelength .lambda. in Angstrom units by using the relation .lambda.=12.397/E). With perfectly parallel atomic layers, only rays within a few arc seconds of p will be diffracted (i.e., the "acceptance angle" is only a few seconds of arc), so that one can obtain a large diffraction efficiency only if the rays are nearly parallel, i.e. only if the source is very far away. As seen in FIG. 5, this problem can be overcome (i.e. the acceptance angle can be increased). FIG. 5a shows that for Laue diffraction, if imperfections are either naturally present or else artificially introduced within the crystal so that all the crystal planes are no longer parallel to each other, rays coming at different angles 39 will still find planes 40 for which the Bragg condition is obeyed. As seen in FIG. 5a the imperfections in the crystal give rise to a three dimensional mosaic structure. The angle 41 between the rays 35 with the lowest angle p and those 39 with the largest p is the acceptance angle. FIG. 5b shows that for Bragg diffraction the acceptance angle can be increased if the crystal is curved in the direction of the radiation beam. Rays coming at different angles 139 will still find planes 140 for which the Bragg condition will be obeyed. The angle 141 between the rays 135 with the lowest angle p and the rays 139 with the largest p is the acceptance angle. The curved shape of the crystals produces a significant focusing effect. The highest degree of focusing for Bragg diffraction occurs when the radius of curvature is equal to L / sin p , where L is the distance from the source to the lens. Furthermore, a mosaic structure in the crystal produces an increase in the acceptance angle in the same manner as described above for Laue diffraction. Each crystal diffraction lens 18 utilizes a plurality of diffracting crystals. Possible crystalline materials include, but are not limited to, silicon, quartz, tin, molybdenum, germanium, and copper. FIG. 6 is a view of FIG. 3 along lines 6--6 depicting a typical embodiment of a lens 18 in the Laue diffraction configuration. Each lens 18 comprises a support substrate 43 typically a metal plate. Stainless steel, brass, tungsten, and aluminum are suitable materials for the substrate 43 with stainless steel, brass, and tungsten having the advantage of better shielding the detector from radiation that was not diffracted by the crystals 42. Regions of the surface of the plate 43 define a series of apertures 44 arranged as concentric rings 45. Each ring contains a plurality of diffracting crystals 42 of the same material and orientation. The material and orientation are determined according to the procedure described below. The innermost ring has a diameter of about 2.7 cm and the outermost ring has a diameter of about 11.6 cm. FIG. 7 is a view of FIG. 3 along lines 6--6 and depicts a typical embodiment for lens 18 in the Bragg configuration. The significant difference is that the curved apertures 125 are much wider than the corresponding apertures 44 in FIG. 6a. For both Laue and Bragg diffraction, the diffracting crystals are mounted onto the plate in such a manner that once mounted, all the crystals in a ring will be so oriented as to use the same set atomic layers to satisfy the Bragg condition. In a typical embodiment, the crystals in a given ring are all of the same material but crystals in different rings may be of different materials. The first step in determining the material and orientation of the diffracting crystals is to select the energy of the radiation that will be observed and the focal length F of the focussing means 18 one wants to achieve. In the simplest embodiment of the invention, a single lens is utilized, in a lens/detector array 17, but a lens/detector assembly 17 having a plurality of lenses is also suitable. Where lenses of focal length F1, F2, F3, etc . . . are placed in close proximity or contact with each-other, the focal length of the combination is given by equations 2 through 6. Equation 2 gives the focal length for one lens, where p is the Bragg angle used in the lens and R is the radius of the crystal ring. EQU F=R/(tan 2p) (2) Equation 3 gives the focal length for two lens, where p.sub.1 and p.sub.2 are the Bragg angles used in the first and second lenses and R.sub.1 and R.sub.2 are the radii used in the first and second lens, respectively. EQU F.sub.12 =(R.sub.1 -R.sub.2)/tan 2p.sub.1 +R.sub.2 /tan (2p.sub.1 +2p2)(3) Equation 4 gives the focal length for three lenses, where p.sub.1, p.sub.2 and p.sub.3 are the Bragg angles used in the first and second and third lenses and R.sub.1, R.sub.2 and R.sub.3 are the radii used in the first, second and third lenses, respectively. EQU F.sub.123 =(R.sub.1 -R.sub.2)/tan 2p.sub.1 +(R.sub.2 -R.sub.3)/tan (2p.sub.1 +2p.sub.2)+R.sub.3 /tan (2p.sub.1 +2p.sub.2 +2p.sub.3)(4) If the lenses are very close together, then the R's become approximately equal and the approximate formula for the focal length is given by equation 5. EQU F.sub.12 . . . n=R(Ave)/tan (2p1+2p.sub.2 +. . . +2pn) (5) If all of the Bragg angles are quite small, the focal length can be approximated by equation 6 EQU 1/F.sub.12 . . . n=1/F.sub.1 +1/F.sub.2 +. . . +1/F.sub.n (6) The set of atomic layers to be used for each ring 45 is determined by the condition that all the rings must have the same focal length F. For rays near the lens axis (small p) the relation between lens-source distance S, lens-detector distance D, and focal length F is given approximately by equation 7. EQU (1/F=(1/S)=(1/D) (7) In practice S and D as shown in FIG. 3 are both chosen to be 2F and the image formed onto the detector is about the same size as the source if the source is larger than the crystal elements in the crystal. If the crystals are bigger than the source, the image will be about twice the size of the crystals. Then the Bragg angle p is arctan R/(2F) where R is the radius of the ring. The Bragg condition yields the relation between the ring radius, focal length, radiation wavelength .lambda., and atomic layer spacing d, given by equation (8). EQU R/F=tan 2 arcsin(.lambda./2d.sub.hkl)! (8) For F>>R, i.e., for small angles, Equation 8 yields EQU d.sub.hkl =.lambda.F/R (9) In practice, a gamma ray with a specific energy (and therefore wavelength .lambda.) is selected. Then, the crystalline plane spacings of an available crystal are tabulated. This information is combined with the desired focal length F to arrive at the respective radii R for the crystal rings, pursuant to equation 10: EQU R=d.sub.hkl /.lambda.F (10) Finally, the size of the crystals are chosen. Alternately, .lambda. is determined from the desired gamma ray energy, then F is chosen, and the available values of dhkl are identified, so that the values of R for the rings are suitable. Copper and germanium are suitable for radiation energies above 100 keV. Lower atomic number materials such as quartz, silicon, and beryllium are more suitable for low energy gamma rays (below 100 keV.) Laue Diffraction Lens In a preferred embodiment for a Laue diffraction lens, copper crystals grown at and obtainable from a facility such as the Institut Langevin-Langmuir (ILL) in Grenoble, France, are utilized. Copper crystals naturally exhibit enough imperfections in their crystal lattice so that their acceptance angle is of the order 200 to 500 seconds of arc, i.e. between 0.06 and 0.15 degrees. Heating and then compressing copper crystals increases the acceptance angle even further. Referring to FIG. 8a, ILL typically provides cylindrical copper crystals 51 of 10 cm. in diameter and 25 cm. long, with a predetermined crystal orientation. Thin slabs 53, of 2 to 3 mm thickness 55, are cut parallel to the planes 57 designated by the Miller indices that have been selected. As shown in FIG. 8b, the slabs 53 are then cut in turn into crystals 42 with faces 59 approximately 2 mm square. The faces 59 are perpendicular to the planes 57. Mounting of the crystals 42 onto the plate 43 can occur in a variety of ways. One way is to first place the plate 43 against a rigid flat surface so that the flat surface is accessible though the concentric ring apertures 45. A number of crystals 42 are placed on the flat surfaces within the confines of the ring 45 with the face 59 of the crystals 42 (i.e., the face that corresponds to the plane selected as described supra) flush against the rigid flat surface. Enough crystals are placed in the ring to virtually fill the ring aperture. The crystals are then cemented together. Upon completion of the mounting procedure, the face 59 of the crystals that is perpendicular to the planes 57 whose Miller indices have been selected is perpendicular to the lens/detector assembly axis 23. (see FIG. 3) The area of this crystal face (2 mm square is suitable, as suggested supra) and the width of the mosaic structure determine the ultimate size of the image spot at the detector location 19. Crystal face 59 surface areas as small as 1 mm square are suitable. Copper and Germanium crystals are suitable for radiation energies above 100 keV. Lower atomic number materials such as quartz, silicon, and beryllium are more suitable for low energy gamma rays, i.e., below 100 keV. FIG. 9a is a detailed view of one of the curved apertures 44 shown in FIG. 6 that contain the crystals 42. As shown, the crystals can be large enough to entirely fill the face of the aperture 44 or, if smaller than the aperture, they can be stacked on top of each other. In either configuration, once in place the crystals are then cemented into the containing means 44. In the instant embodiment, the containing means is the curved apertures. The face 59 of the crystals that is perpendicular to the planes 57 whose Miller indices have been selected is parallel to the plane of the substrate 43. Alternatively, as shown in FIG. 9b, one may cut the crystals into thin strips 61 having a length 63 of perhaps 1 to 20 cm. The strips 61 are arranged in stacks 66 of a predetermined height 62 (for example 1 mm to 4 mm in height), and then bent into circular arc sections of the same radius as the ring 45 to be mounted therein. This procedure is more suitable when the crystalline materials are malleable. As an option, and as depicted in FIG. 9c, one may increase the fraction of the radiation that is diffracted onto the detector if the crystals 42 are cut with a wedge-like cross section and then stacked in the aperture 44 so that radiation 41 incident on each crystal is diffracted onto the detector. This wedge construction essentially increases the solid angle that the crystals in the lens can cover. As such, lens efficiency is concomitantly increased with a 2- or 3-fold increase in intensity of the diffracted beam. Generally, crystals of malleable materials (e.g. copper, molybdenum and tin) or crystals of materials with low melting points are especially suitable in that they exhibit a high degree of mosaic structure. With suitable treatment, however, many other crystal types (and not just those from malleable elements) can be made to exhibit mosaic structure resulting in acceptance angels of 200 to 500 seconds of arc. Methods for introducing such mosaic structure include neutron irradiation, heating the crystal to near its melting point and then subjecting it to stresses or compression, subjecting the crystal to mechanical vibrations (e.g. sonic vibrations), and introducing impurities (i.e. dopants). Generally, the higher the atomic number of the material, the more efficient it is for diffraction of high energy gamma-rays. FIG. 10 is a cross-sectional view of a Laue diffraction lens 18 taken along lines 10--10 in FIG. 6. Bragg Diffraction Lens A wide variety of crystals are suitable for a Bragg diffraction. This is because the diffracted beam does not have to pass completely through the crystal and be reduced in intensity by absorption of the full thickness of the crystal. The diffraction efficiency for Bragg diffraction is determined by the ratio of the diffraction coefficient per unit length to the absorption coefficient per unit length. Since the ratio of these two quantities remains nearly the same for low gamma ray energies, the diffraction efficiency does not change as dramatically with changing energy as it does in the Laue diffraction case. For Bragg diffraction with bent crystals 42 (see FIG. 12 and FIG. 13), a large mosaic structure is not required in order to achieve a large acceptance angle. All that is required is that one be able to cut the crystals to form bendable strips of suitable dimensions. Exemplary dimensions are strips 1 to 2 mm wide, 0.5 to 2 mm thick, and 2 to 20 cm long. The crystal strips are then bent to a radius of 1 m or more, perpendicular to the crystal's long axis. The needed radius of curvature is equal to the distance from the source to the lens divided by sin p. The typical length of a strip is given by the width of the aperture 125 divided by sin p. In the case of the bent Bragg crystals, the width of the mosaic structure controls the size of the field of view of the lens. Thus, one can adjust the size of the field of view independently of the size of the solid angle subtended by the crystals. Crystalline planes are selected and crystal strips are cut in much the same way as described in conjunction with Laue diffraction. FIG. 11 is a three dimensional view of a Bragg lens. The Bragg crystals 42 are mounted on the concave surfaces of a plurality of coaxial cylindrical supports 159. FIG. 12 is a cross-sectional view of FIG. 11 along lines 12--12, and FIG. 13 is a detailed view of FIG. 12. FIG. 13 depicts how the crystals 42 are mounted on the supports 159. Said supports 159 are in turn mounted in a substrate 43 containing apertures 125 arranged as concentric rings 45. As can be seen in FIG. 13, the apertures 125 corresponding to each ring 45 are much wider than the crystal thickness 150 in order to allow the radiation beam 153 to impact upon all of the crystal face 156. The supports 159 that are provided for the bent crystals are shaped so that the radius of the support surface 162 matches that of the bent crystal. One such support can be a machined surface integrally molded, or removably attached to the substrate 43. The use of bent Bragg crystals in the lens allows one to focus the diffracted beam from an individual crystal into a narrow line parallel to the diffraction planes and on the axis 23 of the assembly 17. This concentrates the diffracted beams from the full lens and makes it possible to use a smaller detector in the focal plane. The length 162 of the crystals 42 and the supports 159 is 2 to 20 cm in the direction of the beam 153. Generally, the longer crystals are closest to the lens axis 23, where the values of sin p are smallest. The length of the crystal strips are adjusted to obtain the maximum diffracted flux. Scanning A Source And Formation Of An Image The lens detector assembly achieves its best performance for sources located on or very near the axis of the assembly. When the source 15 is not situated on the axis of the assembly, the movable platform 16 is advanced until the source is positioned on the axis of the assembly. In order to scan across the source 15, one may change the position of the body 12 using the means provided for moving the table 16. Alternatively, one may change the orientation of the lens/detector assemblies and adjust the source/lens and lens/detector distances as indicated by Equation 3 by means of the tracks 22 on which lenses and detectors are mounted. In yet another alternative, one can move the whole lens system relative to the source. Also, equation 5 shows the focal length's dependence on the wavelength of the radiation. The lens 18 and detector 19 are mounted on tracks 22 allowing the use of a given lens to detect radiation of a different wavelength by adjusting lens-source and lens-detector distances as dictated by equation 3. Electronic sensors are mounted on tracks 22 and their signals are recorded and analyzed by the computer. Instead of relying on tracks 22, imaging of radiation of different wavelengths can also be accomplished by using different lenses, and keeping the elements of the assembly stationary. For example, a source having a first energy can be scanned in toto by moving the Table 16 with respect to the center of the lens/detector arrays (see FIG. 1). If the device is to be used for gamma rays of a second energy, one can construct a plurality of different lenses using crystals with atomic spacings so chosen that one obtains the same focal length as the lenses used to focus the first source. Signals from the detectors 19 are analyzed by a computer in conjunction with the data from the scintillation detectors 20 and those from the sensors on the movable platform 16 and the lens and detector tracks 22. A variety of crystalline materials (Germanium, Silicon, Copper and Quartz) have been found by the applicant to be suitable for the fabrication of x-ray and gamma-ray lenses for energies of around 150 keV. Specifically, a prototype Laue crystal diffraction lens was constructed and tested early in the lens project at Argonne National Laboratory consisting of a ring of quartz crystals cubes, 5 mm.times.5 mm.times.5 mm mounted in a ring with a radius of 18 cm. This lens had a focal length for 140.4 keV gamma rays of 1.58 meters (m). This lens was also tested with the 59.54 keV gamma ray from .sup.241 Am where the focal length was 0.64 m. Tests were made at a series of distances between the lens and the source: 1.28 m, 1.92 m, 2.56 m, 3.20 m, and 3.84 m. These early tests laid the ground work for the development of this technology. While the invention has been described with reference to details of the illustrated embodiment, these details are not intended to limit the scope of the invention as defined in the appended claims. |
abstract | Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions. |
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claims | 1. A two-dimensional radiation detector for a radiographic scanner, the radiation detector comprising:an anti-scatter module;a first aligning means for aligning the anti-scatter module with a spatial focus;a second aligning means for aligning the anti-scatter module;a detector subassembly module, each detector subassembly module including a substrate and an array of detector elements arranged on the substrate to detect radiation, anda radiation absorbing mask formed as a grid and arranged between the array of the detector elements and the anti-scatter module; wherein the second aligning means includes alignment pins that align the anti-scatter module with the detector subassembly modulewherein the second aligning means further includes:alignment openings disposed on the substrate; andalignment openings disposed on the radiation absorbing mask; whereinthe alignment pins are disposed on the anti-scatter module, such that inserting the pins into the radiation absorbing mask alignment openings and the substrate alignment openings aligns the detector element array with the radiation absorbing mask and the anti-scatter module. 2. The radiation detector as set forth in claim 1, wherein the radiation absorbing mask is formed of a radiation absorbing material. 3. The radiation detector as set forth in claim 1, further including:one or more of additional radiation absorbing masks stacked on the alignment pins. 4. The radiation detector as set forth in claim 3, wherein the radiation absorbing mask has stepped edges, which interleave with stepped edges of adjacent radiation absorbing masks. 5. The radiation detector as set forth in claim 1, wherein the anti-scatter module includes:a plurality of anti-scatter vanes formed of a material which is substantially absorbing for radiation produced by the radiographic scanner. 6. The radiation detector as set forth in claim 5, wherein the radiation absorbing mask includes:first strips parallel to the plurality of anti-scatter vanes, which first strips are wider than a thickness of the anti-scatter vanes and are equal or greater than a gap between the detector elements of the array. 7. The radiation detector as set forth in claim 5, wherein the radiation absorbing mask includes:second strips perpendicular to the plurality of anti-scatter vanes, which second strips are of substantially a same dimension as a gap between the detector elements. 8. The radiation detector as set forth in claim 5, wherein the radiation absorbing mask has stepped edges, which interleave with stepped edges of adjacent radiation absorbing masks. 9. The radiation detector as set forth in claim 1, wherein the radiation absorbing mask defines precise apertures, which align with and set a resolution of the elements of the detector array. 10. The radiation detector as set forth in claim 9, wherein the apertures are precisely defined by photochemical etching. 11. The radiation detector as set forth in claim 1, wherein the detector element array includes:a scintillation array that produce scintillation events responsive to radiation produced by the radiographic scanner, wherein the scintillation array includes scintillation elements; anda photodetector element array, each photodetector element of the array being arranged to view one of the scintillation elements of the scintillation array to convert light from the scintillation events into electrical signals. 12. The radiation detector as set forth in claim 11, wherein the scintillation element array is arranged in a two-dimensional rectangular array with a rectangular array of interfaces between adjoining scintillation elements and the radiation absorbing mask includes:a rectangular array of strips of a radiation absorbent material that defines the grid, the strips overlying interfaces between adjacent scintillation elements. 13. The radiation detector as set forth in claim 1, wherein the radiation absorbing mask includes:a first plurality of strips extending along a first direction; anda second plurality of strips extending along a second different direction. 14. A two-dimensional radiation detector for a radiographic scanner, the radiation detector comprising:an anti-scatter module;a first aligning means for aligning the anti-scatter module with a spatial focus;a second aligning means for aligning the anti-scatter module;a detector subassembly module, each detector subassembly module including a substrate and an array of detector elements arranged on the substrate to detect radiation, anda radiation absorbing mask formed as a grid and arranged between the array of the detector elements and the anti-scatter module; wherein the second aligning means includes alignment pins that align the anti-scatter module with the detector subassembly module wherein the radiation absorbing mask includes first alignment openings and the detector subassembly module includes second alignment openings, and the alignment pins extend through the first alignment openings of the radiation absorbing mask and the second alignment openings of the detector subassembly module. 15. The radiation detector as set forth in claim 14, wherein the first direction is perpendicular to the second direction. 16. A radiation detector of a radiographic scanner, the radiation detector includes a plurality of detector modules, each detector module including:an anti-scatter module, including a plurality of vanes and alignment pins; anda rectangular grid including:a plurality of wider strips, arranged parallel to each other, each wider strip being wider than a width of each vane,a plurality of thinner strips, the plurality of thinner strips being arranged perpendicular to the wider strips to form uniform openings, each wider strip is aligned with a corresponding vane. 17. The radiation detector as set forth in claim 16, further including:a detector array including a plurality of detector elements arranged to form a multi-dimensional rectangular array, each two adjoining detector elements of the array being separated by interfaces, the interfaces are aligned with the rectangular grid to place the grid openings between the vanes and the detector elements of the array to define resolution of the radiographic scanner, wherein the detector array includes a substrate with alignment openings, and the alignment pins of the anti-scatter module lie within and extend through the alignment openings in the substrate of the detector array. 18. A computed tomography scanner including:an x-ray source mounted to rotate about an examination region, the x-ray source emitting a cone shaped x-ray beam from a radiation focal point and traversing the examination region;a two-dimensional radiation detector which receives the cone beam of radiation that has traversed the examination region, the radiation detector including a plurality of detector modules, each detector module including:an anti-scatter module, which includes alignment pins,a detector subassembly module aligned with the anti-scatter module, each detector subassembly module including a substrate and an array of detector elements arranged on the substrate to detect radiation, anda radiation absorbing mask formed as a grid, the mask being arranged between and aligned with the array of the detector elements and the anti-scatter module, wherein the alignment pins of the anti-scatter module extend through alignment openings in the mask and alignment openings in the detector subassembly module; anda reconstruction processor for reconstructing signals from the detector element array into a volumetric image. 19. A method for manufacturing a radiation detector for a computed tomography scanner, the method comprising:aligning an anti-scatter module, which includes extending alignment pins, with: a detector subassembly module including a substrate and an array of detector elements arranged on the substrate to detect radiation, and a radiation absorbing mask disposed between the anti-scatter module and the detector elements of the array; andinserting the alignment pins through alignment openings in the mask and alignment openings in the detector subassembly module. 20. The method as set forth in claim 19, further including:forming a radiation absorbing mask by photoetching a radiation opaque material to define a grid. 21. The method as set forth in claim 19, wherein the scanner includes an x-ray source on a rotating gantry that produces a cone of x-rays, which pass through an examination region and strike the radiation detector, the method further including:mounting the anti-scatter module onto the computed tomography scanner, with a spatial focal point of the anti-scatter module being aligned with a focal point of the x-ray source prior to inserting the pins into the alignment openings of the mask and the detector subassembly module. 22. The method as set forth in claim 19, wherein as the pins are inserted in the alignment openings of the radiation absorbing mask, edges of adjacent radiation absorbing masks are interleaved. 23. The method as set forth in claim 19, further including:defining uniform apertures in the radiation absorbing mask to precisely fix an amount of radiation received by each detector element of the array. |
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046613102 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT The invention will be described as applied to a protection system for a nuclear power plant but it should be appreciated that the invention is also applicable to protection systems for other types of complex process apparatus where a reliable, fail-safe protection system with high availability is desired. The nuclear power plant which is depicted schematically in FIG. 1 by the block 1 contains a nuclear reactor 3. The reactor 3 produces through nuclear fission reactions thermal energy which is utilized to generate electricity through a turbine-generator combination (not explicitly shown). Operation of the reactor 3 and its associated components is closely moinitored and the reactor is shut-down, or tripped, if certain carefully selected operating limitations are exceeded. A large number of system parameters are monitored in order to assure that operation remains within the selected limits. In a pressurized water reactor (PWR) these parameters include, for example, such measurements as reactor neutron flux, reactor coolant temperature, pressurizer level and pressure, and steam generator pressure and feedwater flow in addition to status indications such as whether a switch is opened or closed or whether a pump is off or on. Some monitored parameters such as departure from nucleant boiling in the PWR are calculated from measured parameters. The various measured and status parameters are monitored by sensors placed throughout the plant. For reliability, redundant, typically four, sensors are provided for each parameter. These redundant sensors form a set such as the sets 5, 7 and 9 in FIG. 1. One sensor signal from each set of all of the monitored parameters is applied to one of the four Roman numeral identified channels I to IV of the protection system through cabling 19. Where appropriate, the sensor signals are compared with selected set points in processors not shown in FIG. 1 so that all the signals A.sub.1, A.sub.2, to A.sub.n, B.sub.1, B.sub.2 to B.sub.n et cetera applied to the protection system channels are partial trip signals which are high if the mointored parameters is within limits and low if out of limits. The partial trip signals such as A.sub.1 through A.sub.n are logically ORed in each channel. Local bypasses of the individual sensors or global bypasses which bypass the partial trip logic of the channel entirely are also selectively applied within each channel, all in a manner to be discussed in more detail below. The statuses of each partial trip signal, each local bypass, and the global bypass of each channel are communicated to each other channel by electrically isolated, fiber optic, multiplexed data links 13. Based upon all of this information, each channel generates an output 15 which controls the opening and closing of a pair of contactors 17 in the reactor trip switchgear 19. The pairs of contactors 17 in the reactor trip switchgear 19 are arranged in an array such that electric power from a source V will be continuously applied to control rod actuators 21 on the reactor 3 unless at least two paris of contactors 17 are opened by the associated protection channel. This arrangement of switchgear contactors participates in the two out of four voting logic by which the reactor is tripped when two out of four of the protection channels indicate the necessity for a trip, however, as will be seen, logic within the channels ensures that there is a coincidence of parameters generating trip signals in the two channels so that the reactor does not trip on two random signals. The channel logic also modifies the voting logic based upon the bypasses in effect. Removal of power from the control rod actuators 21 permits control rods 23 to drop by gravity down into the reactor 3 where they absorb sufficient neutrons to reduce the reactivity below the critical level thereby shutting down the reactor. Each of the protection channels 11 incorporates a number of basic logic units which in turn each contain a fundamental logic element 25 shown in FIG. 2. The operation of this fundamental logic element is based on flux reversals in tape wound toroidal cores made of "square loop" magnetic material. As shown by the arrows 27 and 29 in FIG. 2, this material can be magnetized in either of two directions. The fundamental logic element 25 consists of a single core, 31, that has three windings: an input winding, 33; a d-c control winding, 35; and an output winding, 37. The winding terminals marked by o in FIG. 2 indicate the polarity of the windings. Under the convention used for indicating polarity, current flowing into the o terminal causes a magnetic flux to be induced in the clockwise direction of the core 27. The hysteresis curve for a typical square loop material is shown in FIG. 3. Three parameters of this curve are important to the operation of the core 31 as a dynamic logic element. Those are the maximum (saturation) magnetic flux, B(m); the residual flux B(r); and the "Coercive force", hc, which is the driving magnetizing force required to switch the core. The fact that a residual flux will remain in the material after the magnetizing force has been removed is paramount to the operation of the device as a logic element. This residual flux, along with the hysteresis of the material, defines two stable states: the clockwise 27 and anti-clockwise 29 flux states shown in FIG. 2. In normal operation, a d-c current flows into the terminal of the winding 35 with sufficient magnitude to saturate the core, 31, in the negative (anti-clockwise) direction 29. This causes the magnetic material of the core to be at operating point a in FIG. 3. When a current pulse is impressed into the terminal of winding 33 with a sufficient magnitude to overcome both the magnetizing force generated by the d-c current in winding 35 and the coercive force, hc, then the core, 31 will switch into the clockwise flux state 27, eventually reaching operating point B. This change of the flux state of the core induces a voltage pulse, e(out), across winding 37 which is a function of the specific magnetic material chosen, the thickness of the tape used to wind the core and the magnitude of the magnetizing force impressed upon the core. The ouptut pulse amplitude, V1, is proportional to the saturation flux, B(m), of the magnetic material and the number of turns in the winding 37, and inversely proportional to the pulse width, or the switching time of the core. When the current pulse is removed from winding 33, the magnetizing force of the d-c current in control winding 35 once again takes over and causes the core to switch back to the operating point B, again causing a voltage pulse to be induced across the output winding 37, but this time of the opposite polarity. This relationship of input to output is shown in FIG. 4. The pulses are impressed into winding 33 in a periodic fashion, thus generating a continuous stream of alternating voltage pulses at the output winding 37. If during this operation, the d-c current in the control winding 35 is removed, then the core will not switch back to operating point A when the current pulse is removed form winding 33, rather, the magnetic material of the core will return to the positive residual state, operating point C in FIG. 3. Subsequent current pulses cause the flux to change only from +B(r) to +B(m), therefore the magnitude of the output pulses at winding 37 is substantially smaller than before when the flux was changing between -B(m) and +B(m). Because the core arrived at point C through point B, the last large output pulse will always be positive. If instead of removing the d-c current in control winding 35, the current pulses in the input winding 33 are removed, two cases must be considered, no current in winding 33 and continuous d-c current in 33. In the first case, the core will be reset to point A by the current in the control winding 35 and a final negative output will be generated. In the second case, the core will always remain at the operating point B and the last output pulse is positive. In either case, the core flux will remain invariant and no further output pulses of any magnitude will be generated. In addition to the magnetic coupling described, there is capacitive coupling between the input and output windings of the core. This capacitance, which is extremely dependent on the specific size and geometry of the core and the windings used but was measured to be 35pF in a sample element, works against the "switching" action of the core because the sharp edges of the input current are differentiated by the capacitance to produce voltage spikes on the output that are not controlled by the d-c current in the control winding. It was found, however, that these spikes were sufficiently narrow, 1 to 2 microseconds, to be discriminated from the wider switching pulse by a simple RC filter. Thus by the operation described above, pulses are gated through the core by d-c current present in winding 35 and blocked if the current in winding 35 is removed. This principle forms the basis for implementing the Reactor Trip Logic System. As mentioned previously, the fundamental logic element 25 of FIG. 2 is a component of basic logic units which form the building blocks for each of the protection channels 11. In such a basic logic unit 39 shown in FIG. 5, the fundamental logic element 25 serves as a switch to which an input stream of current pulses is applied through input winding 33. Current through the control winding 35 is controlled by control circuit 41. When control current is present on winding 35 to core experiences flux reversals and pulses appear across the output winding 37. If there is no current in the control winding, then the pulses are blocked. As explained above, the output pulses from the fundamental logic element 25 are not the same shape as the square input pulses, rather, they are pulses of a given width that are "triggered" by the edges of the input pulse and are either positive or negative depending upon whether the edge was rising or falling. Since it is desirable to retain the exact shape of the input pulse at the output of the switch 25, the ouptut winding 37 is connected to a pulse "discriminator-shaper" 43 that is capable of differentiating between the large output pulses that are present when the switch is closed (current flowing in the control winding) and the very small pulses present when the switch is open, and of reshaping the output to resemble the input. The principle of operation of this stage is that of a "Schmitt Trigger" or bistable with hysteresis. This device is "set" by the positive going pulse and "reset" by the negative. The thresholds provided by the hysteresis provide noise immunity and the ability to act on large pulses while ignoring small pulses. Operational amplifier circuits with positive feedback had been considered to implement this block. These offer the advantage of allowing the switching threshold to be very accurately controlled. However, it was discovered that because of imbalances between the rising and falling slew rates of the operational amplifier used, the input pulse width was not preserved. In other words, there was a small delay associated with turning the op-amp on but no delay in turning it off. The result was that the square input pulse lead to an output pulse of a lesser duty cycle and after a few logic stages, the pulse disappeared altogether. Operational amplifiers with balanced rise and fall times generally require dual power supplies. While such an arrangement could be used, it was felt that this requirement would complicate the design and would unduly affect the reliability of the logic function. Therefore, an alternative circuit based on CMOS logic gates was investigated. This circuit is shown inside the block 43. It is basically two CMOS NOR gates 45 cross coupled to form an R-S flip flop. A pulse at the set input "S" causes the ouptut to go to the high state while a pulse at the reset input "R" causes the output to go low. These pulses are provided by the alternating polarity pulses that appear at the output of the core switch element 25. The two diodes 47 connected from the gate inputs to ground perform the function of referencing the pulses to ground by conducting in the forward direction. In order for this circuit 43 to function properly, two considerations are important. First, the voltage level of the pulses coming from the core 31 must be sufficiently greater than the threshold of the CMOS gates. This threshold is nominally one half of the supply voltage. The second consideration concerns the spikes that are present when the core 31 is off (blocking) due to the capacitive coupling of the windings. To prevent these spikes from switching the gates, a low-pass RC filter 49 is provided at the input to each gate. The time constant of these filters 49 is such as to reduce the amplitude of the spikes while at the same time not detrimentally reducing the amplitude of the wider pulses that appear when the core 31 is undergoing flux reversals. It was decided to set the filter time constant to be approximately one half of the pulse width. The next block of the Basic Logic Unit 39 shown in FIG. 5 is a current amplifier 51. This amplifier must take the output of the discriminator stage and amplify it to provide sufficient current to drive the input coil of the succeeding logic unit. The simplest form of this amplifier is a single transistor connected in a common emitter circuit. While a bipolar transistor could be used, a MOSFET 53 is particularly well suited for this application because it can be directly driven by the CMOS logic gates. The gate voltage required for saturation of the MOSFET 53 is approximately 6 volts, thus this voltage will be taken as the supply voltage of the CMOS gates 45. Therefore, the output winding of the core 31 and the low-pass filter 49 must be designed to provide pulses that are suffciently greater than 3 volts, the nominal threshold of the gates. A zener diode 55 protects the MOSFET 53 from voltage spikes while resistor 57 limits current drawn by any spikes. In some cases, it is required that the microcomputers in the system be able to determine the operating status of the logic units 39. For these cases, a status circuit 59 is provided as shown in FIG. 5. This circuit converts the stream of pulses to a maintained logic level by means of a one-shot multivibrator 61. The one-shot 61 is connected so that it is retriggerable and has a time out of approximately 500 microsecond. This will allow for one pulse in a 5 KHz pulse signal to be "missed" without providing a false indication. A ligth emitting diode 63 provides a visual indication of the status of the logic unit and is turned on when a pulse signal is being generated at the output of the current amplifier 51. An isolated status signal is generated through opto-isolator 65 so that any failures in the reading circuits cannot affect the performance of the logic unit 39. Thus far, the circuitry to reconstruct input current pulses at the output has been discussed. The circuit for the input coil 33 itself simply consists of a current limiting resistor 67 connecting the terminal of the coil to the +15 volt supply with the other terminal being connected to the current sinking output of the previous stage of the pulse generator. The normal operating condition is taken to be the logic state in which a continuous stream of pulses exists and the "tripped" state corresponds to that stream of pulses being blocked or removed (by removing the d-c current from the control coil). A string of basic logic units connected in a "series" fashion will provide a logic "OR" function, i.e., a trip signal at any stage will cause the output of the last stage to be in the logically "tripped" state (no pulses). Any logic function can be represented as a combination of AND's, OR's and inversions. The "AND" function can be implemented by the control coil circuit 41 of the basic logic unit shown in FIG. 5. It was described earlier how the coil 35 effectively "permits" pulses to pass through the magnetic core when there is current flowing in the coil, and "blocks" those pulses when there is no current. To make an "AND" logic, multiple current paths 69, 71 and 73 are provided for the coil current, each of which conducts when the corresponding input is in the false (not tripped) state. Grounding any one or more of the AND inputs in this circuit allows current to flow from the voltage source, through the jumper 75 and current limiting resistor 77 into the terminal of coil 35. Diodes 79 and 81 prevent reverse flow of current between the AND inputs. to provide the inversion function, a path to shunt the current around the coil 35 is shown in FIG. 5. When the INV input is grounded, the current that had been flowing through the control coil 35 instead flows through diode 83 and jumper 85 to the INV input, thus causing the magnetic core 31 to block the input current pulses. To ensure that the current is completely diverted from the coil 35, a 2 volt zener diode 87 is provided to establish the potential of the terminal of coil 35 above ground potential so that the terminal can be pulled down to a lower potential than that of the terminal, even through a diode and transistor. For this purpose, an external jumper must be provided between the on board terminal 89 connected to zener 87 and control path 73. In an alternative arrangement of the control circuit 41, the jumpers 75 and 85 can be replaced by a jumper 91 to allow the voltage for energizing the control coil 35 to come from some external source. In practice, a number of the basic logic units 39, typically four core/amplifier circuits, two of which include status circuits 59, are provided on a single printed circuit (PC) card. All inputs, controls and outputs are brought out to the card edge connectors and connections between the circuits are made on the card cage back plane. FIG. 6 illustrates how the basic logic units 39 are integrated with other equipment to form one of the protection channels 11 shown in FIG. 1. In each channel, a clock generator 91 generates a continuous stream of pulses. These pulses pass by two paths, the trip logic path and the global bypass logic path, to the input of a power converter 93. The pulses may be blocked in either or both of the paths, thus preventing them from arriving at the power converter. In fact, the design is such that only one of the two paths is conducting at any given time, however, both may be blocking. The power converter 93 uses the pulse stream at its inputs to control the reactor trip (RT) switchgear. The design of the converter 93 is such that if there are no pulses at either input, the switchgear is tripped. In normal plant operation, the pulses are routed through the trip logic path and blocked in the global bypass logic leg. If plant conditions require that the reactor be tripped, then the Trip Enable and Global Trip computers 95 and 97 respectively, which also form part of the protection channel, as well as other subsystems of the overall plant protection system, act upon the trip logic path to cause it to block the pulses to the power converter 93, ultimately leading to the tripping of the RT switchgear. During the automatic test of the protection system, and also for maintenance, the pulses are routed through the global bypass logic leg thus rendering the trip logic path incapable of opening the RT switchgear. Interlocks are provided, through the bypass logic, to ensure that sufficient protection capabilities are alway maintained through the unbypassed protection channels. To understand the design of the trip logic, it is helpful to consider the following boolean formula for two-out-of-four (2/4) coincidence logic: EQU L(2/4)[Ai,Bi,Ci,Di]=Ai.multidot.Bi+Ai.multidot.Ci+Ai.multidot.Di+Bi.multido t.Ci+Bi.multidot.Di+Ci.multidot.Di (Equation 1) In this relation, the capital letters represent the partial trip signals for the i-th trip function parameter (e.g. high system pressure) and the ".multidot." and "+" represent the logic AND and OR operators, respectively. If "A" represents the partial trip signal in the channel being considered, then the above equation can be rearranged thusly: EQU L(2/4)[Ai,Bi,Ci,Di]=Ai.multidot.(Bi+Ci+Di)+(Bi.multidot.Ci+Bi.multidot.Di+C i.multidot.Di)=Ai.multidot.Ti+Gi (Equation 2) The latter form of the equation groups the partial trip signals from the other channel sets for convenience. The term T is defined as the Trip Enable signal and the term G is defined as the Global Trip. Since there are multiple functional parameters that can trip the reactor, the overall trip logic function is defined by the equation: EQU RT=A1.multidot.T1+G1+A2.multidot.T2+G2+ . . . +An.multidot.Tn+Gn=A1.multidot.T1+A2.multidot.T2+ . . . +An.multidot.Tn+G(*) Where: G(*) is the logical sum of all Gi (Equation 3). This logic equation 3 is implemented by the basic logic units 39 which are connected in series to form the "trip logic path" shown in FIG. 6. For clarity of the presentation, the basic logic units 39 are for the most part illustrated in simplified schematic form with some units, 39b and c, shown in abbreviated form due to space limitations. The A, T and G (*) signals which are applied to the control circuits of the several basic logic units 39 are normally low impedance, current sinking outputs of either microcomputer input-output cards, or bistable or logic cards. In the "tripped" state, they go to the high impedance, current blocking state. The A1 partial trip input for the 39a logic unit which is applied through line 99 is generated for example, by the output transistor of a processor 101 which together with transducer 103 forms the sensor 5 shown in FIG. 1. The processor 101 for instance, compares the output of the transducer 103 with selected limits and generates an output which is low when the parameter is within limits and high when it is out of limits. The T1 trip enable input for logic unit 39a, which is applied through line 105 and is an indication of whether or not there is a partial trip signal for that same parameter in one of the other protection channels, is generated in the trip enable computer 95 from data collected from the other channels through data link unit 13b. This basic logic unit 39a then performs the two out of four trip logic on the first monitored parameter since in the absence of a partial trip for that parameter in another channel, the T1 input will be low to prevent blocking of the pulse signal through logic unit 39a even if the signal A1 is in the "trip" state. Even so, the partial trip status of the signal A1 is reported to the global trip computer 97 through line 107 and is transmitted to the other channels through data link 13a where it generates a trip enable signal for that parameter in those other channels. Thus, when a corresponding partial trip occurs in another channel, both that channel and the channel of FIG. 6 will be blocked to trip their respective trip switchgear so that, as explained in connection with FIG. 1, the reactor will be tripped. A trip enable signal is also generated by the Trip Enable Computer 95 if any of the corresponding parameters in two out of three of the other channels are in bypass. This permits the logic unit 39a to trip the channel on partial trip signal alone to implement the one out of two logic for the two remaining unbypassed channels. The other A-T logic units, such as 39b and c, operate in a similar manner for the respective monitored parameters. The parameter level, or local bypass, is implemented, in part by the Trip Enable Computer 95 and Global Trip Computer 97, and in part by hard circuitry which includes the three position switch 109. As indicated, the three positions are "Normal", "Bypass" and "Trip" and the state of the two switch contacts, "a" and "b", for each switch position is indicated in the legend next to the switch. In the "Normal" position, the partial trip signal A1 is connected through the "a" contact to one of the AND inputs of the basic logic unit 39a which then operates in the manner discussed above. In the "Bypass" position, the "b" contact of the switch grounds and AND input to the logic unit, thus blocking the A.multidot.T function and bypassing the associated parameter. The status of the local bypass is sent ot the Global Trip Computer 97 on line 111 for transmission to the other channels along with the partial trip status. This status is sensed through double contacts on the logic unit PC card so that a missing card would be immediately detected by the sense of the partial trip signal. The bypass status is also indicated by an LED 113 next to the switch 109 which is turned on when the switch is in "bypass". A second LED 115 is illuminated as long as there is no partial trip indication. Some parameters such as the DNBR (departure from nucleant boiling) Trip, require an automatic actuation of the local bypass which is connected as shown in dashed line. Since the DNBR trip for each channel of the protection systems is calculated from conditions existing in one loop of the nuclear steam supply system of a PWR and that loop may be out of service, the automatic bypass permits the associated basic logic unit to be bypassed remotely. Finally, a connection is made through a diode 117 to a "Bypass Test Bus" which is grounded momentarily by an automatic tester (not shown) in the course of its test sequence. This allows the tester to verify that all local bypasses can be sensed. Each of the other A.multidot.T basic logic units in the trip logic path are bypassed in a similar manner. The G(*) function of equation 3 is implemented by the basic logic unit 39d in FIG. 6. This global trip signal is generated by the Global Trip Computer 95 and is sent to the logic unit 39d in two inverse logic senses with the inversion being carried out by the computer. The high impedance on trip signal G(*) is connected to an AND input (with a diode) on logic unit 39d and the low impedance on trip signal G(*)' is connected to the inverting INV input. This provides some redundancy for this function since failure of either output transistor in the global trip computer output will not prevent implementation of a global trip. As will be appreciated from examination of the boolean equations, this global trip signal is an indication that there is a coincidence of at least 2 partial trip signals for a given parameter in the other three channels. The global trip signals then trip the remaining channels to provide a redundancy that assures that the reactor is shutdown. The Global Trip Computer 95 also generates a global trip signal to block pulses in the logic unit 39 d if there are two out of four bypasses in coincidence with one out of two partial trips in the remaining channels or if there are three out of four bypasses for any parameter. Since the logic unit 39d is the last unit in the trip logic path, its output defines the condition of the trip logic path. Accordingly, its status is reported to the global trip computer through line 121. The status of the trip logic path is also transmitted to the automatic tester (not shown) so that the effects of the test sequence on the trip logic can be evaluated. The other path for pulses from the clock generator 91 to the converter 93 in the protection channel of FIG. 6 is by way of the global bypass logic leg. This leg is also made from the Basic Logic Units 39. In normal operation, the path of pulses through this leg is blocked. When a global bypass is applied, either manually via a switch on the bypass panel or automatically by the Automatic Test Subsystem, this path conducts the pulses to the power converter and simultaneously blocks the trip logic path. If permissive conditions do not allow the global bypass to be applied, i.e. a similar bypass already exists in another channel, then the global bypass logic leg is blocked by the action of the Global Bypass Permissive (GBP) signal. This condition causes both paths to be blocked and the reactor trip breakers for this channel to be tripped open. This action, together with the modification of the trip enable signal discussed above, implements the one out of two logic which is applied when two channels have been bypassed. By opening the reactor trip breakers on the second bypassed channel, a trip signal in either of the remaining channels will shut down the reactor. The global bypass logic path is made up of three basic logic units 39e, f and g. Logic unit 39e permits the clock pulses to pass through to logic unit 39f if the Global Bypass Permissive (GBP) signal remains low, i.e. no other bypasses exist. This GBP signal is generated by the Trip Enable Computer 95 from data gathered from the other channels and is applied to an AND input of logic unit 39e through line 123. The GBP signal goes high to block the flow of pulses through logic unit 39e if the Trip Enable Computer 95 determines that another channel has been bypassed before this one or if two out of three other channels are tripping. Basic logic unit 39f passes the pulses on to the power converter 93 if a global bypass is applied, either manually through switch 125 or automatically through the automatic test system (ATS) 127, to any one of its AND inputs. The bypass is applied by a low impedance signal. This same action blocks the pulses from the trip logic path unit 39f to the inverter input of logic unit 39g. As will be understood from the discussion in connection with FIG. 5, the control winding of logic unit 39g is normally turned on by current flowing froma 15 volt supply on the logic unit PC board through the control coil and the zener diode in the control circuitry. Grounding of an AND input to the logic unit 39f by a manual or automatic test bypass signal then diverts current from the control winding of logic unit 39g through the inverter input to block the flow of trip logic pulses. The statuses of logic units 39e, f and g are sent to the Global Trip Computer 97 over lines 131, 133 and 135 respectively for use in determining the overall status of the system. In addition to performing the functions described above, the Trip Enable Computer 95 and Global Trip Computer 97 perform various checks on the messages transmitted by the data links and on their own integrity to further enhance the reliability of the system. FIG. 7 illustrates the details of the power converter 93 shown in block form in FIG. 6 and the undervoltage coils on the reactor trip switch gear which it drives. The converter utilizes the pulse stream received from the trip logic path through logic unit 39g or from the global bypass path through logic unit 39f to convert the 24 volt d-c power avilable to the protection system to the 48 volt d-c power required for the undervoltage coils 137. Each one of the undervoltage coils 137 controls one of the pair of contactors in the reactor trip switchgear 19 associated with one of the protection channels (see FIG. 1). The particular converter used is a single ended, buck-boost derived, flyback d-c to d-c power converter. The transformer 139 provides the necessary electrical isolation between the input and output ground returns. The transformer 139 in the buck-boost derived flyback converter, also serves as the inductive energy storage media. When transistor Q1 is "ON", energy is stored in the primary winding 141. During this time, diode 143 does not conduct because of the phase relationship of the secondary winding to that of the primary side. When Q1 turns "OFF", the diode 143 conducts and the energy stored in the magnetic field of the transformer 139 is released to the output filter capacitor 145 and undervoltage coil 137 via the secondary winding 147. As long as pulses are applied to the converter, sufficient output voltage is generated to maintain the undervoltage coils 137 in the energized state. Termination of the input pulse train causes the undervoltage coils to drop out thereby tripping the associated reactor trip switch gear for that channel. The operation of the switchgear can be performed manually by a switch 149. A plurality of the various protection channel switches 149 are arranged in a stack so that a single operation opens all of the reactor trip switchgear devices. The switching transistor Q1 for the power converter is preferably a power MOSFET protected from voltage spikes during switching by zener diodes 151 and 152. Because of the absence of the second breakdown phenomenon in power MOSFETs, they are ideal switching devices for circuits which drive inductive loads such as d-c to d-c converters. The transistor Q1 is controlled by drive logic 153 which includes an exclusive OR gate 155 to which the pulses from either the trip logic path or global logic path are applied. As discussed previously, only one stream of pulses or the other is delivered to the converter, while in the trip condition both pulses streams are blocked. The output of the "EX-OR" gate 155 is connected to a NOR gate 157 directly and through a monostable 159 such as an MC14528. The monostable 159 serves a double function in this circuit: first, in normal operation, it permits the pulses from "EX-OR" gate 155 to pass through the NOR gate 157 to the gate circuit of the switching MOSFET transistor Q1; its second function is to "time-out" the NOR gate 157, when tripped, in order to leave the switching transistor Q1 in the "OFF" state (non-conducting). The monostable 159 is connected in the retriggerable mode with a rising edge trigger. It is retriggered if a valid trigger occurs, followed by another valid trigger, before its output has returend to a quiescent state. The inverting output is used, so that it is retriggered to its "low" state. The timing is set at 300 microseconds so that 5 KHz pulses coming from the logic paths hold the not "Q" output low. When the pulses are interrupted (e.g. as a result of a trip signal), the not "Q" output changes state to "High", pulling the NOR gate 157 output "low" to turn the switching transistor Q1 "OFF". This action causes the voltage on the secondary side of the transformer 139 to drop to zero and deenergize the undervoltage coils 137. Action of the monostable 159 is not required to cause the secondary voltage to drop to zero, loss of drive pulses is sufficient. However, by turning off the transistor Q1 following a trip, components are protected against excessive heating by d-c currents. While specific embodiments of the present 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 arrangements 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. |
abstract | The invention relates to the field of nuclear technology and radiochemistry, more specifically to the production and isolation of radionuclides for medical purposes. The method for producing actinium-225 and isotopes of radium comprises irradiating a solid block of metallic thorium of a thickness of 2 to 30 mm, which is contained within a hermetically sealed casing made of a material which does not react with thorium, with a flow of accelerated charged particles with high intensity. The irradiated metallic thorium is removed from the casing and is either heated with the addition of lanthanum and the distillation of radium or is dissolved in nitric acid with the recovery of actinium-225 by extraction. A target for implementing this method consists of blocks of metallic thorium of a thickness of 2 to 30 mm, which are contained within a hermetically scaled casing made of different materials which do not react with thorium. |
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summary | ||
claims | 1. A power distribution monitoring system for a reactor, comprising: a plurality of in-core nuclear instrumentation assemblies each having a nuclear instrumentation tube, a plurality of fixed neutron detectors housed in the nuclear instrumentation tube and adapted to detect neutron flux of a local power distribution of a power range in a core of the reactor and a gamma-ray thermometer assembly housed in the nuclear instrumentation tube, said gamma-ray thermometer assembly having a plurality of fixed xcex3-ray heating detectors for detecting xcex3-ray heating values and a heater built therein and adapted to calibrate the fixed xcex3-ray heating detectors, said fixed xcex3-ray heating detectors being arranged at least close to the fixed neutron detectors; means for processing neutron flux detection signals based on the detected neutron flux by each of the fixed neutron detectors; means for processing gamma-ray thermometer signals based on the detected xcex3-ray heating values to obtain gamma ray heating measurement signals of the fixed xcex3-ray heating detectors of each of the gamma-ray thermometer assemblies; means for electrically energizing the heater in each of the gamma-ray thermometer assemblies; means for storing a plurality of predetermined time intervals; means for selecting one of the predetermined time intervals for specified xcex3-ray, thermometer assemblies respectively; means for detecting a core state data representing a state of the core of the reactor; means for detecting a change of the core state according to the detected core state data to determine whether a selected predetermined time interval after detecting the change of the core state elapses; and means for gathering at least one of the gamma-ray thermometer signals and the gamma-ray heating values, wherein said gamma-ray thermometer signal being processed by the gamma-ray thermometer process means from a fixed xcex3-ray heating detector, and said gamma-ray heating measurement signals being calculated from the gamma-ray thermometer signal; and means for simulating a power distribution in the core according to at least one of the gamma-ray heating measurement signals and the neutron flux detection signals; wherein said energizing means is adapted to control an electrical energy supplied to the heater according to the selected predetermined time interval to heat the heater, thereby executing a heater calibration of output voltage sensitivities of the fixed xcex3-ray heating detectors of the gamma-ray thermometer assembly when the selected predetermined time interval after detecting the change of the core state elapses. 2. A power distribution monitoring system according to claim 1 , further comprising means for adjusting at least one of a sensitivity and a gain of the fixed neutron detector by using the gamma-ray heating measurement signals calculated from the gamma-ray thermometer signals of the fixed xcex3-ray heating detector when the selected predetermined time interval after detecting the change of the core state elapses, claim 1 said fixed xcex3-ray heating detector and said fixed neutron detector being housed in an identical in-core instrumentation tube, said predetermined fixed xcex3-ray heating detector being located in an identical core axial direction of the adjusted fixed neutron detector. 3. A power distribution monitoring system according to claim 1 , wherein said detecting means inputs the neutron flux detection signals to the simulating means in a place of the gamma-ray thermometer heating measurement signals, when a selected predetermined time interval after detecting the change of the core state has time remaining before the selected predetermined time interval elapses, claim 1 and wherein said simulating means is configured for storing predetermined adaptive correction quantities by simulating power distribution according to the gamma-ray heating measurement thermometer signals at a current point in time, executing a power distribution simulation corresponding to the core state at the current point in time based on the above predetermined adaptive correction quantities and a current core state, obtaining a pseudo gamma-ray heating measurement signal from the neutron flux signal by correcting a difference in response between the change of the neutron flux signal and a change of a gamma-ray thermometer signal, both of said changes being predicted by a simulation while also accounting for a change in a control rod state and a void fraction of fuel nodes around the neutron flux detectors from a previous point in time up to the current point in time, simulating the power distribution at the current point in time while also adapting the power distribution by interpolating and extrapolating correction ratios to obtain an additional correction ratio of a plurality of axial nodes, said correction ratios being obtained by making a comparison between the pseudo gamma-ray heating measurement signal and a simulated equilibrium value of the gamma-ray heating measurement signal in an axial direction of the core, evaluating a power distribution when a gamma-ray heating measurement signal a is in a non-equilibrium transient state, and executing a zero-clear process of the additional correction ratio when the gamma-ray heating measurement signal is in an equilibrium state by simulating the power distribution to obtain the adaptive correction quantity according to the gamma-ray thermometer signal. 4. A power distribution monitoring system for a reactor, comprising: a plurality of in-core nuclear instrumentation assemblies each having a nuclear instrumentation tube, a plurality of fixed neutron detectors housed in the nuclear instrumentation tube and adapted to detect neutron flux of a local power distribution of a power range in a core of the reactor and a gamma-ray thermometer assembly housed in the nuclear instrumentation tube, said gamma-ray thermometer assembly having a plurality of fixed xcex3-ray heating detectors for detecting xcex3-ray heating values and a heater built therein and adapted to calibrate the fixed xcex3-ray heating detectors, said fixed xcex3-ray heating detectors being arranged at least close to the fixed neutron detectors; means for processing neutron flux detection signals based on the detected neutron flux by each of the fixed neutron detectors; means for processing gamma-ray thermometer signals based on the detected xcex3-ray heating values to obtain gamma-ray heating measurement signals of the fixed xcex3-ray heating detectors of each of the gamma-ray thermometer assemblies; means for electrically energizing the heater in each of the gamma-ray thermometer assemblies; means for storing a plurality of predetermined time intervals; means for selecting one of the predetermined time intervals for specified xcex3-ray thermometer assemblies respectively, means for detecting a core state data representing a state of the core; means for detecting a change of the core state according to the detected core state data to determine whether a predetermined time interval after detecting the change of the core state elapses; means for estimating equilibrium signal levels of the gamma-ray heating measurement signals of the gamma-ray heating detectors; and means for executing a current power distribution simulation corresponding to the core state at a current point in time while also adapting the power distribution by interpolating and extrapolating correction ratios that are obtained by making a comparison between gamma-ray thermometer signal reading values simulated from a previous simulated power distribution and the estimated equilibrium signal levels of the gamma-ray heating measurement signals acquired in an axial direction of the core, so as to obtain correction ratios of a plurality of axial nodes, thereby evaluating a current power distribution when gamma-ray heating measurement signals are in a non-equilibrium transient state, and when the selected predetermined time interval after detecting the change of the core state has time remaining before the selected predetermined time interval elapses. 5. A power distribution monitoring system according to wherein said execution means adjusts at least one of a sensitivity and a gain thermometer signal reading value of a fixed xcex3-ray heating detector, said fixed xcex3-ray heating detector and said adjusted fixed neutron detector being housed in an identical in-core instrumentation tube, said fixed xcex3ray heating detector being located in an identical core axial direction of the fixed neutron detector. 6. A power distribution monitoring system according to claim 4 , wherein said execution means simulates reading values of the fixed neutron detectors according to the correction ratios of a current core power distribution, and compares the simulated reading values of the fixed neutron detectors and measured values of the fixed neutron detector at the current point in time so as to adjust at least one of the sensitivity and the gain of the fixed neutron detector, thereby making the measured values coincide with the simulated reading values. claim 4 |
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061838179 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to the fabrication of nano-devices and semiconductor devices. More specifically, the present invention relates to direct write fabrication and lithography techniques and systems employing the Lorenz force interactions of light fields with the dipole moment of atoms to build stable nanostructures of matter via the direct deposition of atoms upon substrates. 2. Description of the Prior Art The nano-device and semiconductor device manufacturing communities consistently strive to develop fabrication techniques and equipment that enable the precise definition of ever smaller structural components of functional devices. These device manufacturers expend extensive financial and technological resources in the exploration and development of improved photolithography process tools and methods on a continual basis. The potential rewards for achieving significant reduction in minimal dimension manufacturable structures, features or spot sizes, are tremendous. Smaller structural components can be used to create smaller devices. Semiconductor devices constructed with smaller feature and structure dimensions operate at faster speeds, consume less power and embody higher functional complexities. Smaller nano-devices function within finer scales of precision. The economic motivation of manufacturers to improve over the capabilities of conventional photolithography has therefore led to the funding of much intensive work in the field of x-ray level photolithography. Yet the fundamentals of prior art photolithography are self-limiting to enhancements in several ways. Conventional photolithography is a multi-step process. Each step typically contributes an error factor to a finished structure. An example of a photolithography process might include a deposition of a specific material onto a substrate, the application of a layer of photoresist over the substrate, a soft bake of the photoresist, an exposure of the photoresist layer to a pattern-masked light source, the development of the photoresist, a hard bake of the photoresist and finally an etching step. Each step in this example photolithography process has limitations in accuracy and result. The validity of the deposition action is dependent upon the uniformity of the layer of material actually deposited. The resolution of the component structures of the device defined in the photoresist development steps is limited by the wavelength of light used, the geometric accuracy of the pattern-mask, the effects of unintended under or over exposure to the light source, and the equally unintended effects of under development or over development of the photoresist after exposure to the light source. The soft baking and/or the hard baking of the photoresist can also be over or under done. The precision of the etch step is affected by the degree of selectivity of the etching agent and the degree of anisotropy achieved in the delivery of the wet or dry etching agent across the entire surface of the substrate. The errors introduced by each step of the photolithography process are further compounded by the fact that the fabrication of most semiconductor devices requires the repeated application of entire cycles of photolithography processes which employ widely varying deposition materials, photoresist compounds, etching chemicals, pressure and temperature requirements. The net result is often an increasing limitation in the smallest achievable individual structure size, as well as reduction in the control of precision in device feature fabrication. The sheer number of individual process steps of a typical device fabricated with conventional photolithography techniques raises the statistical occurrence of contaminating events during the manufacturing process. The elimination of process steps in itself typically results in yield improvement. The dominant trend in the art is to attempt to increase the accuracy of each step of the photolithography process. Accuracy in etching may be increased by reducing the thickness of the deposited film. However, this increases the requirements for uniformity in the deposition phase. In addition, these thin deposition layers are, in some cases, thin enough to experience and exhibit quantum phenomenon and behavior. However, for the most part, these layers have bulk material properties. This can lead to many uncontrolled parameters during the fabrication stage. For example, a convex surface profile of the deposited layer may lead to strained and mismatched lattices, thus creating devices susceptible to failure through numerous mechanisms. The pursuit of better photolithography through the application of higher frequency light waves, such as hard ultraviolet and x-rays, is based upon a relationship expressed in the Rayleigh Criterion between the wavelength of a light wave and its corresponding achievable diffraction limit. The Rayleigh Criterion is given by: EQU d=0.61 .lambda. N.A. where .lambda.=operational wavelength of the emitted light; PA1 N.A.=Numerical Aperture of the imaging optic; and PA1 d=diameter of the minimal achievable spot size. PA1 h=Planck's constant=6.62.times.10.sup.-34 J/second; PA1 m=the mass of the atom in Kg; and PA1 v=the velocity of the atom in meters/second. Replacing .lambda. in the equation of the Rayleigh Criterion with the wavelength .lambda. of a particular light wave will yield the diffraction limit of the light wave emitted at that frequency. Higher frequencies obviously exhibit lower diffraction limits. The employment of shorter illumination wavelengths does theoretically allow for the definition of significantly smaller device feature sizes. Unfortunately, however, the greater photon energies of higher frequency light waves often create other obstacles to the process of manufacturing smaller geometric features. Higher frequency light waves are more likely to damage lithographic mask optics very quickly because of their high levels of material interaction reactivity. Light waves of wavelengths below 200 nm are more prone to uncontrolled scatter and absorption by the substrate and other deposited materials. Higher energy photons are also more prone to damage the material that they are bombarding. Conventional photolithography techniques, including those employing x-rays, still typically include an etching step and require the provision of a substrate composed of a pure single crystal, such as silicon or gallium arsenide. Device structures are currently fabricated with atoms in higher energy states, which result in an increased incidence of lattice mismatch and attendant structural fragility. Furthermore, most conventional photolithography applications entail the production of environmentally hazardous solutions, reagents and by-products. Manufacturing techniques that reduce the generation of biohazardous material by the nano-device industry, and especially the semiconductor device industry, are of significant benefit to the both manufacturers and the world community. Creative and meaningful work has been done in the area of controlling the deposition of atoms through the Lorenz force effect created by the interactions between an atomic dipole and a standing light wave. Much of this work exploits the generation of a Lorenz force caused by the interaction of an atom, behaving like an oscillating dipole, with the oscillating electric field of a laser. The atom behaves like an oscillating dipole acted on by a Lorenz force, where the Lorenz force is proportional to the intensity gradient of the oscillating electric field of the laser. J. J. McClelland and M. R. Scheinfein, for a first example, proposed the use of a laser beam as a means of focusing an atomic beam to create nanometer, or nm, scale spots. (J. J. McClelland and M. R. Scheinfein, "Laser focusing of atoms: a particle-optics approach", J. Opt. Soc. Am. B/Vol. 8, No. 9/September 1991, pp. 1974-1986, which is hereby incorporated by reference) McClelland and Scheinfein envisioned the employment of a TEM.sub.01 laser as an atomic lens whereby the direction of an atomic beam is purposely directed with a focal spot size on the order of one nanometer. T. Sleater et al. have reported on the successful implementation of an atomic lens scheme wherein a cylindrical lens potential was created by positioning a large period, 45 micron, standing light wave perpendicular to a supersonic beam of metastable helium atoms. (T. Sleator, T. Pfau, V. Balykin, and J. Mlynek, "Imaging and Focusing of an Atomic Beam with a Large Period Standing Light Wave", Applied Physics B, 1992, pp. 375-379, which is hereby incorporated by reference) The thinness of the established lens was estimated to be 40 microns. An achieved spot size of four microns was primarily limited by diffraction. Additionally, a microfabricated grating with a period of eight microns was constructed. Chromatic, spherical and diffusive aberrations appeared to have little impact on the spot size. Sleator et al. further suggested that the thinness of the lens could enable lithographic applications in the nanometer range. Sleator et al., with helium atoms excited by copropagation of electrons at an energy level of about 31 eV, has a resulting beam of metastable helium atoms having an average velocity of approximately 1760 m/s and a corresponding deBroglie wavelength of 0.56 angstroms. The deBroglie wavelength is calculated from the following expression: EQU l.sub.db =h/mv= where Sleator et al. further proposed that their technique could be generalized to two dimensions by combining two standing waves. The creation of two dimensional device structures should thereby be executable. Sleator et al. also predicted that an atomic beam exhibiting a particle velocity of 900 m/s should be focusable with a laser power of ten mW into spot sizes of ten nm. In another example where neutral atom lithography was accomplished, Timp et al. used an optical standing wave of 589 nm as an array of cylindrical lenses to focus a perpendicular sodium beam and to thereby construct a grating on a substrate, where a periodicity of 294.3+/-0.3 nm was achieved. (G. Timp, R. E. Behringer, D. M. Tennant, and J. E. Cunningham, "Using Light as a Lens for Submicron, Neutral-Atom Lithography", Physical Review Letters, Vol. 69, No. 11, Sep. 14, 1992, pp 1636-1639, which is hereby incorporated by reference.) G. Timp et al. further described a collimated sodium atomic beam propagating along a y axis and interacting with a perpendicularly oriented standing wave (SW). In preparing the atomic beam for interaction with the standing wave the atomic beam was passed through a "Doppler" optical molasses in order to reduce the transverse velocity and cool the sodium atoms. The average force, U.sub.(z), exerted by the standing wave on the sodium atoms then acts as an array of weak cylindrical lenses and focuses the atomic beam into a grating on the substrate with a period one half of the wavelength of the standing wave, or approximately 589 nm/2=294.5 nm. J. J. McClelland et al. report the use of laser light to control the motion of a chromium atomic beam to fabricate a nanostructure. (J. J. McClelland, R. E. Scholten, E. C. Palm, and R. J. Celotta "Laser-Focused Atomic Deposition" Science, Vol. 262, Nov. 5, 1993, pp. 877-880, which is hereby incorporated by reference.) The resulting nanostructure consisted of a series of lines which showed line widths of 65 nm+/-6 nm, line spacings of 212.78 nm and heights of 34 nm+/-10 nm. The techniques of Timp et al., Sleator et al. and McClelland et al., rely upon the application of the two interaction mechanisms existing between laser fields and atoms, namely the spontaneous force and the dipole or gradient force. The dipole force is discussed above as the Lorenz force. The spontaneous force is used in the construction of an optical molasses, where a laser field repeatedly bombards an atom with photons. The atom will then radiate photons in random directions through spontaneous emission. The resultant effect of this atomic absorption and radiation of photons includes a net transfer of momentum to most of the subject atoms in the direction of the absorbed photons, as the momenta of spontaneously emitted photons will statistically average to zero. Thus, the spontaneous force can be used in an optical molasses apparatus to cool an atomic beam. The conventional art thus shows an interest in using lasers to direct the deposition of atomic beams onto substrates and to form simple structures and features of nano-devices and semiconductor devices. The prior art, however, lacks significant enablement work to allow for the efficient construction of most conventional semiconductor devices, as the creation of large numbers of nanostructures by the direct deposition of atomic beams onto substrates via the application of Lorenz force interactions between atomic beams and light fields requires the production of numerous specific patterns of photonic energy. The technologies of holographic generation, however, offer efficiencies in the creation of the necessary photonic energy patterns. Commercially available software packages, e.g. the ZEMAX-EE product from Focus Software or Wolfram Research's Mathematica package, are capable of calculating from a desired holographic shape and orientation to define the required diffraction pattern to generate the intended hologram. These mathematically powerful software programs indicate the feasibility of back calculating diffraction patterns upon the basis of a mathematical definition of the desired image. The design of photonic lenses to focus a beam of atoms using Lorenz force interactions requires that the shape, energy states of individual atoms and isotopic composition of the atomic beam be precisely anticipated. High intensity evanescent waves have been demonstrated by R. Kaiser et al., (R. Kaiser, Y. Levy, N. Vansteenkiste, A. Aspect, W. Seifert, D. Leipold and J. Mlynek "Resonant Enhancement of Evanescent Waves with a Thin Dielectric Waveguide", Optics Communications, Vol. 104, No. 4, 5, 6, (1994), pp. 234-240, which is hereby incorporated by reference.) in conjunction with thin dielectric plates to be used as an atomic mirror. The technique of Kaiser et al. couples a laser beam to a dielectric wave guide by optical tunneling through a solid gap. An enhanced evanescent wave is thereby produced in a vacuum above the wave guide. This evanescent wave functions as an atomic mirror. There is a long felt need in the industries of nano-device and semiconductor device manufacturing to efficiently and accurately define more robust structural components of devices with smaller dimensions than the prior art allows. There is also a long felt need to limit the environmental impact of manufacturing semiconductors by reducing the volume of toxic chemicals generated for and by semiconductor processing. Alternatives to subtractive processing techniques, whereby material is first applied to a substrate or structure and then selectively removed, may therefore offer significant value to the art of nano-device and semiconductor device manufacture. SUMMARY AND OBJECTS OF THE INVENTION It is, therefore, an object of the present invention to construct a method and an apparatus for the direct deposition of particles, atoms or molecules with high accuracy on a substrate. It is another object of the present invention to provide a method and apparatus for direct write fabrication, or direct write lithography, of nanostructures of a nano-device or a semiconductor device through the skillful manipulation of Lorenz force interactions of dipole moments of atoms and light fields. It is a further object of the present invention to provide a method and an apparatus capable of fabricating nanostructures on amorphous substrates. It is yet another object of the present invention to provide a method and an apparatus for the planarization of amorphous substrates. It is still a further object of the present invention to provide a method and apparatus for the achievement of deposition/device layer thickness uniformity and control with improved precision. It is still another object of the present invention to improve and simplify the manufacturing processes of nano-devices and semiconductor devices. This elimination of process steps typically provides for greater device yields by reducing yield losses due to statistical contamination events. It is yet a further object of the present invention to provide a method and apparatus capable of producing semiconductor devices with increased robustness and reduced incidence of failure from common causes such as electrostatic discharge events and thermal runaway. In accordance with the above objects of the invention, the present invention provides a direct write fabrication, or direct write lithography method and apparatus employing the Lorenz force, wherein the interaction of light fields of a photonic lens or lenses with the dipole moment of atoms is used to control the direct deposition of atoms upon appropriate substrates. The method of the present invention precisely controls the amplitude, phase, polarization and dimensions of light fields through the use of holographic techniques and computer generated holographic elements optionally in conjunction with other optical elements to directly manipulate the energy state of individual atoms, thus determining their exact bond sites and energies. In particular, the capability of the preferred embodiment to construct nanostructures with particles, such as atoms or molecules, in lower energy states enables the creation of stronger and more rugged device features. The preferred embodiment of the present invention comprises the additive fabrication process of deposition without requiring the inclusion of subtractive process steps such as etching. Furthermore, the preferred embodiment of the method of the present invention eliminates or reduces the purity requirement in many applications for expensive single crystal substrates by enabling the depositional construction of nanostructures on amorphous substrate materials. The method of the present invention thus allows the fabrication of device structures on inexpensive amorphous silicon dioxide, e.g. glass, ceramic substrates, or other suitable materials known in the art. Because the preferred embodiment is purely additive, the method of the present invention can reduce the routine preparation and generation of significant volumes of many of the toxic and corrosive chemicals and by-products currently found in conventional art photolithography and etching process steps. The preferred embodiment of the present invention can be used for substrate planarization and/or the fabrication of semiconductor devices, electro-optical devices, electrical devices, mechanical devices and other devices with nanostructures, or nano-scale features known in the art, thus avoiding the need for multiple lithography, deposition and etch processes and permitting the creation of semiconductor and other nano-fabricated devices with smaller critical dimensions than has heretofore been achieved. The method of the preferred embodiment includes the propulsion of a stream or a vector comprising a substantially isotopically homogeneous matter, such as a highly pure, isotopically specific atomic beam. The stream, vector or particle beam may form a particle, molecular or atomic beam comprising a multiplicity of particles such as single atoms or molecules. The content of the particle beam is filtered from an originating feedstock to comprise a particular and predesignated atomic or molecular isotope. The particle, atomic or molecular beam is generated and processed through ionizing and deionizing steps, detuning stages, an optical molasses region and an output to impose a highly collimated, low energy level uniformity among the individual atoms of the beam. The particle beam is then delivered with predetermined vectors and energy states into a fabrication reactor. The particle beam then interacts with one or more holographically generated laser fields or photonic lenses resident within the fabrication reactor. These photonic lenses are specifically generated to precisely focus the particle beam onto a substrate and thereby create device features and nanostructures by direct write fabrication. The method of the preferred embodiment requires a process step of a preparation for generation of one or more photonic lenses. This preparatory work includes the steps of: (1.) mathematically modeling a feature or structure to be built, (2.) identifying a particular elemental or molecular isotope that will be used to construct the feature or structure, (3.) determining the vector, energy states and entrance point of the isotope molecules or atoms at the moment of insertion of the particle beam into the fabrication reactor in relationship to the intended position of the substrate during the direct write fabrication action, (4.) back calculating the required characteristics of one or more photonic lenses that will direct, by means of Lorenz force interaction, the particle beam of known parameters to form the mathematically modeled feature or structure, (5.) back calculating the laser frequency or frequencies and the diffraction pattern or patterns required in one or more optical lenses to generate the required photonic lens or lenses as defined in step 4 and (6) prepositioning a laser source or sources which can generate the required laser frequencies, as determined in step 5, in relationship to i.) a substrate, ii.) the optical lens or lenses, where the lens or lenses contain the necessary diffraction patterns as calculated in step and iii.) the fabrication reactor, whereby the required photonic lens or lenses are established within the fabrication reactor coincident with and in proper orientation to the introduction position of the particle beam containing the preselected isotope and the substrate in order to build the predesignated feature or structure. The method of the present invention includes the inventive application of concepts and techniques used to shape light beams in the optical arts to condition and shape a particle beam and to establish the energy characteristics, instantaneous quantum nature and flow vector of the particle beam to create nano-device and semiconductor device features and structures on a prepositioned substrate. The method of the present invention as carried out in the preferred embodiment requires the provision of a feedstock into a high temperature vacuum crucible from which a feedstock vapor is generated. A particular isotope of the feedstock vapor is then isolated by a photo-ionization step wherein a laser tuned to a specific frequency chosen to selectively ionize the desired isotopic species, and with minimal affects to the remainder of the feedstock vapor, is focused at the feedstock vapor. The newly formed ions of the isotopic species are then directed into and through a particle beam generator by means of the momentum imparted to the selected isotopic atomic species from the photo-ionization action, and by means of conventional art ionic pumps and electromagnetic field buffering. The beam of ions is then de-excited by application of detuned radio frequency energy and neutralized by interaction with an electron source. The electrically neutral particle beam next leaves the particle beam generator after passing through, and being more precisely shaped by, a grazing incidence nozzle. The particle beam is immediately inserted from an exit port of the grazing incidence nozzle into an optical molasses cooling chamber. The optical molasses cooling chamber reduces the energy level of the isotopic species to near ground level yet allows the particle beam to continue onward and through an output coupler. The output coupler of the preferred embodiment includes a set of tuned traveling/evanescent wave plates. Certain alternate preferred embodiments of the present invention optionally include magnetic mirrors in combination with electromagnetic probing energy and/or other suitable components and methods in the design of various alternate output coupler embodiments. The output coupler further uniformly orders the individual isotopic species atoms into a narrow band of low energy states and delivers the particle beam to a fabrication reactor. The particle beam then comprises a multiplicity of atoms of the preselected isotopic species wherein virtually all of the isotopic atoms exhibit an energy level within an anticipated and narrow band of energy states. A photonic lens is synchronously generated in the fabrication reactor to direct the particle beam towards a location on a prepositioned substrate in order to form a desired structure. The control of the flow vector of the particle beam is accomplished by means of Lorenz force interactions of the photonic lens or lenses with the dipole moments of the individual atoms of the particle beam. The term flow vector is defined as the three dimensional direction and speed of the atoms of the particle beam in either an individual or a collective sense. Each photonic lens is imposed within the reactor by means of passing a laser beam of a preselected frequency and energy content through an optical lens, where the optical lens contains a particular and previously calculated and recorded holographic diffraction pattern. The method of the present invention teaches that a photonic lens is generated from a material holographic lens, e.g. glass, where the photonic lens directs the flow of a relatively isotopically pure atomic or molecular particle beam towards a substrate to effect the fabrication of a nano-scale device feature of a predetermined size, shape and orientation within the substrate. The form and strength of the desired photonic lens is calculated from a precise knowledge of the physical, quantum and electromagnetic properties of the particular isotopic species to be controlled, the position of the substrate in relation to the pathway of the particle beam and a mathematical description of the size, shape and orientation of the desired nano-feature within the substrate. The diffraction pattern of the material holographic lens used to generate the photonic lens is back calculated from a knowledge of the nature of the laser beam frequency and strength available, the nature of the holographic lens material, the physical position of the holographic lens in relationship to the particle beam and the substrate, and the shape, size, orientation and strength of the photonic lens. These, together with the various ancillary objects and features, will become apparent to those possessing the ordinary skill in the art as the following description proceeds, a preferred embodiment being shown with reference to the accompanying drawings, by way of example only, wherein: |
description | The present invention relates to the field of materials used in the nuclear field, in particular materials intended to exhibit the best possible resistance to the physicochemical conditions encountered under nominal conditions and during a nuclear reactor accident, for instance in a pressurized water reactor (PWR) or a boiling water reactor (BWR). The invention relates more particularly to the process for manufacturing of a nuclear component, to the nuclear component, and to the uses thereof against oxidation and/or hydriding. The constituent zirconium alloy of current nuclear fuel claddings becomes oxidized on contact with water which constitutes the coolant for PWR or BWR nuclear reactors. Since the oxide form is fragile and the uptake of hydrogen associated with oxidation leads to the precipitation of zirconium hydrides which are embrittling, the service life of the claddings is partly limited by the maximum acceptable thickness of oxide and the associated content of absorbed hydrogen. To ensure good residual mechanical properties of the cladding directed toward ensuring optimum confinement of the nuclear fuel, the residual thickness of healthy and ductile zirconium alloy must be sufficient and/or the fraction of hydrides must be sufficiently limited to ensure good residual mechanical properties of the cladding directed toward ensuring optimum confinement of the nuclear fuel. The possibility of limiting or retarding such oxidation and/or hydriding may thus prove to be crucial under the conditions of an accident. These conditions are reached, for example, in the case of hypothetical accident scenarios such as RIA (“Reactivity Insertion Accident”) or LOCA (“Loss of Primary Coolant Accident”), or even under conditions of dewatering of the spent fuel storage pool. They are among others characterized by high temperatures which are generally greater than 700° C., in particular between 800° C. and 1200° C., and which may be reached with a high rate of temperature increase. At such temperatures, the coolant is in the form of steam. Oxidation under the conditions of an accident is much more critical than under the conditions of normal nuclear reactor functioning, since the deterioration of the cladding, which is the first barrier for confinement of the fuel, is more rapid and the associated risks are greater. These risks are among others the following: emission of hydrogen; embrittlement of the cladding at high temperature by oxidation or even, under certain conditions, hydriding of the cladding; embrittlement of the cladding by quenching, caused by the abrupt decrease in temperature during the massive introduction of water for the securing of the nuclear reactor core; low mechanical strength of the cladding after quenching or cooling, in the case among others of handling operations after an accident, aftershocks of earthquakes, etc. Given these risks, it is essential to limit as far as possible the oxidation and/or hydriding of the cladding at high temperature, or even at very high temperature, so as to improve the safety of nuclear reactors among others using water as coolant. The very high temperatures are situated at the extreme limit or even beyond that of the high temperatures comprised between 700° C. and 1200° C. that are set by the accident regulatory conditions. Now, the regulatory criteria governing the design-basis accidents according to the scenario of LOCA type defined in the 1970s stipulated that the maximum temperature of the cladding should not exceed 1204° C. (2200° F.) and stipulated a maximum degree of “ECR” oxidation of 17%. The degree of “ECR” (“Equivalent Cladding Reacted”) oxidation is the percentage of thickness of metal cladding transformed into zirconia (ZrO2) resulting from the oxidation of the zirconium contained in the nuclear fuel cladding, assuming that all the oxygen that has reacted forms stoichiometric zirconia. To take into account the additional embrittling effect associated with the hydriding of the cladding in service, this degree of acceptable residual “ECR” oxidation may now be substantially less than 17% under certain conditions, for instance a hydridized cladding in service of up to several hundred ppm by mass, which corresponds in practice to a degree of oxidation of the cladding which must not exceed a few minutes at 1200° C. Improvement in the resistance to oxidation and/or hydriding at very high temperature would advantageously make it possible to obtain additional safety margins, among others by proportionately preventing or retarding the deterioration of the cladding in the event of worsening or persistence of the accident situation. One of the aims of the invention is thus to prevent or attenuate one or more of the drawbacks described above, by proposing a nuclear component and the process for manufacturing same which improves the resistance to oxidation and/or hydriding, among others in the presence of water vapor. Another aim of the invention may be to improve this resistance to oxidation and/or hydriding at high temperature between 700° C. and 1200° C., or even at very high temperature above 1200° C.; among others when these temperatures are reached with a temperature increase rate which is comprised between 0.1° C./second and 300° C./second. Another aim of the invention may be to improve the oxidation and/or hydriding resistance time, beyond which time the integrity of the nuclear component, in particular the nuclear fuel confinement, is no longer ensured. Another aim of the invention may be to improve the potential for industrialization of the manufacturing process, among others by proposing a process that is not only versatile also economical and more environmentally friendly. The present invention thus relates to a process for manufacturing a nuclear component via the method of chemical vapor deposition of an organometallic compound by direct liquid injection (DLI-MOCVD), the nuclear component comprising: i) a support containing a substrate based on a metal chosen from, zirconium, titanium, vanadium, molybdenum or base alloys thereof, the substrate being coated or not with an interposed layer placed between the substrate and at least one protective layer; ii) said at least one protective layer coating said support and composed of a protective material comprising chromium chosen from a partially metastable chromium comprising a stable chromium crystalline phase and a metastable chromium crystalline phase, an amorphous chromium carbide, a chromium alloy, a carbide of a chromium alloy, a chromium nitride, a chromium carbonitride, a mixed chromium silicon carbide, a mixed chromium silicon nitride, a mixed chromium silicon carbonitride, or mixtures thereof; the process comprising the following successive steps: a) vaporizing a mother solution containing a hydrocarbon-based solvent free of oxygen atoms, a precursor of bis(arene) type comprising chromium; and containing, where appropriate, an additional precursor, a carbon incorporation inhibitor or a mixture thereof; the precursors having a decomposition temperature comprised between 300° C. and 600° C.; b) in a chemical vapor deposition reactor in which is located said support to be covered and the atmosphere of which is at a deposition temperature comprised between 300° C. and 600° C. and at a deposition pressure comprised between 13 Pa and 7000 Pa (or even from 130 Pa to 4000 Pa); introducing the mother solution vaporized in step a) and then depositing said at least one protective layer onto said support, this deposition being brought about by this introduction of the mother solution vaporized in step a). More particularly, the protective material comprising chromium is the amorphous chromium carbide. More particularly, the manufacturing process of the invention thus relates to a process for manufacturing a nuclear component via the method of chemical vapor deposition of an organometallic compound by direct liquid injection (DLI-MOCVD), the nuclear component comprising: i) a support containing a substrate based on a metal chosen from zirconium, titanium, vanadium, molybdenum or base alloys thereof, the substrate being coated or not with an interposed layer placed between the substrate and at least one protective layer; ii) said at least one protective layer coating said support and composed of a protective material comprising chromium which is an amorphous chromium carbide; the process comprising the following successive steps: a) vaporizing a mother solution containing a hydrocarbon-based solvent free of oxygen atoms, a precursor of bis(arene) type comprising chromium; and containing, where appropriate, an additional precursor; the precursors having a decomposition temperature comprised between 300° C. and 500° C.; b) in a chemical vapor deposition reactor in which is located said support to be covered and the atmosphere of which is at a deposition temperature comprised between 300° C. and 600° C. and at a deposition pressure comprised between 13 Pa and 7000 Pa; introducing the mother solution vaporized in step a) and then depositing said at least one protective layer onto said support. In comparison with the processes of the prior art, the manufacturing process of the invention among others improves the resistance of a nuclear component with respect to oxidation and/or hydriding, while at the same time offering a deposition process with great potential for industrialization enhanced by the possibility of recycling. The manufacturing process of the invention uses a method of chemical vapor deposition of an organometallic compound by direct liquid injection (known as DLI-MOCVD, according to the English acronym for “Direct Liquid Injection—Metal Organic Chemical Vapor Deposition”). This method is described, for example, in the following documents: “F. Maury, A. Douard, S. Delclos, D. Samelor, C. Tendero; Multilayer chromium based coatings grown by atmospheric pressure direct liquid injection CVD, Surface and Coatings Technology, 204 (2009) 983-987” (reference [1]), “A. Douard, F. Maury; Nanocrystalline chromium-based coatings deposited by DLI-MOCVD under atmospheric pressure from Cr(CO)6, Surface and Coatings Technology, 200 (2006) 6267-6271” (reference [2]), WO 2008/009714 (reference [3]) and WO 2008/009715 (reference [ ]). As a reminder, the principle of the DLI-MOCVD technique is to directly introduce, into a chemical vapor deposition chamber, under continuous or pulsed conditions, a precursor of the metal to be deposited in vaporized form. With this aim, starting with a feed tank under pressure (for example at a pressure of 3×105 Pa of an inert gas), a mother solution containing at least one organometallic compound as precursor is introduced into an evaporator. It is then split up into microdroplets to form an aerosol which is flash-vaporized. Flash evaporation consists in rapidly vaporizing a compound outside the pressure and temperature conditions predicted by its saturated vapor pressure law. The evaporator is heated to a temperature such that the precursor and its solvent are vaporized, without, however, bringing about decomposition at this stage. The temperature is conveniently comprised between the boiling point of the solvent and the decomposition temperature of the precursor and, incidentally, that of the solvent, for example 200° C. The parameters for injection of the mother solution are preferably set using a computer program. They are adjusted so as to obtain a mist of very fine and numerous droplets, in order to obtain satisfactory flash evaporation under reduced pressure. The liquid injection thus constitutes a well-controlled source of organometallic precursor, which does not limit the possibilities for optimization of the parameters of the coating deposition process. This flexibility in the formulation of the mother solution makes it possible to deposit a wide variety of coatings with a single physical vapor deposition device. The composition, structure, geometry and physicochemical characteristics of a coating may in particular be broken down into numerous variants, among others for a coating such as a protective layer, an interposed layer or a liner. The vaporized mother solution is entrained by a stream of inert gas from the evaporator to the deposition zone of the reactor in which the substrate to be covered was placed. The carrier gas used is preferably preheated to the temperature of the evaporator to obtain more efficient vaporization. Said gas is inert so as not to react with the reagents present, for example by oxidizing them. Nitrogen is generally chosen for its low cost, but helium, which benefits from better thermal conductivity, or argon, which has greater protective power, may also be employed. According to the manufacturing process of the invention, the chemical vapor deposition reactor chamber is heated to a deposition temperature comprised between 300° C. and 600° C. For this temperature range, the organometallic precursor, in particular of bis(arene)metal type, decomposes without degrading the solvent, so as to prevent as much as possible the production of reaction byproducts that are liable to pollute the chamber by becoming deposited on the reactor walls or even on the substrate. If the substrate to be covered is metallic, the temperature may be limited to the resistance temperature of the metal so as to prevent it from undergoing any deformation or phase transformation. For example, when the substrate is constituted of an alloy of zirconium, the deposition temperature can be comprised between 350° C. and 550° C., or between 350° C. and 500° C. The reduced deposition temperature for the DLI-MOCVD process is an advantage when compared with other vapor chromization CVD processes which use gaseous transition metal halides and operate at temperatures which may reach 900° C. The chemical vapor deposition reactor is placed under reduced pressure, at which the main steps of the deposition are performed, from the vaporization of the mother solution up to the optional collection of the effluent obtained on conclusion of the manufacturing process. The reduced pressure is generally from a few Torr to a few tens of Torr. They are thus moderately reduced pressures with regard to the pressures from about 10−3 Torr to 10−4 Torr of industrial PVD processes which require high vacuum equipment. One of the advantages of the DLI-MOCVD deposition process for coating a nuclear component with a protective layer is that said layer can be deposited on the inner surface and/or the outer surface of the substrate coated or not with an interposed layer. Protection of the inner layer of a nuclear component via the manufacturing process of the invention is particularly advantageous when the substrate to be coated is of large size, for example in the case of a nuclear fuel tubular cladding 1 cm in diameter and about 4 m long. It partly solves the oxidation and/or hydriding problems encountered in the nuclear sector. Specifically, by way of example, during a loss of primary coolant accident (LOCA) scenario in second and third generation reactors, a zirconium alloy fuel cladding is subjected to a sudden rise in internal temperature and pressure, which lead to ballooning and to accelerated oxidation of the cladding. The combination of these two phenomena may lead to bursting of the cladding, and thus to rupture of the nuclear fuel confinement and of the fission products it contains. The inner surface of the cladding thus exposed is particularly sensitive to oxidation and to secondary hydriding, namely massive localized hydriding which results from the depletion in oxygen of the water vapor atmosphere via the confinement effect. The cladding may then become degraded by cracking during the quenching caused by the rewatering of the accident-hit core, or even during post-quenching mechanical stresses (earthquake aftershocks, handling, etc.). This degradation may possibly lead to a loss of efficiency of cooling of the fuel assemblies and progress toward an uncontrolled degraded situation (“serious accidents”). Coating of the inner surface of a nuclear fuel cladding with at least one protective layer aids in limiting, retarding or even preventing internal oxidation and/or secondary hydriding under LOCA conditions. It is also advantageous for combating secondary hydriding under nominal conditions in the event of incidental piercing of the cladding and for combating the problem of the fuel pellet-cladding interaction. The substrate onto which is deposited at least one protective layer via the manufacturing process of the invention is composed of a metal or of one of its base alloys thereof. Preferentially, the substrate is composed of zirconium or of a zirconium-based alloy. The zirconium-based alloy may comprise, by weight: from 0% to 3% of niobium; from 0% to 2% of tin; from 0% to 0.5% of iron; from 0% to 0.2% of chromium; from 0% to 0.2% of nickel; from 0% to 0.2% of copper; from 0% to 1% of vanadium; from 0% to 1% of molybdenum; from 0.05% to 0.2% of oxygen. The zirconium alloy may in particular be chosen from an alloy meeting the constraints of the nuclear field, for example Zircaloy-2, Zircaloy-4, Zirlo™, Optimized-Zirlo™ or M5™. The compositions of these alloys are such that they comprise, by weighs, for example: Zircaloy-2 alloy: 1.20% to 1.70% of Sn; 0.07% to 0.20% of Fe; 0.05% to 1.15% of Cr; 0.03% to 0.08% of Ni; 900 ppm so 1500 ppm of O; the remainder is zirconium. Zircaloy-4 alloy: 1.20% to 1.70% of Sn; 0.18% to 0.24% of Fe; 0.07% to 1.13% of Cr; 900 ppm to 1500 ppm of O; less than 0.007% of Ni; the remainder is zirconium. Zirlo™ alloy: 0.5% to 2.0% of Nb; 0.7% to 1.5% of Sn; 0.07% to 0.28% of at least one element chosen from Fe, Ni, Cr; up to 200 ppm of C; the remainder is zirconium. Optimized-Zirlo™ alloy: 0.8% to 1.2% of Nb; 0.6% to 0.9% of Sn; 0.090% to 0.13% of Fe; 0.105% to 0.145% of O; the remainder is zirconium. M5™ alloy: 0.8% to 1.2% of Nb; 0.090% to 0.149% of O; 200 ppm to 1000 ppm of Fe; the remainder is zirconium. Preferentially, the zirconium-based alloy is Zircaloy-2 or Zircaloy-4. According to a particular geometry, the nuclear component manufactured via the process of the invention may comprise an interposed layer positioned between the substrate and at least one protective layer. In this embodiment, the support is formed by the combination of the substrate and of the interposed layer. The interposed layer may serve as a diffusion barrier. When the substrate comprises zirconium or an alloy of zirconium, the interposed layer constitutes in particular a diffusion barrier which limits, indeed even prevents: the diffusion of the chromium from the external layer toward the zirconium-based internal layer, which results in an accelerated consumption of the external layer(s) in addition to its oxidation to give chromic oxide; the formation of a eutectic above 1330° C. approximately, which may potentially harm the mechanical strength of the component and its ability to be cooled. The interposed layer may be deposited onto the substrate via a wide variety of deposition methods, more particularly by performing a DLI-MOCVD deposition or by plasma-enhanced chemical vapor deposition (CVD), onto the outer surface of the substrate and/or onto its inner surface. These two methods may be used when the interposed layer is deposited onto the outer surface of the substrate. On the other hand, only deposition by DLI-MOCVD makes it possible to cover the inner surface of the substrate, generally when the interposed layer is composed of a metallic material. The operating conditions for deposition by DLI-MOCVD are then those described elsewhere in the present description. In particular, the precursor comprising an interposed material chosen from chromium, tantalum, molybdenum, tungsten, niobium or vanadium may be of the bis(arene) type according to the variants indicated in the present description. The plasma-enhanced CVD deposition method is, for its part only used for deposition onto the outer surface of the substrate, generally when the interposed layer is composed of a ceramic material. Preferentially, the interposed layer is then deposited onto the outer surface of the substrate by performing the plasma-enhanced CVD deposition using a mixture comprising least one titanium, aluminum or silicon halide and a gaseous nitrogen precursor, as illustrated, for example, by the document “S. Anderbouhr, V. Ghetta, F. Blanquet, C. Chabrol, F. Schuster, C. Bernard, R. Madar; LPCVD and PACVD (Ti, Al)N films: morphology and mechanical properties; Surface and Coatings Technology, Volume Issues 2-3, Jul. 18, 1999, pages 103-110” (reference [5]). Preferentially, the titanium, aluminum or silicon halide is a titanium, aluminum or silicon chloride. It is chosen, for example, from TiCl4, AlCl3 and SiCl4 or mixtures thereof. Preferentially, the interposed layer may comprise at least one interposed material chosen from chromium, tantalum, molybdenum, tungsten, niobium, vanadium, alloys thereof, a titanium nitride, a titanium carbonitride, a mixed titanium silicon nitride, a mixed titanium silicon carbide, a mixed titanium silicon carbonitride, a mixed titanium aluminum nitride, or mixtures thereof. The interposed material composed of a titanium nitride, a titanium carbonitride, a mixed titanium silicon nitride, a mixed titanium silicon carbide, a mixed titanium silicon carbonitride or a mixed titanium aluminum nitride is a ceramic interposed material: each of these materials is generally denoted, respectively, as TiN, TiCN, TiSiN, TiSiC, TiSiCN or TiAlN, without this implying any stoichiometry; the nitrogen, carbon, silicon and aluminum atoms generally being insertions into the titanium metal matrix. Preferentially, the thickness of the interposed layer is from 1 μm to 5 μm. According to another particular geometry, the nuclear component manufactured via the process of the invention may further comprise a liner placed on the inner surface of said support. This liner generally acts as a diffusion barrier or improves the robustness of the nuclear component with respect to possible chemical or mechanical interactions. Said at least one protective layer may be an outer protective layer which coats the outer surface of said support and/or, when the nuclear component comprises an inner volume which may or may not be open, an inner protective layer which coats the inner surface of said support coated or not coated with the liner. A nuclear component which comprises an inner volume is, for example, a nuclear fuel cladding, a guide tube, a plate fuel (the inner volume of which is not open, it nevertheless being possible for this component to be manufactured by assembling several parts that may comprise a surface onto which the inner protective layer is deposited, these surfaces forming—after assembly of the parts—the inner surface of the support covered with the protective layer), or an absorber rod. The liner does not necessarily constitute a coating: it may be a part that is assembled or fitted subsequently into the nuclear component. It may also be obtained by hot coextrusion during the manufacture of the substrate. Alternatively, the liner may be deposited, at a deposition temperature comprised between 200° C. and 400° C., onto the inner surface of the support by chemical vapor deposition of an organometallic compound (MOCVD) or DLI-MOCVD with, as precursor(s), a titanium amide and further a precursor comprising silicon, a precursor comprising aluminum and/or a liquid additive comprising nitrogen as precursor if the material of which the liner is composed comprises, respectively, silicon, aluminum and/or nitrogen. The MOCVD deposition method is described, for example, for the titanium homolog which may be chromium for this MOCVD process in the document “F. Ossola, F. Maury: MOCVD route to chromium carbonitride thin films using Cr(NEt2)4 as single-source precursor: growth and mechanism., Adv. Mater. Chem. Vap. Deposition, 3 (199) 137-143.” (reference [6]). Preferentially, during the deposition of the liner by MOCVD or DLI-MOCVD, the liquid additive comprising nitrogen is ammonia, or optionally a molecular precursor comprising a titanium-nitrogen bond. A high concentration of the liquid additive comprising nitrogen generally promotes the formation of a nitride at the expense of a carbide, the carbon of which comes from the organometallic precursors. The MOCVD or DLI-MOCVD deposition temperature may be comprised between 300° C. and 400° C. so as to promote as best as possible the proportion of the amorphous structure in the material of which the liner is composed and thus the performance of the liner as a diffusion barrier. Alternatively, the MOCVD or DLI-MOCVD deposition temperature may be between 400° C. and 550° C. or between 400° C. and 500° C., so as to increase the deposition rate. Preferentially, the material of which the liner is composed comprises titanium nitride, a titanium carbonitride, a mixed titanium silicon nitride, a mixed titanium silicon carbide, a mixed titanium silicon carbonitride, a mixed titanium aluminum nitride, or mixtures thereof. The liner generally has a thickness of from 1 μm to 10 μm. Other deposition methods may also be suitable depositing the liner, for instance CVD, plasma CVD or DLICVD deposition, as illustrated, respectively, by the documents “Jin Zhang, Qi Xue and Songxia Li, Microstructure and corrosion behavior of TiC/Ti(CN)/TiN multilayer CVD coatings on high strength steels. Applied Surface Science, 2013. 280: pages 626-631” (reference [7]), “A. Weber, C.-P. Klages, M. E. Gross, R. M. Charatan and W. L. Brown, Formation Mechanism of TiN by Reaction of Tetrakis (dimethylamido)-Titanium with Plasma-Activated Nitrogen. Journal of The Electrochemical Society, 1995. 142(6): pages L79-L82.” (reference [8]) or “Y. S. Li, S. Shimada, H. Kiyono and A. Hirose, Synthesis of Ti—Al—Si—N nanocomposite films using liquid injection PECVD from alkoxide precursors. Acta Materialia, 2006. 54 (8): pages 2041-2048” (reference [9]). In order to deposit at least one protective layer onto the support or the substrate, the manufacturing process of the invention comprises two steps in its general embodiment: a step a) of vaporization of the mother solution comprising the precursor(s) of the protective layer, a step b) of deposition of the vaporized mother solution, during which the protective layer is formed on the substrate. Step a) of the vaporization of the mother solution is preferentially performed at a vaporization temperature comprised between 120° C. and 220° C. The mother solution contains a solvent, a precursor of bis(arene) type comprising chromium; where appropriate, an additional precursor and a carbon incorporation inhibitor. The choice of the solvent contained in the mother solution generally satisfies several criteria. First, the boiling point of the solvent is less than the temperature of the evaporator in order to make possible flash evaporation in the evaporator. It does not contain any oxygen, to prevent oxidation of the deposits by cracking. It is chemically inert with respect to the precursor in solution and is liquid under standard temperature and pressure conditions, namely, according to the present description, atmospheric pressure and a temperature of 25° C. Finally, the solvent does not decompose significantly in the reactor, so as to be recovered in the effluent leaving the reactor and to avoid or limit any pollution. For these reasons among others, the solvent of the mother solution is a hydrocarbon-based solvent, i.e. it is composed solely of carbon and hydrogen. Preferably, the solvent belongs to a chemical family similar to that of the ligands of at least one precursor compound, for example the precursor of bis(arene) type comprising chromium belonging to the aromatic hydrocarbon (or arene) family. This is because, during the passage through the reactor, this precursor decomposes thermally, releasing its ligands one after the other. The reaction byproducts are thus essentially free arenes, which mix all the better with the solvent when they are chemically similar or even identical to it. As a result, the compounds collected in the effluent leaving the reactor (precursor or unconsumed reagent, byproducts of the DLI MOCVD reaction and solvent) are generally all aromatic hydrocarbons. Preferentially, the solvent is thus a monocyclic aromatic hydrocarbon, which is liquid under standard conditions, with a boiling point below 150° C. and a decomposition temperature of greater than 600° C. Even more preferentially, it is chosen from benzene, or a benzene substituted with one or more identical or different groups chosen independently from methyl, ethyl and isopropyl groups. Even more preferentially, the solvent is toluene, mesitylene (1,3,5-trimethylbenzene) or ethylbenzene. It is also possible to use a mixture of these compounds as solvent. One of the main constituents of the mother solution is a precursor of bis(arene) type comprising chromium, and, where appropriate, an additional precursor. Depending on the composition chosen for the mother solution, various protective materials described below may be deposited by decomposition of the precursors during the deposition step b). Since the protective materials contain chromium, the mother solution contains at least the solvent and the precursor of bis(arene) type comprising chromium, the concentration of which may be chosen within a wide range. This concentration has an influence mainly on the deposition rate according to step b): the more the mother solution is concentrated in precursor, the higher the growth rate of the coating. The concentration in the mother solution of the precursor of bis(arene) type comprising chromium may be comprised between 0.1 mol·L−1 and 4.4 mol·L−1 (concentration of the pure precursor), generally between 0.1 mol·L−1 and 1 mol·L−1, typically 0.35 mol·L−1. Besides the precursors of the protective layer and the solvent, mother solution may also comprise a carbon incorporation inhibitor which prevents or limits the deposition of a protective material comprising carbon: such a material may be a carbide, a mixed carbide, a carbonitride or a mixed carbonitride, these materials possibly comprising, for example, as atomic percentages, 35% of carbon and optionally 2% to 3% of oxygen often localized on the surface of the protective layer. A small amount of carbon may occasionally be deposited with the chromium during step b), without, however, forming a carbide, even in the presence of the inhibitor. The inhibitor is a nucleophilic compound, generally a chlorine-based or sulfur-based additive, free of oxygen atoms. Its decomposition temperature is greater than 500° C., which prevents or limits the heterogeneous decomposition of the aromatic ligands of the precursor of bis(arene) type comprising chromium during which, by dissociation of the metal-ligand bonds of the precursor, some of the hydrocarbon ligands decompose under the catalytic effect of the substrate and supply their carbons so as to form ceramics of carbide type. Preferentially, the inhibitor is a monocyclic aromatic hydrocarbon substituted with a thiol group or at least one chlorine, and even more preferentially the inhibitor is thiophenol (C6H5SH) or hexachlorobenzene (C6Cl6). The inhibitor may be present in a concentration equal to 1% to 10% of the molar concentration of the chromium precursor in the mother solution, for example 2%. Once the vaporization of the mother solution has been performed in step a), the deposition step b) may be performed in a hot-walled reactor conventionally used in this field and operating under reduced pressure. The reactor in its entirety is heated to the temperature required for the deposition, so that the walls, the reactive gas phase circulating in the reactor and the substrate to be covered are at the same temperature. This type of reactor is also known as an “isothermal” reactor (or “quasi-isothermal” reactor, since a few temperature gradients may possibly remain). Use may also be made of a cold-walled reactor, in which only the reactor is not at the deposition temperature but at a lower temperature. In this case, the reactor yield, determined from the consumption of precursor, is lower. According to the invention, the chemical vapor deposition reactor is at a deposition temperature comprised between 300° C. and 600° C., so that any precursor of bis(arene) type present in the mother solution decomposes without, however, the solvent undergoing degradation. This avoids the generation of byproducts that are liable to pollute the chamber by becoming deposited on the reactor walls or even on the substrate. The deposition temperature according to step b) is preferentially comprised between 350° C. and 550° C., even between 350° C. and 500° C., even more preferentially between 350° C. and 450° C., so as to avoid or further limit any deformations or phase transformations of the substrate when the latter is metallic. Alternatively, the deposition temperature according to step b) is comprised between 300° C. and 400° C., which improves the density of the protective layer and promotes its amorphous nature, and thus its resistance to oxidation, to hydriding and/or to the migration through the nuclear component of undesired material such as a fissile material. The deposition step b) is performed on the final layer of the support. For example, for deposition onto the outer surface of the support, the deposition of the protective layer is performed on the substrate or on the final interposed layer depending on whether the support contains, respectively, a substrate that is bare or coated with at least one interposed layer. After step b), the manufacturing process of the invention may comprise the following step: c) performing on said at least one protective layer at least one step chosen from a subsequent treatment step of ionic or gaseous nitridation, ionic or gaseous silicidation, ionic or gaseous carbosilicidation, ionic or gaseous nitridation followed by ionic or gaseous silicidation or carbosilicidation. This post-treatment step improves the heat resistance and the tribological properties of the protective layer, more particularly its scratch resistance. These post-treatment methods are known to those skilled in the art, and are described, example, in the document “S. Abisset, F. Maury, R. Feurer, M. Ducarroir, M. Nadal and M. Andrieux; Gas and plasma nitriding pretreatment of steel substrates before CVD growth of hard refractory coatings; Thin Solid Films, 315 (1998) 179-185” (reference [10]). Since the ionic treatment is a plasma-enhanced treatment for which the minimum deposition temperature may be about 400° C., it is applicable only to the outer surface of a protective layer. The minimum deposition temperature for a gaseous carbosilicidation and a gaseous silicidation is, respectively, about 900° C. and about 800° C. In general, the silicidation and carbidation methods use, respectively, silane or one of its homologs (SinH2n+2) and a hydrocarbon (for example CH4, C3H8, C2H2 or C2H4) and may operate, respectively, with or without a plasma a minimum temperature of about 800° C. or 400° C. In general, steps a), b) and/or c) are performed with a carrier gas so as to inject any chemical species into the chemical vapor deposition reactor. The carrier gas comprises at least one rare gas, which is usually chemically inert with respect to be various chemical species present in the reactor. The rare gas may be chosen from xenon or krypton, but preferably from nitrogen, helium or argon. The carrier gas is, for example, at a pressure comprised between 0.2 Pa and 2 Pa. Once the various steps of the manufacturing process of the invention have been performed, the gaseous effluent leaving the chemical vapor deposition reactor comprises precursor molecules, the solvent and, where appropriate, the inhibitor which have not been consumed or pyrolyzed. The effluent may also comprise free ligands dissociated from the precursor, which are of the same aromatic family as the solvent. They are incorporated into the base solvent, with which they are miscible, and act themselves as solvent. A major and unexpected advantage is that the major these compounds leaving the reactor at low temperature are monocyclic aromatic molecules, generally of chemical structure similar or identical to that of the initial compounds, namely the precursor or the solvent. It is thus advantageous to collect them. They are gaseous on leaving the reactor due to the temperature and pressure conditions, but liquid under standard conditions. The mixture thus collected will form a solution, known as daughter solution, which may be introduced into the feed tank of the reactor as a new mother solution that can be used in step a) of the manufacturing process of the invention. The recycling process resulting therefrom is described in patent application FR 1562862 (reference [11]). By virtue of its features, the manufacturing process of the present invention allows such recycling and can thus function in a closed cycle, which has many advantages reduction or even elimination of the discharging of environmentally unfriendly substances, economic saving via the optimum use of the precursors, and, as illustrated below, an increase in the hardness of the protective coating. On conclusion of the manufacturing process of the invention, the support is coated with at least one protective layer. This protective coating makes it possible among others to combat the oxidation, the hydriding and/or migration of any undesired material inside or outside the nuclear component, for instance any fissile material derived from the nuclear fuel. The protective materials that may be deposited via the manufacturing process of the invention are varied. They are described below. According to one embodiment of the invention, the mother solution contains the inhibitor and the precursor of bis(arene) type comprising chromium; such that, at a deposition temperature comprised between 300° C. and 450° C., the protective material comprising a partially metastable chromium is obtained. In this case, the inhibitor is preferably a monocyclic aromatic hydrocarbon substituted with a thiol group, for instance thiophenol. The protective layer comprising a partially metastable chromium generally has a columnar structure. The constituent columnar grains of the columnar structure may have a mean diameter of from 100 nm to 1 μm. The metastable chromium crystalline phase generally comprises chromium of centered cubic crystallographic structure according to the Pm-3n space group. Preferentially, the stable chromium crystalline phase comprises chromium of centered cubic crystallographic structure according to the Im-3m space group and the metastable chromium crystalline phase comprises chromium of centered cubic crystallographic structure according to the Pm-3n space group. For example, the metastable crystalline phase comprising chromium of centered cubic crystallographic structure according to the Pm-3n space group represents from 1 atom % to 10 atom % of the partially metastable chromium. The Im-3m and Pm-3n space groups are described, for example, on the following website: https://en.wikipedia.org/wiki/Space_group. Their structure and their proportion may be determined by X-ray diffraction (XRD). Obtaining this Im-3m+Pm-3n partially metastable polycrystalline chromium is unexpected: to the inventors' knowledge, only the manufacturing process of the invention makes it possible to deposit a protective layer of such a crystallographic structure. This crystallographic structure has a certain advantage for nuclear safety. Specifically, the metastable chromium crystalline phase of Pm-3n type disappears irreversibly above 450° C. Now, a nuclear reactor, for instance a PWR reactor, functions under nominal conditions at about 380° C. If an XRD analysis, for example during a non-destructive control (NDC), reveals a posteriori the disappearance of the chromium metastable phase of Pm-3n type in the partially metastable chromium protective layer, it may be deduced therefrom that the environment around the nuclear component obtained via the manufacturing process of the invention has been exposed, albeit briefly, to temperatures above 450° C., i.e. well above the nominal operating temperature of 380° C. Such a detection possibility, even a posteriori, of an abnormal increase in temperature constitutes a major advantage for managing the safety of a nuclear reactor. The protective layer comprising a partially metastable chromium thus acts as an internal probe of the nuclear component, without this, however, harming the overall behavior of the component in a nuclear environment. According to another embodiment of the invention, the mother solution contains the precursor of bis(arene) type comprising chromium; such that, at a deposition temperature comprised between 300° C. and 500° C., the protective material comprising an amorphous chromium carbide is obtained. In an ideal amorphous solid, of glass type, the atoms are arranged randomly forming a three-dimensional network. In the case of compounds of carbide or nitride type based on transition metals, a certain degree of short-range order may be demonstrated, for example over a distance of less than 10 nm. At this scale, the amorphous solid is of nanocrystalline nature: it is the disorientation of these domains relative to each other and the more disordered zones which separate them which gives them an amorphous structure. Amorphous chromium carbide may be denoted as “a-CrCx”: the term “a-” means amorphous, and the coefficient x indicates that the carbide does not have exactly the stoichiometry of one of the three stable chromium carbide compounds (Cr23C6; Cr7C3; Cr3C2). Its composition may be close to Cr7C3 but intermediate with that of Cr3C2. Above a deposition temperature of 500° C., chromium carbide is no longer amorphous but polycrystalline. Obtaining a protective material comprising an amorphous chromium carbide via the manufacturing process of the invention may have numerous advantages. As illustrated hereinbelow, amorphous chromium carbide is of very high hardness, for example between 22 GPa and 29 GPa, which is particularly high when the precursors are recycled on conclusion of the manufacturing process. It is moreover generally free of grain joints, which makes it an efficient diffusion barrier, even at high temperature, which has among others the effect of limiting the penetration of chromium in a zirconium-based substrate which would thus be weakened. These structural characteristics of amorphous chromium carbide may be globally preserved even after oxidation at 1100° C. followed by quenching in water. A protective layer composed of amorphous chromium carbide is thus an excellent solution for protecting a nuclear component with respect to oxidation, hydriding, the migration of any undesired material, and/or damage during the handling of the nuclear component. According to another embodiment of the invention, the mother solution contains the precursor of bis(arene) type comprising chromium, an additional precursor chosen from a precursor of bis(arene) type comprising vanadium, a precursor of bis(arene) type comprising niobium, a precursor comprising aluminum, or the mixture of these additional precursors; such that a protective material comprising a chromium alloy chosen from a chromium/vanadium alloy, a chromium/niobium alloy, a chromium/vanadium/niobium alloy or a chromium/aluminum alloy is obtained in the presence of the inhibitor or such that a protective material comprising a carbide of the chromium alloy chosen from a carbide of a chromium/vanadium a carbide of a chromium/niobium alloy, a carbide of a chromium/vanadium/niobium alloy or a carbide of a chromium/aluminum alloy is obtained in the absence of the inhibitor. The carbide of the chromium alloy results from the incorporation of carbon into one of the abovementioned chromium alloys, forming an insertion carbide: it is thus a carbide of the chromium/vanadium alloy, a carbide of the chromium/niobium alloy, a carbide of the chromium/vanadium/niobium alloy or a carbide of the chromium/aluminum alloy (preferably the mixed carbide Cr2AlC of “MAX phase” type), which may be denoted, respectively, as CrVC, CrNbC, CrVNbC, CrAlC, without this notation referring to any stoichiometry. Preferably, the chromium all or the corresponding carbide thereof are mixed alloys or mixed carbides (namely, they do not comprise any other metallic element in a significant content, for example a content of greater than 0.5 atom %) and/or they are chromium-based alloys or the corresponding carbide of the chromium-based alloy thereof. These alloys improve the ductility of the protective layer of the nuclear component. The contents of each element in the alloy or in its mixed carbide are chosen by a person skilled in the art so as to obtain the mechanical properties, including the ductility, which are desired in a nuclear environment. For example, the atomic content of vanadium or niobium in these alloys may be comprised between 10% and 50%. In general, the mole ratios between the precursor of bis(arene) type comprising chromium and the precursor of bis(arene) type comprising vanadium or the precursor of bis(arene) type comprising niobium are similar or correspond, respectively, to the stoichiometric ratio in the chromium alloy between chromium and vanadium or niobium. Preferentially, the chromium alloy is metastable or partially metastable, i.e. it is totally or partly formed from a metastable crystalline phase, which is promoted by reducing the deposition temperature. According to another embodiment of the invention, the mother solution contains the precursor of bis(arene) type comprising chromium, a liquid precursor comprising nitrogen as additional precursor being present in the mother solution or a gaseous precursor comprising nitrogen being present in the chemical vapor deposition reactor; such that the protective material comprising a chromium nitride is obtained in the presence of the inhibitor or such that the protective material comprising a chromium carbonitride is obtained in the absence of the inhibitor. In a nuclear fuel cladding, a protective material comprising a chromium nitride and/or a chromium carbonitride may further have the advantage of combating the interaction between the inner surface of the cladding (in particular of the substrate) and the nuclear fuel pellet, known as the pellet-cladding interaction. With this aim, the deposition of the protective layer may be completed by a passivation step directed toward partially oxidizing the protective layer, for example by placing it in contact with oxygen or water. Preferentially, the deposition temperature according to step b) is comprised between 300° C. and 400° C., even more preferentially between 300° C. and 500° C., for example 480° C., such that the chromium nitride is amorphous. Once all the inhibitor has been consumed to avoid the incorporation of carbon during the deposition of the chromium nitride, a chromium carbonitride may be formed with the remaining precursors. A mixture of nitride and carbonitride is then obtained. Preferentially, the chromium nitride or chromium carbonitride is CrN, Cr2N or Cr2(C,N). The nature of the deposited nitride may depend on the ratio R between the partial pressure of the precursor comprising nitrogen and that of the precursor of bis(arene) type comprising chromium: for a given temperature, the hexagonal phase of the nitride Cr2N is preferentially obtained for low values of R and the cubic phase of the nitride CrN is obtained for the highest values. A person skilled in the art will also be able to vary the deposition temperature or the deposition pressure so as to promote the production of CrN or Cr2N. According to another embodiment of the invention, the mother solution contains the precursor of bis(arene) type comprising chromium, a precursor comprising silicon as additional precursor; such that, at a deposition temperature comprised between 450° C. and 500° C., the protective material comprising a mixed chromium silicon carbide is obtained. Where appropriate, the protective material comprising a mixed chromium silicon carbide may be included in the composition of an interphase layer positioned between a layer of metallic nature and a layer of ceramic nature (for example of a composite material such as SiC/SiC) so as to promote the adhesion between these two layers of different nature. Preferentially, the mixed chromium silicon carbide is amorphous, which may be promoted, for example, by an atomic percentage of silicon similar to chemical doping (for example an atomic percentage comprised between 1% and 3%) which retards the crystallization of the protective material and preserves the microstructure of the amorphous chromium carbide. As illustrated hereinbelow, the amorphous mixed chromium silicon carbide has good durability. Preferentially, the amorphous mixed chromium silicon carbide has the general formula CrxSiyCz, with the stoichiometric coefficients “x” comprised between 0.60 and 0.70, “y” comprised between 0.01 and 0.05, and “z” comprised between 0.25 and 0.35. The mixed chromium silicon carbide may be of the “MAX phase” type. The MAX phases herein are ternary carbides defined by the formula Mn+1AXn, in which M is chromium, A is silicon and X is carbon. This class of materials is characterized by a hexagonal crystal structure containing a stack of nanometric layers, and a small proportion of nonmetallic atoms (25%, 33% and 37.5% when n is equal to 1, 2 and 3, respectively). These materials have both a metallic nature and properties similar to those of ceramics. The mixed chromium silicon carbide of “MAX phase” type according to the invention preferably contains the silicon atom in an atomic percentage comprised between 15% and 30%. Preferentially, it is chosen from a mixed carbide of formula Cr2SiC, Cr3SiC2, Cr5Si3C2 or mixtures thereof, these carbides comprising, as atomic percentages, 25% of silicon, 17% of silicon, 30% of silicon and 25% of aluminum. According to another embodiment of the invention, the mother solution contains the precursor of bis(arene) type comprising chromium, a precursor comprising silicon as additional precursor, a liquid precursor comprising nitrogen as additional precursor being present in the mother solution or a gaseous precursor comprising nitrogen being present in the chemical vapor deposition reactor; such that, at a deposition temperature comprised between 450° C. and 550° C., the protective material comprising a mixed chromium silicon nitride is obtained in the presence of the inhibitor or such that the protective material comprising a mixed chromium silicon carbonitride is obtained in the absence of the inhibitor. Preferentially, the mixed chromium silicon carbonitride or the mixed chromium silicon nitride is amorphous, this structure being favored in particular by a moderate deposition temperature. The mixed chromium silicon nitride may have the general formula Crx′Siy′Nw′, with the stoichiometric coefficients “x′” comprised between 0.23 and 0.57, “y′” comprised between 0.003 and 0.240, and “w′” comprised between 0.42 and 0.56. The mixed chromium silicon carbonitride may have the general formula Crx″Siy″Cz″Nw″, with the stoichiometric coefficients “x″” comprised between 0.20 and 0.56, “y″” comprised between 0.005 and 0.220, “z″” comprised between 0.05 and 0.34 and “w″” comprised between 0.06 and 0.50. To obtain the protective material comprising a mixed chromium silicon carbide, a mixed chromium silicon nitride or a mixed chromium silicon carbonitride, the molar percentage of the precursor comprising silicon may be from 10% to 90% in the mother solution, preferentially from 10% to 25%, for example 15%. The molar percentage is defined here as the ratio between the number of moles of the precursor comprising silicon/(the sum of the number of moles of the precursor comprising silicon and of the precursor of bis(arene) type comprising chromium). Whether it is gaseous or liquid, the precursor comprising nitrogen is generally present in excess amount in the chemical vapor deposition reactor. Its concentration is, for example, 100 to 200 times greater than that of the precursor of bis(arene) type comprising chromium. According to another embodiment of the invention, the mother solution further contains, as additional precursors, at least one precursor of bis(arene) type comprising an addition element chosen from yttrium, aluminum, vanadium, niobium, molybdenum, tungsten, a precursor comprising aluminum or yttrium as addition elements or mixtures thereof; such that the protective material is doped with the addition element. This doping may concern any protective material mentioned in the present description. When the protective material is doped with the addition element, it generally preserves the microstructure of the corresponding undoped protective material within which the addition element is usually in the form of an insertion element or even, in certain cases, in the form of a substitution element, for instance a substitution of chromium for silicon. Preferentially, the protective material comprises the addition element in a content of from 1 atom % to 10 atom %. This content increases with the deposition temperature and the molar percentage in the mother solution of the precursor of bis(arene) type comprising an addition element which is generally greater than the atomic percentage of the addition element in the protective material. The precursors intended to obtain each protective material described previously may have a variable composition. Preferentially, the element M chosen from chromium, vanadium, niobium or the addition element is present, respectively, in the precursor of bis(arene) type comprising chromium, the precursor of bis(arene) type comprising vanadium, the precursor of bis(arene) type comprising niobium or the precursor of bis(arene) type comprising the addition element. Each of these precursors of bis(arene) type thus comprises the corresponding element M. Preferably, the element M is in the oxidation state zero (M0) so as to have a precursor of bis(arene) type comprising the element M0. More particularly, the element M is the chromium or the addition element which is present, respectively, in the precursor of bis(arene) type comprising chromium or in the precursor of bis(arene) type comprising the addition element. Similarly, the element M chosen from chromium, tantalum, molybdenum, tungsten, niobium, or vanadium present in the precursor of bis(arene) type comprising an interposed material chosen from chromium, tantalum, molybdenum, tungsten, niobium, or vanadium is preferably in the oxidation state zero (M0) so as to have a precursor of bis(arene) type comprising the element M0. In the precursor organometallic compound of bis(arene) type, the element M0 is “sandwich” complexed by the organic ligands, namely the substituted or unsubstituted arene groups. Since the element M0 has the same oxidation state as in the protective coating that is deposited (the mixed or non-mixed carbides, nitrides, carbonitrides generally being insertion compounds, the element M0 preserves the oxidation state zero therein), the precursors of bis(arene) type usually decompose without a complex reaction, for instance a redox reaction generating numerous byproducts. Preferentially, the precursor of bis(arene) type comprising the element M0 is a precursor of bis(arene) type free of oxygen atoms, of general formula (Ar) (Ar′) M0 in which Ar and Ar′, which may be identical or different, each represent, independently of each other, an aromatic group of benzene type or benzene type substituted with at least one alkyl group. Since the stability of the metal-ligand bond increases substantially with the number of substituents on the benzene nucleus, it is advantageous to choose a precursor in which Ar and Ar′ represent two sparingly substituted aromatic ligands. Preferentially, the aromatic groups Ar and Ar′ then each represent a benzene ligand or a benzene ligand substituted with one to three identical or different groups chosen from a methyl, ethyl or isopropyl group. The precursor of bis(arene) type comprising the element M0 may thus be chosen from at least one compound of formula M0(C6H6)2), M0(C6H5Et)2, M0(C6H5Me)2 or M0(C6H5iPr)2. It then decomposes from about 300° C. If it has a decomposition temperature of greater than 600° C., it is generally not selected so as to avoid decomposition of the solvent and to limit the formation of byproducts. Preferably, the precursor of bis(arene) type comprising the element M0 has the formula M0(C6H5Et)2, since its liquid state under the conditions of the vaporization step a) significantly decreases the amount of solvent in the mother solution and thus to increase the deposition rate during step b). By way of example, when the metal is chromium, the precursor may be a sandwich chromium compound, such as bis(benzene)chromium (known as BBC, of formula Cr(C6H6)2), preferably bis(ethylbenzene)chromium (known as BEBC, of formula Cr(C6H5Et)2), bis(methylbenzene)chromium (of formula Cr(C6H5Me)2) and bis(cumene)chromium (of formula Cr(C6H5iPr)2) or a mixture thereof. The precursor of bis(arene) type comprising the element M0 may also be an asymmetric derivative of formula (Ar)(Ar′)Cr, where Ar and Ar′ are different; or a mixture of these bis(arene)chromium compounds which may be rich in one of these compounds. Only BBC exists in the form of a powder. It may be injected in the form of a solution, but the concentration is then rapidly limited by its low solubility in hydrocarbon-based solvents. The other precursors cited are liquid and may be injected directly without solvent, but this does not make it possible to satisfactorily control the microstructure of the deposits. Their use in solution is preferred, as this allows wide variation in the concentration of said solution, and better adjustment of the injection conditions and consequently of the physical properties. Advantageously, the mother solution may contain various precursors, without having a negative impact on the manufacturing process of the invention. In particular, the exact nature of the aromatic ligands of the metal is not critical, provided that these ligands preferably belong to the same chemical family of sparingly substituted monocyclic aromatic compounds. The byproducts of the DLI-MOCVD reaction that are derived from the initial reagents may thus be reintroduced into the chemical vapor deposition reactor, even if the products collected at the reactor outlet have structural variations among themselves. The purity of the initial mother solution is not a critical point either, which makes it possible to use commercial solutions which may contain, for example, up to 10% of derived compounds. As it is possible to recycle these derived compounds in the process itself, the recycled mother solutions which will be used for a subsequent deposition will contain various bis(arene)s as precursors. The mother solution may thus contain a mixture of several different precursors of general formulae (Ar) (Ar′) M0. Such properties improve the capacities for industrialization of the manufacturing process of the invention. As regards the precursors comprising a nitrogen atom, generally, the liquid precursor comprising nitrogen is hydrazine and/or the gaseous precursor comprising nitrogen is ammonia. In general, and unless otherwise indicated, the concentration in the chemical vapor deposition reactor of ammonia or hydrazine is, respectively, from 1 to 10 times or 1 to 500 times greater than the concentration of the precursor of bis(arene) type comprising chromium. The precursor comprising silicon may, for its part, be an organosilane compound which preferably comprises a group of phenylsilane or alkylsilane type. Preferentially, the precursor comprising silicon is then chosen from diphenylsilane (C6H5)2SiH2, monophenylsilane (C6H5)SiH3, diethylsilane (C2H5)2SiH2, triethylsilane (C2H5)3SiH, or mixtures thereof. The precursor comprising aluminum may be a tri(alkyl)aluminium of formula AlR3, in which R is chosen, for example, from CH3, C2H5 and C(CH3)3. The precursor comprising yttrium may be an yttrium alkyl complex, particularly in which the alkyl group has large steric hindrance, more particularly a tri(alkyl)yttrium of formula YR3. It may thus be tris(cyclopentadienyl)yttrium (denoted YCp3) in which R══C5H5. The precursor comprising yttrium may also be an yttrium amigo complex, more particularly a tri(amido)yttrium of general formula YL3. It may thus be tris[N,N-bis(trimethylsilyl)amide]yttrium. In the light of the foregoing, the protective layer deposited onto the support, among others onto the substrate not coated with an interposed layer, thus has a wide variety of compositions. As an example, the nuclear component may comprise: a zirconium-based substrate and a protective layer comprising partially metastable chromium (denoted Cr/Zr), or; a zirconium-based substrate and a protective layer comprising amorphous chromium carbide (denoted CrC/Zr), or; a molybdenum-based substrate and a protective layer comprising amorphous chromium carbide (denoted CrC/Mo), or; a zirconium-based substrate and a protective layer comprising mixed chromium silicon carbide CrxSiyCz (denoted CrxSiyCz/Zr). The composition may also vary within a protective layer itself. Thus, a protective layer may have a composition gradient: when the protective material is composed of several chemical elements, for example elements A and B for a material of formula AxBy, a composition gradient consists in varying the atomic percentage x or y of at least one element A or B according to the thickness of the coating. This variation may be continuous or in stages. In practice, to vary this atomic percentage, two mother solutions which are more or less rich in element A or B may be injected simultaneously at a variable flow rate during the growth of the protective layer using two injectors. A protective layer with a composition gradient is particularly advantageous as interposed layer since it gradually accommodates the mechanical properties of the protective coating to be more coherent with the corresponding properties of the materials with which it forms a junction. The protective material deposited in a composition gradient is, for example, the alloy of chromium or its carbide, for example a chromium/vanadium alloy. The protective coating comprising at least one protective layer may also have a variety of structures, of physicochemical characteristics and/or of geometries within one or more layers. Thus, the protective layer may have a columnar or equiaxed structure. Preferably, the structure is equiaxed when the protective layer has a diffusion-barrier function. The diffusion-barrier property of a protective coating is influenced by its microstructure. Thus, usually, the anti-diffusion performance of an exogenous chemical element increases in the following order: columnar<equiaxed<amorphous. In the columnar structure, the grains of the protective material are elongated, since they grow in a preferential direction, generally after the protective layer reaches a critical thickness, for instance a few tens to hundreds of nanometers. The columnar structure is also favored by secondary growth by epitaxy, generally at high temperature and by making the stack grow very rapidly. On the other hand, in the equiaxed structure, the grains of the protective material also grow in all their preferential growth directions. The protective layer may also have a particular density. It is, for example, comprised between 90% and 100% (preferably 95% to 100) of the density in solid form of the stable metallic chromium of centered cubic crystallographic structure according to the Im-3m space group. Preferentially, said at least one protective layer has a hardness comprised between 15 GPa and 30 GPa, preferably between 20 GPa and 29 GPa. Generally, the denser a protective layer is, the more efficient it is as a diffusion barrier, for example for preventing the diffusion of oxygen, hydrogen or any undesired material. The density of a protective layer is generally higher when it forms part of a multilayer coating, or is formed at a reduced deposition rate, for example by working at reduced temperature or pressure. The geometry of the protective coating comprising the protective layers may also be variable. Each of said at least one protective layer may have a thickness of from 1 μm to 50 μm, even more preferentially from 1 μm to 25 μm, or even from 1 μm to 15 μm. Alternatively, said at least one protective layer may have a thickness of from 10 μm to 50 μm: this protective layer of minimum thickness promotes the resistance to oxidation. Preferentially, the cumulative thickness of said protective layers is from 2 μm to 50 μm. The nuclear component may comprise from 1 to 50 protective layers. The protective coating may also be a monolayer or multilayer coating. Several identical or different protective layers of composition may form, respectively, a homogeneous multilayer protective coating or a heterogeneous multilayer protective coating. Preferentially, the homogeneous multilayer protective coating generally disfavors the production of a columnar structure. It may comprise protective layers composed of the partially metastable chromium (its density promoting the diffusion-barrier role of the homogeneous multilayer coating) or of the chromium alloy, for example with a thickness of about 400 nm, or even from 100 nm to 400 nm. A multilayer material differs from a monolayer material of overall equivalent chemical composition, among others by the presence of an interface between the layers. This interface is such that it generally corresponds to a perturbation of the microstructure at the atomic scale. It is identified, for example, by means of a fine characterization technique such as high-resolution transmission electron microscopy (TEM), EXAFS (Extended X-Ray Absorption Fine Structure) spectroscopy, a Castaing microprobe which gives a composition profile, or even a posteriori by the optional presence of oxide(s) since the interface zone is a preferential oxidation pathway. A multilayer material is generally obtained via a process which performs a sequenced deposition of each monolayer. In order to deposit the homogeneous multilayer protective coating, the deposition of each layer may be separated by a waiting time, for example between 1 minute and 10 minutes, so as to purge the chemical vapor deposition reactor. By way of example, when the protective coating is of the heterogeneous multilayer type, it may comprise protective layers composed of: chromium and amorphous chromium carbide (denoted Cr/CrC), or chromium and chromium nitride (denoted Cr/CrN), or amorphous chromium carbide and chromium nitride (denoted CrC/CrN), or chromium, amorphous chromium carbide and chromium nitride (denoted Cr/CrC/CrN), or chromium and mixed chromium silicon carbide (denoted Cr/CrxSiyCz), or mixed chromium silicon carbide and amorphous chromium carbide (denoted CrxSiyCz/CrC), or mixed chromium silicon carbide and chromium nitride (denoted CrxSiyCz/CrN), or chromium and mixed chromium silicon nitride (denoted Cr/CrSixNy), or mixed chromium silicon nitride and amorphous chromium carbide (denoted CrSixNy/CrC), or mixed chromium silicon nitride and chromium nitride (denoted CrSixNy/CrN), or mixed chromium silicon carbide and mixed chromium silicon nitride (denoted CrxSiyCz/CrSixNy). In order to manufacture a protective coating comprising several layers of different composition, one possibility is to use as many injectors as there are layers of different composition, each injector making it possible to introduce a mother solution of specific composition into the chemical vapor deposition reactor. When the heterogeneous multilayer protective coating comprises a chromium protective layer, the last protective layer of the coating is generally a protective layer of ceramic nature (carbide, nitride or carbonitride, which may or may not be mixed). The nuclear component is generally a component of the nuclear reactor core, more particularly a nuclear fuel cladding, a spacer grid, a guide tube, a plate fuel or an absorber rod. Preferentially, the nuclear fuel cladding is in the form of a tube or a plate. The plate may result more particularly from the assembly of two subunits. The invention also relates to a nuclear component which is obtained or obtainable by the manufacturing process of the invention. The invention also relates to a nuclear component comprising: i) a support containing a substrate based on a metal chosen from zirconium, titanium, vanadium, molybdenum or base alloys thereof, the substrate being coated or not coated with an interposed layer placed between the substrate and at least one protective layer; ii) said at least one protective layer coating said support and composed of a protective material comprising chromium chosen from a partially metastable chromium comprising a stable chromium crystalline phase and a metastable chromium crystalline phase, an amorphous chromium carbide, a chromium alloy, a carbide of a chromium alloy, a chromium nitride, a chromium carbonitride, a mixed chromium silicon carbide, a mixed chromium silicon nitride, a mixed chromium silicon carbonitride, or mixtures thereof. These two types of nuclear component, namely the nuclear component according to the invention and the nuclear component which is obtained or obtainable by the manufacturing process of the invention, may be in one or more of the variants described in the present description for the manufacturing process of the invention, among others the variants relating to the structure, the physicochemical characteristics (for instance the density), the geometry and/or the composition of the nuclear component. These variants relate among others, but not exclusively, to: the substrate, the interposed layer, the liner, the protective layer including its combination with a particular substrate, the protective coating or the type of nuclear component as are detailed in the present description, among others the description of the manufacturing process of the invention. For these two types of nuclear component, the protective material comprising chromium is more particularly the amorphous chromium carbide. The nuclear component thus can comprise: i) a support containing a substrate based on a metal chosen from zirconium, titanium, vanadium, molybdenum or base alloys thereof, the substrate being coated or not coated with an interposed layer placed between the substrate and at least one protective layer; ii) the at least one protective layer coating the support and composed of a protective material comprising chromium which is an amorphous chromium carbide. For these two types of nuclear component, the particular embodiments of the invention, optionally taken in combination, may be such that: the substrate is composed of zirconium or of a zirconium-based alloy. the interposed layer comprises at least one interposed material chosen from chromium, tantalum, molybdenum, tungsten, niobium, vanadium, alloys thereof, a titanium nitride, a titanium carbonitride, a mixed titanium silicon nitride, a mixed titanium silicon carbide, a mixed titanium silicon carbonitride, a mixed titanium aluminum nitride, or mixtures thereof; the thickness of the interposed layer is from 1 μm to 5 μm; the nuclear component further comprises a liner placed on the inner surface of the support; the at least one protective layer an outer protective layer which coats the outer surface of the support and/or, when the nuclear component comprises an inner volume which may or may not be open, an inner protective layer which coats the inner surface of the support coated or not coated with the liner; the material of which the liner is composed comprises a titanium nitride, a titanium carbonitride, a mixed titanium silicon nitride, a mixed titanium silicon carbide, a mixed titanium silicon carbonitride, a mixed titanium aluminum nitride, or mixtures thereof; the liner has a thickness of from 1 μm to 10 μm; the protective material is doped with an addition element chosen from yttrium, aluminum, vanadium, niobium, molybdenum, tungsten, or mixtures thereof; the protective material comprises the addition element in a content of from 1 atom % to 10 atom %; the nuclear component comprises: a zirconium-based substrate and a protective layer comprising amorphous chromium carbide, or; a molybdenum-based substrate and a protective layer comprising amorphous chromium carbide, or; a molybdenum-based substrate and a protective layer comprising amorphous chromium carbide, or. the at least one protective layer has a composition gradient; the at least one protective layer has an equiaxed structure; the at least one protective layer has a density comprised between 90% and 100% of the density in solid form of the stable metallic chromium of centered cubic crystallographic structure according to the Im-3m space group; the at least one protective layer has a hardness comprised between 15 GPa and 30 GPa; each of the at least one protective layer has a thickness of from 1 μm to 50 μm, or even from 10 μm to 50 μm; several protective layers of identical or different composition form, respectively, a homogeneous multilayer protective coating or a heterogeneous multilayer protective coating; the heterogeneous multilayer protective coating comprises protective layers composed of: chromium and amorphous chromium carbide, or amorphous chromium carbide and chromium nitride, or chromium, amorphous chromium carbide and chromium nitride, or mixed chromium silicon carbide and amorphous chromium carbide, or mixed chromium silicon nitride and amorphous chromium carbide. the nuclear component is a nuclear fuel cladding, a spacer grid, a guide tube, a plate fuel or an absorber rod. The invention also relates to the use of these types of nuclear component for combating oxidation and/or hydriding in a humid atmosphere comprising water, in particular in the form of water vapor. The invention also relates to the use of these types of nuclear component, for combating hydriding in an hydrogenated atmosphere comprising hydrogen, in particular an hydrogenated atmosphere comprising more than 50 mol % of hydrogen and/or water in addition, in particular in the form of water vapor (for example at a minimum of 25 mol % of water, among others 25% to 50%). The humid atmosphere may comprise from 25 mol % to 100 mol %, preferentially from 50 mol % to 100 mol % and even more preferentially from 75 mol % to 100 mol % of water vapor. The humid atmosphere or the hydrogenated atmosphere may further comprise an additional gas chosen from air, nitrogen, carbon dioxide, or mixtures thereof. Preferentially, the purpose of these uses is to combat oxidation and/or hydriding: in which the humid or hydrogenated atmosphere is at a temperature comprised between 25° C. and 1400° C., or even between 25° C. and 1600° C., more particularly a temperature comprised between 200° C. and 1300° C., even more particularly between 1200° C. and 1300° C., or even between 1300° C. and 1600° C.; or preferentially between 800° C. and 1200° C.; and/or up to at least 5000 seconds, more particularly between 1000 seconds and 5000 seconds, in particular when the temperature is comprised between 1200° C. and 1300° C.; and/or in the presence of a temperature increase rate which is comprised between 0.1° C./second and 300° C./second; and/or following quenching with water of the nuclear component, in particular in which the quenching takes place at a temperature comprised between 25° C. and 400° C. The invention also relates to a probe for temperature variation (preferably a temperature increase above 450° C.) in a nuclear environment, the probe comprising a partially metastable chromium comprising a stable chromium crystalline phase and a metastable chromium crystalline phase. As indicated previously, the use of partially metastable chromium for measuring a temperature variation among others has an advantage in the a posteriori measurement of a temperature increase above 450° C. and thus an advantage for nuclear safety. The metastable chromium crystalline phase may comprise chromium of centered cubic crystallographic structure according to the Pm-3n space group. Preferentially, the stable chromium crystalline phase comprises chromium of centered cubic crystallographic structure according to the Im-3m space group and the metastable chromium crystalline phase comprises chromium of centered cubic crystallographic structure according to the Pm-3n space group. Generally, the metastable crystalline phase comprising chromium of centered cubic crystallographic structure according to the Pm-3n space group represents from 1 atom % to 10 atom % of the partially metastable chromium. The probe may be the component of a system or of a part for non-destructive control (NDC), namely, for example, a control which characterizes the state of integrity of structures or materials, without degrading them, during their production, use or maintenance. The part may thus be of very simple structure since it is essentially its composition which gives it its properties. It thus acts as a tracer and may be left in the nuclear environment in order to be able to be readily analyzed a posteriori for detecting the temperature variation. For the sake of integration and compactness, the nuclear component of the invention may comprise the probe, among others in all the variants described for the nuclear component. In the present description of the invention, a verb such as “to comprise”, “to incorporate”, “to include”, “to contain”, “composed of” and the conjugated forms thereof are open terms and thus do not exclude the presence of additional element(s) and/or step(s) which are added to the initial element(s) and/or step(s) stated after these terms. However, these open terms further target a particular embodiment in which only the initial element(s) and/or step(s), with the exclusion of any other, are targeted; in which case, the open term further targets the closed term “to consist of”, “to be constituted of” and the conjugated forms thereof. The expression “and/or” is directed toward connecting elements for the purpose of simultaneously denoting a single one of these elements, both the elements, or even a mixture thereof or a combination thereof. The use of the indefinite article “a” or “an” for an element or a step does not exclude, unless otherwise mentioned, the presence of a plurality of elements or steps. Any reference sign in parentheses in the claims shall not be interpreted as limiting the scope of the invention. Moreover, unless otherwise indicated, the limit values are included in the ranges of parameters indicated and the temperatures indicated are considered for an implementation at atmospheric pressure. Any coating, layer or liner deposited by means of the manufacturing process of the invention can cover a part or, preferably, all of the surface on which it lies. Still in the present description, any alloy is generally a base alloy. The term “base alloy” of the metal included among others in the composition of the substrate, of the protective material or of the interposed material denotes any alloy based on the metal in which the content of the metal is at least 50% by weight of the metal of the alloy, particularly more than 90%, or even more than 95%. The base metal is, for example, zirconium, titanium, chromium, tantalum, molybdenum, tungsten, niobium or vanadium. An alloy may also contain other chemical elements (for example in a content of greater than 0.5 atom %), in particular a second metal element, and then constitutes a mixed alloy. The carbon insertion element and/or the nitrogen insertion element in an alloy form, respectively, a carbide of the alloy, a nitride of the alloy or a carbonitride of the alloy, which may also be mixed in the presence of a second metal element. The alloy is preferably able to be used in the nuclear sector and/or under irradiation. Other subjects, characteristics and advantages of the invention will now be specified in the description which follows of particular embodiments of the process of the invention, which are given by way of illustration and without limitation, with reference to the appended figures. The examples of deposition of a coating onto a substrate which follow are performed in a chemical vapor deposition reactor placed in a tube furnace (model sold by the company Carbolite). The reactor is composed of a silica tube: it is connected to the gaseous effluent inlet and evacuation system via leaktight connections and to the mother solution via an injector (Vapbox model sold by the company Kemstream). In the examples that follow, the substrate is not coated with an interposed layer. However, these examples may be transposed to the case where the substrate is formed from a substrate coated with an interposed layer. 1. Coating on a Silicon Substrate As a preliminary experiment, the manufacturing process of the invention is transposed to a silicon substrate, a square silicon wafer (side length of 1 cm and 300 μm thick) is placed in the reactor so as to be covered with a protective layer of chromium carbide (CrC), chromium (Cr) or mixed chromium silicon carbide (CrxSiyCz denoted CrSiC) using the following compounds: chromium precursor: bis(ethylbenzene)chromium, known as BEBC (Cr(C6H5Et)2), silicon precursor: bis(phenyl)silane (H2Si(C6H5)2), inhibitor: thiophenol (C6H5SH), solvent: toluene, carrier gas: N2 at a flow rate of 500 sccm, namely 500 cm3/min under standard conditions, evaporator temperature: 200° C., deposition time: 20 minutes. The other conditions for the deposition of the protective coating constituting the outer coating are indicated in Table 1, in particular the injector open time and frequency, and the temperature and pressure in the chemical vapor deposition reactor. The injection parameters (open time and frequency) act essentially on the deposition rate and thus on the deposit thickness by varying the deposition time. The open frequency is, for example, from 1 Hz to 10 Hz, typically 10 Hz. The parameters which act on the physicochemical and structural characteristics of the deposit are essentially the deposition temperature (acting in particular on the structure: amorphous or crystalline; dense or porous; equiaxed or columnar) and the composition of the injected solution. By convention, the silicon substrate covered with coating number N is referred to in Table 1 as sample N: sample 1 thus denotes the silicon substrate covered with coating 1. TABLE 1Injection conditionsTotalOpeningdepositionDepositionCoatingFrequencytimepressuretemperature(N)Injected solution(Hz)(ms)(10+3 Pa)(° C.)CrC (1)BEBC (3.5 × 10−1 M) +100.5-51.3; 6.7450; 500;toluene (50 ml)550Cr (2)BEBC (3.5 × 10−1 M) +100.5-56.7400; 450thiophenol (2 mol %) +toluene (50 ml)CrSiC (3)BEBC (3.5 × 10−1 M) +40.56.7450; 500bis(phenyl)silane(15 mol %) + toluene(50 ml) Table 2 collates the result of the structural analyses of these coatings, and also the result of the heat resistance tests detailed in the following examples (thermal stability: temperature at and above which a crystalline phase appears in XRD in situ under argon) The atomic composition expressed as atomic percentage is measured with a Castaing microprobe (known as EPMA, according to the English acronym for “Electron Probe Microanalysis”). Variation of the injection parameters gives several coating thicknesses, which are generally between 0.5 μm and 10 μm: the thicknesses selected for the oxidation tests that follow are indicated in Table 2. As regards their structure, Table 2 shows that the CrC and CrSiC coatings are amorphous (no crystalline structure), the CrC coating nevertheless being polycrystalline for a deposition temperature of 550° C. As regards the Cr coating, it comprises two crystalline phases: a main phase composed of centered cubic (bcc) chromium crystals in the Im-3m space group (lattice parameter a=2.88 Å) and a phase in very minor proportion composed of chromium crystals of centered cubic crystallographic structure in the Pm-3n space group (lattice parameter a=4.59 Å) when the DLI-MOCVD deposition temperature is less than or equal to 450° C. TABLE 2DepositionThermalCoatingThicknesstemperatureAtomicstability(N)(μm)(° C.)Structurecomposition(° C.)CrC (1)2; 5; 9400AmorphousCr0.64C0.33O0.03Cr7C3 580500Cr3C2 590Cr2O3 610550PolycrystallineCr0.61C0.32O0.07Cr (2)4; 6400Polycrystalline,Cr0.69C0.12S0.08O0.11Metastable450two-phased:Cr phasemain phasedisappearsCr bcc (Im-3m);at 450° C.metastablephaseCr bcc (Pm-3n).Multilayerstructureto increase thedensity.CrSiC (3)4450AmorphousCr0.66Si0.02C0.29O0.03Cr7C3 750500CrSi2 7502. Evaluation of the Heat Resistance The coatings must withstand the temperature under an oxidative atmosphere and act as a diffusion barrier. Their heat resistance under an inert atmosphere is first studied by XRD in situ as a function of the temperature under Ar. The temperature at which a structural change appears (crystallization, phase transformation) is measured. Such structural changes are liable to accelerate the diffusion of oxygen through the coating to the substrate, via the grain joints or microcracks. The coating must therefore be able to maintain its structural integrity at the highest possible temperatures. The heat resistance results obtained for the wafers coated in the preceding example are collated in Table 2. 2.1. Heat Resistance of an Amorphous Si/CrC Wafer Under an Inert Atmosphere A silicon substrate passivated with a layer of a few hundred nanometers composed of amorphous silicon nitride (generally Si3N4), which is commercially available, is used to avoid any diffusion with the protective layer deposited on its surface. Alternatively, a bare silicon substrate may be passivated by thermal CVD with a mixture (SiH4+NH3) or by plasma (PECVD) with NH3. The passivated silicon substrate is then coated with amorphous chromium carbide according to the conditions of Table 1 (specific conditions: injection at 10 Hz, deposition temperature=500° C., deposition pressure=6700 Pa). In a chamber flushed with a stream of argon to limit or even prevent any contamination with oxygen, this sample is heated from 30° C. to 800° C. at a temperature increase rate of 5° C./minute while in parallel being analyzed by X-ray diffraction (XRD). After natural cooling to 30° C., a final X-ray spectrum is recorded. Analysis of the X-ray spectra obtained shows that the sample maintains an amorphous structure up to 570° C. It then crystallizes in the form of Cr7C3 at about 580° C., and then Cr3C2 at about 590° C., and finally becomes oxidized to Cr2O3 at about 610° C. due to traces of oxygen present on account of an airtightness defect of the reactor: these three phases remain up to 800° C. with a proportion of Cr2O3 which increases with the temperature. These results indicate that the amorphous chromium carbide coating preserves its physical integrity up to 300° C. under an inert atmosphere of argon. 2.2. Heat Resistance of a Partially Metastable Si/Cr Wafer Under an Inert Atmosphere A silicon substrate passivated as previously is coated with partially metastable chromium according to the conditions of Table 1 (specific conditions: deposition temperature=400° C., deposition pressure=6700 Pa). It is then placed in the reactor flushed with a stream of argon, in which it is gradually heated from 30° C. to 600° C. at a temperature increase rate of 1° C./second. In parallel, it is analyzed by XRD (acquisitions of 35 minutes each at 30° C. and every 50° C. from 350° C.) Analysis of these spectra shows that the partially metastable chromium is polycrystalline: it contains a stable phase (Im-3m) and a metastable phase (Pm-3n). The metastable phase (Pm-3n) disappears after 450° C. In parallel, the stable phase (Im-3m) remains throughout the heat treatment, the corresponding XRD peaks become finer and more intense by virtue of the improvement in the crystallinity. Above 550° C., chromium oxide Cr2O3 appears as a result of the traces of oxygen present in the in situ analysis chamber. On conclusion of the temperature cycle, the sample is cooled naturally to 30° C.: the metastable phase is no longer present, since it has undergone an irreversible transformation. Only the stable phase remains, along with chromium oxides in small proportion. These results indicate that the chromium metal coating has a heat resistance under an inert atmosphere in accordance with conventional chromium of centered cubic (bcc) structure which is stable under normal conditions. Moreover, it is seen that the minor phase of metastable chromium becomes transformed at about 450° C. to form the main stable chromium phase: this is coherent with the literature which states that metastable chromium is transformed into stable chromium after 450° C. 2.3. Heat Resistance of an Amorphous Si/CrSiC and Steel/CrSiC Wafer Under an Inert Atmosphere A passivated silicon substrate is coated with an amorphous mixed chromium silicon carbide according to the conditions of Table 1 (specific conditions: mole ratio of diphenylsilane to BEBC=15%, deposition temperature 450° C.), and is then placed in the reactor flushed with a stream of argon. This sample is heated from 30° C. to 750° C. at a temperature increase rate of 1° C./second while in parallel being analyzed by XRD (acquisitions every 50° C. from 400° C.) In a similar manner to these operating conditions, a 304L stainless-steel substrate coated with an amorphous mixed chromium silicon carbide at a deposition temperature of 500° C. is also analyzed by XRD (acquisitions every 50° C. from 600° C.) during heating from 30° C. to 1100° C. Comparative analysis of the XRD spectra for these two samples shows similar behavior: the amorphous nature disappears at about 750° C. by crystallization of a chromium carbide Cr2C3 and of the compound CrSi2. Furthermore, it shows that since CrSi2 appears on the two substrates, this compound is formed with the silicon of the coating and not that of the substrate, the silicon nitride passivation layer applied previously to the surface of the silicon substrate constituting a diffusion barrier. Above 750° C., other crystalline phases appear: Cr2O3 after 850° C. and Cr3C2 after 1000° C. It is thus noted that the addition of Si to the chromium carbide coating delays its crystallization toward higher temperatures (750° C. instead of 580° C.) 3. Evaluation of the Oxidation in Air 3.1. Resistance to Oxidation in Air at 25° C. of an Amorphous Zr/CrC Wafer In order to evaluate its resistance to mild oxidation at room temperature, a wafer based on the zirconium alloy Zircaloy-4 (dimensions: 45 mm×14 mm×1 mm) is coated with a single 4 μm protective layer of amorphous chromium carbide in accordance with the deposit:ion conditions indicated in Table 1 for coating 1 (deposition temperature=450° C., deposition pressure=6700 Pa). It resides in air at room temperature for a few days, which results in very mild surface oxidation of the chromium carbide coating due to the passage into the ambient air, as for the majority of metals and alloys. The elemental composition by mass according to the depth in this sample is then determined for the elements zirconium, chromium, oxygen and carbon by glow discharge spectrometry (GDS). It is shown in FIG. 1 and reveals the presence of a nonzero oxygen content at the interface between the substrate and the coating. The mean coating composition values found via this technique are in agreement with those measured by microprobe (EPMA, Electron Probe Micro-Analyzer): as weight percentages relative to the total mass of the wafer, 90% of Cr, slightly less than 10% of C and a few % of O, which corresponds in atomic percentages to a composition close to Cr7C3. The Transmission Electron Microscopy (TEM) images shown in FIGS. 2A and 2B are acquired using a microscope (JEM 2100 sold by the company Jeol) equipped with a field emission gun operating at 200 kV: the zones delimited on these images confirm this result of the profile of FIG. 1, among others the presence of a layer of zirconium oxide ZrO2 (zone b) with a thickness of less than 400 nm, which is at the interface between the non-oxidized substrate (zone a) and the coating (zone c). These results are confirmed by complementary analyses Energy Dispersion Spectroscopy (EDS) which indicate a composition, atomic percentages, of about 40% of O and 60% of Zr on the oxidized interfacial zone as opposed to 5% of O and 95% of Zr on the non-oxidized zone of the substrate. The formation of a passivation layer at the surface of the Zircaloy-4 is normal in the absence of specific surface stripping. This layer may possibly disappear in the course of the oxidation and temperature increase experiments. The Zircaloy-4 substrate was already surface-oxidized on receipt. No other source of oxidation could be identified during the phase of deposition of the chromium carbide protective layer. The sample thus coated underwent no additional oxidation on conclusion of storage in air under the ambient temperature conditions. 3.2. Resistance to Oxidation in Air at 800° C. of an Amorphous Si/CrC Wafer The silicon wafer covered with amorphous CrC manufactured according to the conditions of Example 2.1 is subjected to aging in air at 800° C. for increasing times of 15, 30, 45, 60, 90, 120 and 180 minutes. For each temperature, the sample is cooled naturally and then analyzed by XRD at room temperature. Analysis of the XRD spectra shows that, under air at 800° C.: after 15 minutes of treatment, the same three phases seen previously appear: Cr7C3, Cr3C2 and Cr2O3. The amorphous phase appears to disappear since no broad spread-out lump is visible at about 2 theta=30° C.; after 90 minutes at 800° C., there is virtually no more Cr2O3, as if the oxygen had consumed all the chromium belonging to the carbides. Residual peaks characteristic of Cr3C2 are still present, even after 180 minutes at 800° C. These results indicate that the amorphous chromium carbide coating withstands oxidation in air at 800° C. for at least 15 minutes. Beyond these conditions, it begins to be oxidized and to crystallize. 3.3. Resistance to Oxidation as a Function of the Temperature in Dry or Humid Air of Bare or Coated Zr or Mo Wafers In order to study the resistance to oxidation in dry or humid air as a function of the temperature, thermogravimetric analyses (TGA) are performed on TGA wafers (dimensions: 6 mm×4 mm×1 mm, wafers pierced with a hole 1 mm in diameter and then suspended on the beam of the TGA balance) of Zircaloy-4 or molybdenum coated by the manufacturing process of the invention with protective layers of various compositions (amorphous CrC, partially metastable Cr or amorphous CrxSiyCz) and various thicknesses (9 μm, between 5 μm and 6 μm, or between 2 μm and 3 μm). The deposition conditions are those of Table 1 for the corresponding protective layers which cover all the faces of the wafers. These conditions are completed by the following specific parameters: amorphous CrC: T=450° C. and P=6700 Pa; partially metastable Cr: T=400° C., amorphous CrxSizCy: T=500° C. These analyses consist in gradually heating each sample from 25° C. to 1200° C. at a temperature increase rate of 90° C./minute, and then in maintaining a temperature of 1200° C. They are performed with dry or humid air (27.5% relative humidity). The progress of the oxidation of each sample is evaluated by measuring its gain in mass (due to the formation of oxide). For comparative purposes, each analysis is repeated on a wafer without a protective layer. 3.3.1. Results for an Amorphous Zr/CrC Wafer Zircaloy-4 wafers of various thicknesses (2 μm, 5 μm or 9 μm) each covered with a protective layer of amorphous CrC are subjected to TGA analysis in air between 25° C. and 1200° C. For comparative purposes, an uncoated Zircaloy-4 wafer is subjected to the same analysis. As illustrated in FIGS. 3A (dry air) and 3B (humid air), during the temperature increase from 25° C. to 1200° C. on a Zircaloy-4 substrate coated with amorphous CrC, the gains in mass change in an equivalent manner in dry air or in humid air, irrespective of the thickness of the protective layer. On the other hand, in comparison, the extent of the oxidation is much less than that for the bare substrate: the coated substrate gains only about 0.15% in mass as opposed to about 3% mass for the bare substrate, i.e. 20 times more. During the temperature stage at 1200° C., the three Zircaloy-4 substrates coated with amorphous CrC become gradually oxidized over time. This slowing-down in catastrophic oxidation at and beyond which the substrate disintegrates is smaller for the humid air atmosphere which is the more oxidizing (FIG. 3D) relative to the dry air (FIG. 3C), but advantageously increases with the thickness of the protective layer. XRD analyses appear to indicate that three crystalline phases coexist in the coating: crystalline Cr2O3, amorphous chromium oxides and amorphous chromium carbide CrC. In comparison, the bare substrate becomes immediately oxidized to ZrO2. The oxygen penetration is thus total in the bare substrate, whereas it is blocked or greatly limited up to 900° C. Above this temperature, it becomes partial and gradual in the coated substrates, since the amorphous CrC coating slows down the oxidation kinetics. The amorphous CrC coating thus indeed protects against and/or delays the oxidation of the Zircaloy-4 substrate during heating from 25° C. to 1200° C., even in humid air. This beneficial effect increases with the thickness of the coating. 3.3.2. Results for an Amorphous Mo/CrC Wafer Molybdenum wafers of various thicknesses (2 μm or 5 μm) each covered with a protective layer of amorphous CrC are subjected to TGA analysis in air between 25° C. and 1200° C. For comparative purposes, an uncoated molybdenum wafer is subjected to the same analysis. As illustrated in FIGS. 4A (dry air) and 4B (humid air), during the temperature increase from 25° C. to 1200° C. on a molybdenum substrate coated with amorphous CrC, the gains in mass change less quickly in dry air than in humid air, which is a more oxidative atmosphere. Moreover, the thicker the amorphous CrC coating, the later the sample becomes oxidized and the more it is capable of withstanding high temperature: up to about 1000° C. for a 2 μm coating, or even more than 1100° C. for a 5 μm coating. The resistance to oxidation for greater thicknesses of the amorphous CrC coating could not be tested: it should nevertheless change favorably. In comparison, the bare molybdenum substrate becomes oxidized immediately in dry or humid air: above 600° C. after a gain in mass due to the formation of molybdenum oxide, the bare substrate is then total destroyed by formation of volatile oxides, which leads to rapid and negative losses of mass. Since the various samples are rapidly destroyed at a temperature of 1200° C., the corresponding isothermal TGA analysis is not illustrated. The amorphous CrC coating thus indeed protects against and/or delays the oxidation of the molybdenum substrate during heating from 25° C. to 1100° C., even in humid air. This beneficial effect increases with the thickness of the coating. 3.3.3. Results for a Partially Metastable Zr/Cr Wafer Zircaloy-4 wafers are each covered with nine protective layers of partially metastable Cr so as to form multilayer coatings of different thicknesses (4 μm or 6 μm). The stack of the nine protective layers is made by injecting the mother solution for 15 minutes and then stopping it for 5 minutes: a multilayer coating is thus obtained by repeating this deposition/pause cycle nine times. The resumption of injection after the 5-minute pause creates an interface which may be visualized by TEM analysis of the multilayer coating. This avoids the columnar growth and densifies the coatings. The coated wafers are subjected to TGA analysis in air between 25° C. and 1200° C. For comparative purposes, an uncoated Zircaloy-4 wafer is subjected to the same analysis. As illustrated in FIGS. 5A (dry air) and 5B (humid air), during the temperature increase from 25° C. to 1200° C. on a Zircaloy-4 substrate coated partially metastable Cr, the gains in mass change in an equivalent manner in dry air or in humid air, irrespective of the thickness of the protective layer. On the other hand, in comparison, the extent of the oxidation is smaller than that for the bare substrate: the substrate coated with partially metastable Cr gains only about 0.8% in mass as opposed to about 3% mass for the bare substrate (i.e. 75 times more), which is nevertheless less protective than the amorphous CrC coating analyzed previously with a gain in mass equal to 0.15%. During the temperature stage 1200° C., the two Zircaloy-4 substrates coated with partially metastable Cr become gradually oxidized over time as illustrated by FIGS. 5C (dry air) and 5D (humid air). Compared with what was found with the amorphous Zr/CrC wafer (Example 3.3.1), the multilayer coating of partially metastable Cr shows in the first seconds a delay of oxidation in comparison with the bare substrate, rapidly followed by an acceleration of this oxidation. The origins of this unfavorable acceleration at 1200° C. relative to the bare substrate have not yet been entirely elucidated. At this stage, various hypotheses are being studied with a view to optimizing the protective effect of the partially metastable Cr coating with respect to oxidation. However, advantageously, even at 1200° C., the partially metastable Zr/Cr wafer shows less substantial desquamation than the bare Zr wafer, which ensures better mechanical strength. XRD analyses indicate that, on conclusion of the temperature stage at 1200° C., all of the partially metastable Cr coating is oxidized to Cr2O3 and the substrate to ZrO2. The partially metastable Cr coating greatly prevents or greatly limits the penetration of oxygen up to about 900° C., above which temperature this penetration becomes partial and gradual. The partially metastable Cr coating thus indeed protects against and/or delays the oxidation of the Zircaloy-4 substrate. 3.3.4. Results for a Zr/CrxSiyCz Wafer A Zircaloy-4 wafer 4 μm thick covered with a protective layer of amorphous mixed chromium silicon carbide CrxSiyCz is subjected to TGA analysis in dry or humid air between 25° C. and 1200° C. with a temperature stage at 1200° C. For comparative purposes, an uncoated Zircaloy-4 wafer is subjected to the same analysis, the oxidation in dry or humid air giving, in this case, identical results. As illustrated in FIG. 6A (25° C. to 1200° C.), the gain in mass changes similarly in dry air or in humid air for the Zircaloy-4 substrate coated with amorphous mixed chromium silicon carbide CrxSiyCz: the oxidation is delayed up to 700° C. in humid air and 800° C. in dry air. In comparison with the bare substrate, the coated substrate shows better resistance to oxidation. FIG. 6B (stage at 1200° C.) logically shows better resistance to oxidation in dry air than in humid air of the Zircaloy-4 substrate coated with amorphous mixed chromium silicon carbide CrxSiyCz, without, however, improving this property relative to the bare substrate. However, advantageously, even at 1200° C., the Zr/CrxSiyCz wafer shows less substantial cracking and swelling than the bare Zr wafer, which ensures better mechanical strength. XRD analyses indicate that, on conclusion of the temperature stage at 1200° C., all of the amorphous mixed chromium silicon carbide CrxSiyCz coating is oxidized (among others to Cr2O3 for chromium) and the substrate to ZrO2. The amorphous mixed chromium silicon carbide CrxSiyCz coating thus indeed protects against and/or delays the oxidation of the Zircaloy-4 substrate. Its resistance to oxidation and its mechanical strength are of intermediate magnitude relative to those of amorphous CrC and of partially metastable Cr. 3.3.5. Evaluation of the Resistance to Oxidation and to Hydriding after Exposure of a Zr/CrC Wafer to 1100° C. Followed by Quenching in Water Zircaloy-4 wafers (dimensions: 45 mm×14 mm×1 mm) of three different thicknesses (2 μm to 3 μm, 5 μm to 6 μm, or 9 μm) each covered with a protective layer of amorphous CrC as used in Example 3.3.1 are oxidized in dry sir for 14 minutes, and then subjected to quenching in water for 10 seconds. The Scanning Electron Microscopy (SEM) images of a 9 μm section of the wafer resulting from this treatment are shown in FIGS. 7A and 7B (zoom into a corner of the wafer): although the substrate has a rough finish, i.e. it has a certain surface roughness, the coating is of uniform thickness and it perfectly covers all the unevennesses. The mechanical strength of the wafers is also preserved. The absence of degradation of the Zircaloy-4 on conclusion of this test at 1100° C. with quenching in water is coherent with the oxidation of Zircaloy-4 which increases as a function of the temperature and is significant only at and above 1300° C. (Example 3.3.1). The lower temperature (1100° C.) and the short time (14 minutes are such that the oxidation is limited here to the coating (see FIGS. 7A and 7B) which ensures good protection of the Zircaloy-4. In accordance with GDS profiles not shown herein, three main layers of different chemical composition may be identified on the SEM images of FIGS. 7A and 7B, i.e. going from the external medium to the Zircaloy-4 substrate: an outer layer of oxidized coating about 4 μm thick comprising a zone A of 2 μm of surface chromium oxides (Cr2O3) and a zone B of 2 μm of partially oxidized coating (Cr3C2 resulting from the recrystallization of the amorphous CrC during the temperature increase+Cr2O3); an intermediate layer comprising a zone C of non-oxidized coating (amorphous CrC) at the interface with the outer coating and with a zone D formed from chromium and carbon which has diffused into the Zircaloy-4 with which it is at the interface (with possible formation of cubic zirconium carbide ZrC with a lattice parameter of about 4.7 Å, in particular for the thinnest coating of 2 μm to 3 μm); a layer formed from Zircaloy-4 (zone E). The amorphous CrC coating thus protects the Zircaloy-4 substrate with respect to oxidation for the three wafers, since no formation of ZrO2 is detected. Only the coating is partially oxidized to Cr2O3. Table 3 collates the thicknesses evaluated by GDS for zones A, B, C and D for the three Zr/CrC wafers, and also the corresponding overall gains in mass. TABLE 6Zone D:Zone B:ThicknessCrCOofInitialZone A:transitionZone C:diffusionthicknessThicknesszoneThicknesszone inGain inCrCCr2O3thicknessCrCsubstratemass(μm)(μm)(μm)(μm)(μm)(mg · cm−2)91.52.53150.4145-61.52.52110.5432-31.52.5160.288 Table 3 is completed by FIG. 7C which groups the SEM images of a cross section for each of the three wafers, given that the total thickness of the oxidized coating is greater than the thickness of the initial protective coating. The SEM image of the wafer 5 μm to 6 μm thick shows a crack resulting from accidental rupture of the initial coating. Table 3 and FIG. 7C show that the greater the thickness of the initial CrC coating, the thicker the non-oxidized coating (zone C) and the thicker also is the diffusion zone of the non-oxidized coating in the substrate (zone D). On the other hand, the thickness of the oxidized coating (zone A) does not change: about 2 μm of Cr2O3 and 2 μm of partially oxidized transition zone (chromium oxycarbide) (zone B). Since the oxidation phenomenon is superficial, and since the oxidation conditions are the same, it is coherent that the thickness of Cr2O3 and of the transition zone should be identical insofar as the CrC coating has not all been consumed. The oxidation is partial depending on the thickness: less than half of the total thickness of the coating is oxidized for the thinnest coating (2 μm to 3 μm) and a quarter for the thickest (9 μm). The gain in mass is greater for the wafer with a protective layer 9 μm thick, which is also illustrated by FIG. 7C which shows greater progress of the oxidation within the coating. The ratio of the gain in mass between the wafers with a 9 μm coating and a 2-3 μm coating is comparable to the ratio of the oxidized coating thicknesses: 1.44 versus 1.59. Complementary analyses show that the elemental compositions of the coatings are comparable for each wafer in oxidized and non-oxidized zones, with the curves reaching the same mass percentages and following the same trends. In particular, a diffusion of chromium and of carbon is observed in the Zircaloy-4 substrate, among others in the form of ZrC, which advantageously reinforces the adhesion of the coating to the substrate. Moreover, the non-oxidized coating zone C (amorphous CrC) maintains a very dense microstructure, formed from small grains of crystalline chromium carbides of submicron size. As regards the resistance to hydriding, GDS analysis of the 9 μm thick wafer illustrated by FIG. 7D shows a mass percentage of hydrogen of 10 ppm which is not significant in the Zircaloy-4 substrate after oxidation at 1100° C. followed by quenching in water, which confirms the hydrogen-permeation barrier property of the amorphous chromium carbide coating. Even under oxidative conditions at 1100° C. followed by quenching, a nuclear component obtained via the manufacturing process of the invention can maintain a certain degree of mechanical integrity and have a comfortable residual margin of resistance to oxidation/hydriding. 3.3.6. Conclusion Unexpectedly, the inventors have discovered that a nuclear component obtained by the manufacturing process of the invention has improved resistance to oxidation and/or hydriding, in particular at high and very high temperature, among others in the presence of water vapor. Such properties cannot be anticipated with regard to the chemical and metallurgical specificities of zirconium and of the zirconium alloys used for nuclear applications, among others with regard to their chemical composition, surface state, crystalline texture, final metallurgical state (work-hardened or more or less recrystallized), which are properties that are liable to have an influence on the quality and behavior of the coatings. In particular, at low temperature, the α phase of a zirconium alloy (denoted “Zr-α”, of compact hexagonal crystalline structure) transforms into the β phase (denoted “Zr-β”, of centered cubic crystallographic structure) in a temperature range typically extending from 700° C. to 1000° C., On passing from the Zr-α structure to the Zr-β cubic structure, the alloy undergoes local dimensional variations. These variations are in principle unfavorable to the mechanical strength of an outer layer which would cover a zirconium-based inner layer, on account among others of the incompatibility of their expansion coefficients. These adhesion difficulties are accentuated by the mechanisms of diffusion of chemical species that are faster in the Zr-β phase than in the Zr-α phase, and which can modify the interface between the substrate and its coating. Now, the various protective layers deposited via the manufacturing process of the invention have shown very good adhesion to a zirconium-based substrate, even under extreme conditions. 4. Microstructure of a Partially Metastable Si/Cr Monolayer or Multilayer Coating A passivated silicon substrate is provided with a partially metastable chromium monolayer coating according to the conditions of Example 2.2, or with a multilayer coating of similar thickness and chemical composition while additionally observing a waiting time of 5 minutes between the deposition of each of the nine constituent layers of the multilayer coating. The microstructure of these coatings is illustrated by the SEM images represented by: FIG. 8A: monolayer coating obtained with a deposition rate of 5 μm/hour; FIG. 8B: monolayer coating obtained with a deposition rate of 3 μm/hour; FIG. 8C: multilayer coating obtained with a deposition rate of 1 μm/hour. These images show that the monolayer coatings have here a columnar microstructure, whereas the multilayer coating has an equiaxed microstructure. FIG. 8C in particular shows the interface which exists between each of the nine layers which are visualized individually within the multilayer coating. Moreover, the density of the coating increases gradually while reducing the rate of deposition of the partially metastable chromium by reduction of the injection frequency and time. The density of the multilayer coating is estimated as 7.7±0.6 g·cm−3 via RBS (“Rutherford Backscattering Spectrometry”) analyses: taking the uncertainties into account, it is thus of the same order of magnitude as the optimum density of 7.2 g·cm−3 for the stable solid chromium. The deposition conditions thus have an influence on the microstructure of the coating and on its density. In general, the quality of a coating is inversely proportional to its deposition rate, as shown evolution from a porous columnar growth to a dense equiaxed growth obtained reducing the deposition rate. Reducing the deposition rate may be obtained by reducing the deposition temperature, or by increasing the deposition pressure or the precursor concentration of the mother solution injected into the CVD deposition reactor. 5. Hardness of the Coatings In order to measure the hardness of the coatings obtained previously or obtained under similar conditions, nanoindentation experiments are performed. For the amorphous CrC coating, the hardness measurements are performed on a coating obtained from fresh precursor (3.5 μm thick) or from precursor recycled under conditions similar to those presented in Example 6 (1 μm thick). The nanoindentation machine is equipped with a Berkovich indenter (triangular-based pyramid with an angle of 65.27° between the vertical and the height of one of the faces of the pyramid). The measures are taken in compliance with the rule of the tenth: the indenter drives in by less than one tenth of the thickness of the coating. A measurement cycle takes place in three stages: increasing load up to the maximum load, in 30 seconds; maintenance of the maximum load for 30 seconds; removal of the load for 30 seconds. The calculations made by the measurement and analysis software take into account a Poisson coefficient of the coating of 0.2. The results are collated in Table 4. The parameters “H” and “E” are the hardness and the Young's modulus. The parameters “H/E” and “H3/E2” evaluate the durability of the coating: H/E: compares the elastic breaking strength or the abrasion resistance between different coatings. H3/E2: characterizes the elastic behavior of a material and is proportional to the resistance to penetration under load and to the plastic deformation. TABLE 4ThicknessHEH/EH3/E2Coating(μm)Substrate(GPa)(GPa)(—)(GPa)CrxCy3.5Zircaloy-422.7 ± 2.3266.5 ± 25.28.5 × 10−21.6 × 10−1CrxCy1.0Zircaloy-428.9 ± 6.4315.0 ± 41.29.2 × 10−23.6 × 10−1recycledCr(S)3.5Zircaloy-4 9.7 ± 0.8344.8 ± 32.02.8 × 10−27.7 × 10−3columnarCr(S)5.5Zircaloy-416.9 ± 0.6280.1 ± 72.76.0 × 10−26.2 × 10−2denseCrxSizCy5.5Zircaloy-419.9 ± 2.8180.8 ± 17.61.1 × 10−14.58 × 10−1 Cr—— 6.1 ± 0.4224.2 ± 21.42.7 × 10−24.5 × 10−3solidZircaloy-4—— 4.4 ± 0.4108.8 ± 15.74.0 × 10−27.2 × 10−3solid For the various coatings, these results show that: partially metastable Cr: this coating has a high hardness, of about 17 GPa. This hardness is twice as high as that of the commercially available electrolytic hard chromium. It is in the high hardness range of solid metallic chromium, generally comprised between 6 GPa and 25 GPa depending on the production process. amorphous chromium carbide CrC: this coating has an appreciably high hardness, comprised between 22 GPa and 29 GPa. The best hardnesses are obtained when the precursors not consumed on conclusion of the manufacturing process of the invention are recycled. In comparison, the hardness range of a chromium carbide of the prior art generally ranges between 5 GPa and 24 GPa depending on the process used and the deposition conditions. The high hardness range of the amorphous chromium carbide coating obtained via the manufacturing process of the invention is unexpected: it is generally accepted that a chromium carbide is much less hard in amorphous form than in crystalline form, or even that the presence of carbon would soften a chromium-based coating. amorphous CrSiC: in comparison with the amorphous CrC coating, this coating has a hardness of about 20 GPa which is smaller, but, advantageously its Young's modulus of 180 GPa is much lower. The amorphous CrSiC coating thus has a very high durability (H/E=11.1×10−2 and above all H3/E2=4.6×10−1 GPa), which competes with high-durability coatings specially designed for this purpose. 6. Recycling of the Precursors The example of a deposition process with recycling which follows is taken from patent application FR 1562862 (reference [11]). It relates to the deposition of an amorphous chromium carbide coating on a silicon substrate and illustrates by analogy the possibility of recycling one or more of the precursors or of derivatives thereof which remain on conclusion of the manufacturing process of the invention. The recycling is performed with a cryogenic trap. The deposition of an amorphous chromium carbide CrC coating is performed under the following conditions: Injection conditions: injector open time: 0.5 ms frequency: 10 Hz Reagent: BEBC (5 g) Solvent: Toluene (50 mL) Carrier gas: N2 (flow rate of 500 sccm, i.e. 500 cm3/min under standard conditions) Deposition time: 20 minutes Deposition temperature: 450° C.; Deposition pressure: 6666 Pa Evaporation temperature: 200° C. Temperature of the cryogenic trap: 120° C. Amount of daughter solution recovered: 30 mL Two experiments N1 and N2 were performed with a BEBC mother solution. In a third experiment, the two daughter solutions obtained on conclusion of N1 and N2 are recovered to form a recycled mother solution which is used as source of precursor for a third deposition operation N3. For N1 and N2, the thickness of the coating is typically 5 μm. A coating about 1.5 μm thick is obtained on conclusion of N3. The concentration of BEBC is determined and the yield for N1 and N2 is calculated (see Table 5). TABLE 5[BEBC] in[BEBC] inYieldtheDaughtertherelativeInjectedinjectedsolutiondaughterto theExperimentsolutionsolutionrecoveredsolutionprecursorNo.(ml)(g/ml)(ml)(g/ml)(%)N1550.078300.03160%N2500.083300.03459%N330 + 300.04035NotN/Ameasured,sinceverysmall Complementary analyses show that the compositions and amorphous microstructures of the coatings do not depend on the precursor concentration of the injected solution. On the other hand, a daughter solution containing a recycled precursor makes it possible to deposit a coating with a hardness higher than that obtained with the initial precursor of the mother solution. 7. Geometry of the Nuclear Component According to the Invention The nuclear component obtained via the manufacturing process of the invention is described on a cross section with reference to FIGS. 9A, 9B, 9C, and 9D in the nonlimiting particular case of a tubular geometry. When the nuclear component is solid or does not delimit any accessible internal volume, it generally does not comprise any coating deposited on the inner surface of the substrate. According to a first embodiment of the invention, the cladding illustrated by FIG. 9A comprises a substrate 1 whose inner surface delimits a volume that is capable of receiving the nuclear fuel. The substrate 1 forms a support (not coated with an interposed layer for this embodiment) on which is placed a protective layer 2 composed of a protective material which improves the oxidation resistance of the cladding. According to a second embodiment illustrated by FIG. 9B, the cladding is further provided with an interposed layer 3 placed between the substrate 1 and the protective layer 2. In this case, the combination of the substrate 1 and of the interposed layer 3 forms the support. The interposed layer 3 is composed of at least one interposed material which is capable of preventing or limiting the diffusion of the protective material of the protective layer 2 toward the substrate 1. According to a third embodiment illustrated by FIG. 9C, the cladding according to the second embodiment is further provided with an inner protective layer 2B in contact with the internal volume of the cladding and arranged here on the inner face of the substrate 1 (thus not coated with a liner for this embodiment). The inner protective layer 2B completes the protection imparted by the outer protective layer 2A. According to a fourth embodiment illustrated by FIG. 9D, the cladding according to the third embodiment is further provided width a liner 4 positioned between the inner protective layer 2B and the substrate 1. The liner 4 constitutes a diffusion barrier. Needless to say, depending on the desired properties, other embodiments are possible as a function of the presence or absence in each embodiment of the interposed layer 3 of the outer protective layer 2A, of the inner protective layer 2B and/or of the liner 4. The present invention is in no way limited to the embodiments described and represented, and a person skilled in the art will know how to combine them and to make thereof, by means of his general knowledge, numerous variants and modifications. [1] F. Maury, A. Douard, S. Delclos, D. Samelor, C. Tendero; Multilayer chromium based coatings grown by atmospheric pressure direct liquid injection CVD; Surface and Coatings Technology, 204 (2009) 983-987. [2] A. Douard, F. Maury; Nanocrystalline chromium-based coatings deposited by DLI-MOCVD under atmospheric pressure from Cr(CO)6; Surface and Coatings Technology, 200 (2006) 6267-6271. [3] WO 2008/009714 [4] WO 2008/009715 [5] S. Anderbouhr, V. Ghetta, E. Blanquet, C. Chabrol, F. Schuster, C. Bernard, R. Madar; LPCVD and PACVD (Ti,Al)N films: morphology and mechanical properties; Surface and Coatings Technology, Volume 115, Issues 2-3, Jul. 18, 1999, pages 103-110. [6] F. Ossola, F. Maury; MOCVD route to chromium carbonitride thin films using Cr(NEt2)4 as single-source precursor: growth and mechanism., Adv. Mater. Chem. Vap. Deposition, 3 (1997) 137-143. [7] Jin Zhang, Qi Xue and Songxia Li, Microstructure and corrosion behavior of TiC/Ti(CN)/TiN multilayer CVD coatings on high strength steels. Applied Surface Science, 2013. 280: Pages 626-631. [8] A. Weber, C.-P. Klages, M. E. Gross, R. N. Charatan and W. L. Brown, Formation Mechanism of TiN by Reaction of Tetrakis(dimethylamido)-Titanium with Plasma-Activated Nitrogen Journal of The Electrochemical Society, 1995. 142(6): pages L79-L82. [9] Y. S. Li, S. Shimada, H. Kiyono and A. Hirose, Synthesis of Ti—Al—Si—N nanocomposite films using liquid injection PECVD from alkoxide precursors. Acta Materialia, 2006. 54(8): pages 2041-2048. [10] S. Abisset, F. Maury, E. Feurer, M. Ducarroir, M. Nadal and M. Andrieux; Gas and plasma nitriding pretreatment of steel substrates before CVD growth of hard refractory coatings; Thin Solid Films, 315 (1998) 179-185. [11] FR 1562862 filed on Dec. 18, 2015. |
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044252986 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Since the vertical baffle plates in pressurized water reactor cores are not welded together, gaps may exist between the plates that may allow jetting of the reactor coolant water through the gaps. In order to determine if such gaps exist after a period of reactor operation and in order to determine if corrective procedures are required, it is first necessary to measure the gaps between the plates. The invention described herein provides apparatus for remotely measuring gaps between plates in a nuclear reactor for determining if such gaps need to be reduced. Referring to FIG. 1, a pressurized water nuclear reactor is referred to generally as 20 and comprises a reactor vessel 22 having an inlet 24 and an outlet 26. A core plate 28 is suspended in the lower portion of reactor vessel 22 and serves to support fuel assemblies (not shown) when reactor 20 is operating. A generally cylindrical core barrel 30 is disposed in reactor vessel 22 for directing the flow of reactor coolant which may be water from inlet 24 down the annulus between reactor vessel 22 and core barrel 30 and into the lower end of reactor vessel 22. From the lower end of reactor vessel 22, the reactor coolant flows upwardly through holes in core plate 28, through the core area of the reactor, and out through outlet 26. In this manner, the reactor coolant passes in heat transfer relationship to the fuel assemblies which are normally disposed on core plate 28. Referring now to FIGS. 1-4, a plurality of vertical baffle plates 32 are disposed within core barrel 30 and on core plate 28 and define the outer perimeter of the reactor core in a conventional manner. While baffle plates 32 are connected together to form a rigid structure, they are not welded together along their lengths. Since baffle plates 32 are not welded or otherwise sealed along their lengths, gaps 34 may exist between adjoining baffle plates 32. However, when gaps 34 exceed approximately 0.0015 inches, it is generally advisable to reduce the width of the gap in order to prevent jetting of water therethrough as shown in FIG. 3. The jetting of water through gaps 34 can cause vibrations or lead to damage of the fuel assemblies in the core. One means of reducing gap 34 is by the use of peening apparatus 36 which can be used to hammer the edge of one of the baffle plates 32 so as to deform that baffle plate 32 near the end thereof as shown in FIG. 4. By slightly deforming the end of baffle plate 32 along its length, the adjoining baffle plates can be made to touch or otherwise reduce gap 34 therebetween. In this fashion, gap 34 can be reduced or eliminated thereby reducing or eliminating the jetting of water through gaps 34. However, before gaps 34 are attempted to be closed, it is first necessary to determine if a gap 34 exists between two selected baffle plates 32 and the size of the gap. Once the gap is determined to be of such a size that it would be advisable to close, then peening apparatus 36 may be employed to reduce the gap. Referring now to FIGS. 1 and 5, the baffle maintenance apparatus is referred to generally as 40 and comprises a vertical support member 42 with a bottom plate 44 attached thereto at its lower end and a top plate 46 attached thereto at its upper end. Vertical support member 42 may be a stainless steel metal member approximately 14 feet in length and capable of extending from core plate 28 to above the top of baffle plates 32 for supporting equipment to perform operations on baffle plates 32. A plurality of pins 48 are attached to the bottom of bottom plate 44 and are capable of being disposed in flow holes of core plate 28 for aligning and stabilizing vertical support member 42. In addition, an extension member 50 may be attached to top plate 46 for positioning baffle maintenance apparatus 40 on core plate 28. A plurality of vertical rods 52 are also attached to top plate 46 and bottom plate 44 and are arranged parallel to vertical support member 42 for supporting and guiding carriage 54 along rods 52. Carriage 54 is slidably mounted on rods 52 and attached to drive means 56 for selectively moving carriage 54 along rods 52. Drive means 56 comprises a hydraulic motor 58 of approximately 2,000 in-lb torque that is mounted on vertical support member 42 and connected to chain drive mechanism 60 by means of gears 62. Chain drive mechanism 60 is also connected to carriage 54 for moving carriage 54 vertically along rods 52 under the influence of motor 58. In this manner, carriage 54 can be moved along the entire length of baffle plates 32 for performing maintenance operations thereon. Referring to FIGS. 5, 6 and 7, a gripper support assembly 64 is attached to top plate 46 and extends vertically substantially parallel to vertical support member 42 for supporting gripper mechanism 66. Gripper mechanism 66 comprises a mechanical latch 68 pivotally connected to hydraulic cylinder 70 which may be a 2-inch stroke Bimba hydraulic cylinder model number H-092-DUZ. When piston 72 of hydraulic cylinder 70 is moved upwardly latch 68 is caused to grip the top end of baffle plate 32 thereby providing an attachment for baffle maintenance apparatus 40 at the top of baffle plate 32. Similarly, when hydraulic cylinder 70 causes piston 72 to move downwardly, latch 68 is disengaged from baffle plate 32 as shown in phantom in FIG. 6. Gripper support assembly 64 also comprises a plurality of guide members 74 attached to gripper support assembly 64 and arranged to slide over the top of baffle plate 32 for guiding gripper mechanism 66 into proper alignment with baffle plate 32 and for providing additional stability for baffle maintenance apparatus 40. In general, two guide members 74 are provided with each gripper mechanism 66 and with each guide member 74 arranged at right angles to each other for contacting each of the two adjacent baffle plates 32. Thus, it can be seen that baffle maintenance apparatus 40 may be lowered onto core plate 28 with pins 48 disposed in core plate 28 and with gripper mechanism 66 capable of gripping the top of baffle plate 32 thereby positioning carriage 54 in proper relationship to baffle plates 32 for performing maintenance thereon. Referring now to FIGS. 8-11, carriage 54 comprises a mounting member 80 that is slidably disposed on rods 52 and a carriage plate 82 pivotally attached to mounting member 80 by means of pivot pin 84. The attachment of carriage plate 82 to mounting member 80 by means of pivot pin 84 allows carriage plate 82 to pivot in a horizontal plane as shown in phantom in FIG. 11. This pivoting of carriage plate 82 provides a means by which carriage plate 82 can be moved along rods 52 without interfering with baffle plates 32 but allowing carriage plate 82 to be pivoted into a position for performing operations on gaps 34 between baffle plates 32. As can be seen in the drawings, carriage plate 82 is provided with at least two attachment points for pivot pin 84 so that carriage plate 82 can be reversed to perform operations on the opposite baffle plate. A second hydraulic cylinder 86 which may be similar to hydraulic cylinder 70 is attached to carriage plate 82 and to mounting member 80 for selectively pivoting carriage plate 82 with respect to mounting member 80. Carriage plate 82 may have a wheel 88 rotatably disposed thereon for contacting a side of a baffle plate 32 while allowing carriage 54 to move vertically with respect to baffle plate 32. In this manner, carriage plate 82 can be pivoted into contact with baffle plate 32 by means of hydraulic cylinder 86 and can be moved vertically along rods 52 while wheel 88 is in contact with baffle plate 32. Still referring to FIGS. 8-11, a camera support 90 is mounted on mounting member 80 and provides a means to mount camera 92 and light source 94. Camera 92 which may be a Westinghouse Electric Corporation ETV 1250 and light source 94 which may be a 100-watt underwater light source are pivotably mounted on camera support 90 so that they can be manually directed toward the particular gap 34 on which operations are to be performed. Referring now to FIGS. 8 and 12, peening apparatus 36 comprises an hydraulic hammer 96 which may be a Model CH18 Stanley chipping hammer and is attached to table 98. A pressure-compensated pump (not shown) rated at approximately 18 GPM and 3000 psi and powered by a 20 HP motor is connected to and drives hydraulic hammer 96. Table 98 is slidably attached to bar 100 with bar 100 being fixedly attached to carriage plate 82 by means of posts 102. Table 98 is also attached to peening drive means 104. Peening drive means 104 comprises an hydraulic cylinder 106 which may be similar to hydraulic cylinder 70 and is mounted on drive rod 108. Drive rod 108 is slidably disposed through stop 110 which is attached to carriage plate 82. A coil spring 112 is disposed around drive rod 108 and between stop 110 and hydraulic cylinder 106 for damping the reciprocal movements of hydraulic cylinder 106. Drive rod 108 threaded on the end of hydraulic cylinder 106 for adjusting the compression of coil spring 112. A plurality of nipples 116 are mounted on hydraulic cylinder 106 and extend through slots 118 in carriage plate 82 for providing attachment of hydraulic lines to hydraulic cylinder 106. When hydraulic hammer 96 is activated, hammer chisel 120 reciprocates within hydraulic hammer 96 at a rate of approximately 2,000 times per minute. This causes hydraulic hammer 96 to vibrate at the same rate which causes table 98 to also vibrate. Since table 98 is attached to hydraulic cylinder 106 and to hydraulic hammer 96, the vibration of hydraulic hammer 96 is partially absorbed by coil spring 112. Thus the vibration caused by hydraulic hammer 96 can be dampened by coil spring 112. Hydraulic cylinder 106 also provides a means by which hydraulic hammer 96 maybe moved toward or away from baffle plates 32. By activating hydraulic cylinder 106, piston 121 can be moved into or out of hydraulic cylinder 106 thereby sliding table 98 and hydraulic hammer 96 horizontally relative to baffle plates 32. Referring to FIG. 13, a plurality of hydraulic control valves 122, 124, 126 and 128 which may be Parker directional control valves model number D3W1-D-Y are mounted on an assembly and attached to vertical support member 42. Each control valve has a piston 130 with a wheel 132 on the end thereof for contacting actuator shaft 134. The movement of piston 130 into and out of the control valve opens or closes the valve thereby activating the equipment to which it is connected. Actuator shaft 134 is connected to drive cable 136 which is connected to a drive mechanism (not shown) that is located remote from the contol valves. As actuator shaft 134 is moved relative to the control valves, each wheel 132 contacts actuator shaft 134, sequentially, thereby moving piston 130 and operating the corresponding control valve. In this manner, the control valves can be opened or closed in sequence from a remote location. Control valve 122 is connected hydraulically to hydraulic cylinder 70 for activating gripper mechanism 66. Control valve 124 is connected hydraulically to hydraulic cylinder 86 for selectively pivoting carriage plate 82 with respect to mounting member 80. Control valve 126 is connected hydraulically to hydraulic hammer 96 for activating or deactivating hydraulic hammer 96. And, control valve 128 is connected hydraulically to hydraulic cylinder 106 for moving table 98 and hydraulic hammer 96 toward or away from baffle plates 32. In the non-operating position, actuator shaft 134 does not contact any wheels 132 of the control valves. However, as actuator shaft 134 is moved upwardly, actuator shaft 134 sequentially activates control valves 122, 124, 126 and 128 in succeeding order. Likewise, as actuator shaft 134 is moved downwardly, the control valves are deactivated in reverse order. Referring now to FIGS. 8 and 14-23, a gauging mechanism 140 is attached to carriage plate 82. Gauging mechanism 140 is arranged on carriage 54 to be near baffle plates 32 for determining the width of gaps 34 between plates 32. Gauging mechanism 140 comprises a base 144 fixedly attached to plate 142 by means of screws 146 and a movable platform 148 slidably mounted on base 144. Platform 148 is formed in a "T" type configuration and has a plurality of holes 150 therein through which are disposed pins 152. Pins 152 are disposed through holes 150 and attached to base 144 in a manner to allow platform 148 to slide on pins 152 and laterally relative to base 144. Base 144 has a substantially cylindrical slot 154 therein that has a rectangular opening 156 along its top length. A cylindrical rod 158 is slidably disposed in slot 154 and has a post 160 attached thereto in a manner to extend through opening 156. Rod 158 is attached to a push-pull type cable (not shown) that leads to a location outside reactor vessel 22 so that rod 158 may be slid in slot 154 by pushing or pulling the cable. Platform 148 also has a diagonal groove 162 extending from one corner of platform 148 to the corner diagonally opposite therefrom. Post 160 extends into and is slidably disposed in groove 162 for sliding platform 148 on pins 152 and relative to base 144. When rod 158 is moved to a position as shown in FIG. 17, post 160 is located approximately midway along groove 162 and midway along slot 154. In such a position, platform 148 is centered on base 144 as shown in FIG. 18. When rod 158 is moved by the cable to a position as shown in FIG. 19, then post 160 is moved to an extreme end of slot 154 and groove 162 which causes platform 148 to slide on pins 152 into a configuration as shown in FIG. 20. Similarly, when rod 158 is pushed into a position as shown in FIG. 21, post 160 is moved to the other extreme end of slot 154 and groove 162 which causes platform 148 to be slid laterally on pins 152 and into a configuration as shown in FIG. 22. It can be seen that the movement of post 160 in groove 162 causes platform 148 to slide laterally with respect to base 144. In this manner the movement of rod 158 can result in the lateral movement of platform 148 relative to base 144. Referring now to FIGS. 14, 15, 16 and 23-30, gauging mechanism 140 also comprises a gauge 170 rotatably mounted on platform 148. Gauge 170 comprises a substantially cylindrical rod 172 slidably disposed in a substantially cylindrical slot 174 of platform 148. Slot 174 extends the entire length of platform 148 and has a rectangular opening 176 along its top length. Rod 172 has a post 178 attached thereto and extending through opening 176. Rod 176 is attached to a push-pull type cable that extends to a location remote from reactor vessel 22 for moving rod 172 through slot 174. Gauge 170 also comprises a short cylindrical member 180 and a circular member 182 rotatably mounted on platform 148. Post 178 extends through a bore in cylinder member 180 and through circular member 182 and has a nut 184 attached to the top end thereof. Nut 184 prevents cylindrical member 180 and circular member 182 from being separated from post 178 but allows cylindrical member 180 and circular member 182 to rotate about post 178. Circular member 180 has a plurality of feeler gauges 186 attached thereto. Feeler gauges 186 may be thin metal strips ranging from 0.0015 inches to 1.050 inches in thickness for being disposed in gaps 34 to determine the width of the gaps between baffle plates 32. For example, circular member 180 may have four feeler gauges 186 disposed thereon at 90.degree. to each other. In addition, a like number of pins 188, 190, 192 and 194 are attached to the underside of circular member 180 and arranged at approximately 90.degree. to each other and equidistant between each feeler gauge as shown in FIG. 14. A spring loaded stop 200 is mounted on platform 148 and comprises a dog 202 rotatably disposed on pin 204 with a spring 206 disposed therearound. Dog 202 is arranged to rotate in one direction and is spring-loaded to return to its original position but is prevented from rotating in the opposite direction by post 208. A set screw 210 is mounted on platform 148 and arranged to contact cylindrical member 180 for limiting the travel of cylindrical member 180. Referring now to FIGS. 27-30, in its initial position, gauge 170 is arranged to have a feeler gauge 186 substantially in alignment with opening 176 so as to be able to be inserted into a gap 34 between two baffle plates 32 as shown in FIG. 27. When rod 172 is pulled through slot 174 by cables, post 178 is moved trough opening 174 and toward set screw 210 until a pin such as pin 190 contacts dog 202 as shown in FIG. 28. Since post 208 prevents dog 202 from rotating in a clockwise motion as viewed in FIG. 28, pin 190 is prevented from moving toward set screw 210. However, because circular member 182 is rotatably mounted on post 178, as post 178 is moved closer to set screw 210, circular member 182 rotates in a clockwise direction as shown in FIG. 29. It can be seen in FIG. 29 that pin 190 slides along stop 202 as circular member 182 rotates. As post 178 is moved even closer to set screw 210, circular member 182 continues to rotate and completes a 90.degree. rotation as circular member 182 contacts set screw 210 as shown in FIG. 30. At this point, rod 172 and post 178 may be pushed away from set screw 210 and into the original configuration of FIG. 27 except that circular member 182 will have been rotated 90.degree.. Since dog 202 can rotate about pin 204 in a counterclockwise direction, dog 202 can rotate and allow a pin such as pin 188 to pass by as shown in phantom in FIG. 30. In this manner, by moving rod 172 through slot 174, a plurality of feeler gauges 186 can be rotated and extended into a gap 34 between two adjacent baffle plates 32 in order to measure the width of the space between the two baffle plates. It should be noted that platform 148 can be arranged to mount gauge 170 on the side thereof rather than on the top so as to rotate feeler gauges 186 in a vertical plane rather than in a horizontal plane as illustrated in FIG. 31. OPERATION When it is desired to measure gap 34 between two adjacent baffle plates 32, nuclear reactor 20 is shut down and all of the fuel assemblies are removed in a conventional manner. With reactor vessel 22 still filled with reactor coolant, baffle maintenance apparatus 40 is lowered into reactor vessel 22 by means of extension member 50. As baffle maintenance apparatus 40 is lowered into reactor vessel 22, pins 48 are inserted into flow holes in core plate 28 and guide members 74 are slid over the top ends of baffle plates 32. In this position, vertical support member 42 is aligned substantially vertically relative to baffle plates 32. Next, actuator shaft 134's drive mechanism which is located remote from reactor vessel 22 is activated which causes actuator shaft 134 to move upwardly relative to control valves 122-128. As actuator shaft 134 contacts wheel 132 of control valve 122, piston 130 is depressed which opens control valve 122 and activates gripper mechanism 66. Activation of gripper mechanism 66 causes latch 68 to engage the top of baffle plate 32 as shown in FIG. 6. At this point baffle maintenance apparatus is firmly positioned in reactor vessel 22. As actuator shaft 134 continues to move upwardly, control valve 124 is similarly activated which causes hydraulic cylinder 86 to pivot carriage plate 82 into a position as shown in FIGS. 9, 10 or 11 depending on the particular gap 34 that is to be inspected. The pivoting of carriage plate 82 moves carriage plate 82, for example, from a position as shown in phantom in FIG. 11 to the position as shown in full in FIG. 11. It should be understood from a review of the drawings and particularly FIG. 10, that wheel 88 can be mounted on either side of carriage plate 82 and that carriage plate 82 can be mounted in a reverse manner on mounting member 80 for accessing the various corners of baffle plates 32. With carriage plate 82 in this configuration, gap 34 may be either inspected with gauging mechanism 140 or gap 34 may be closed by activating peening apparatus 36. Also, with carriage plate 82 in this configuration, camera 92 and light source 94 are directed toward the particular gap 34 to be inspected so that working personnel located remote from reactor vessel 22 may view the work area. Next, working personnel or automatic equipment may manipulate the cables connected to rod 172 for rotating a selected feeler gauge 186 (as previously described) into position for being inserted into the selected gap 34. Also, by moving rod 172, the selected feeler gauge 186 may be moved into a position as shown in FIG. 27 which will cause the feeler gauge to be inserted into the selected gap 34. In addition, cables may be manipulated to move rod 158 so as to laterally position the feeler gauge so that the feeler gauge may be inserted in gap 34. All of this may be performed in view of camera 92 so that the operating personnel may verify the movements. Similarly, additional feeler gauges 186 may be inserted into gap 34 so that operating personnel may determine the width of gap 34. If it is determined that gap 34 has a width greater than desired, actuator shaft 134 may again be moved so as to activate hydraulic hammer 96. With hydraulic hammer 96 vibrating, actuator shaft 134 may then activate control valve 128 for moving table 98 and hydraulic hammer 96 toward the baffle plate 32 for closing the gap 34 between the baffle plates. When hydraulic hammer 96 contacts baffle plate 32, a portion of the edge of the baffle plate is deformed which causes gap 34 to be reduced as shown in FIG. 4. In this fashion, each gap 34 may be inspected and if necessary, the gap may be reduced to eliminate or greatly reduce the jetting of water through the gap when the reactor is operating. In addition, drive means 56 may be employed in either the gauging or peening operation for moving carriage 54 including the gauging and peening apparatus along the length of each gap 34 so that the entire length of the gap may be inspected or reduced. Of course, baffle maintenance apparatus 40 may be deactivated and removed from reactor vessel 22 in the reverse manner to which it was inserted. Therefore, the invention provides apparatus for determining the size of a gap between two adjacent baffle plates in a nuclear reactor so that those gaps that are too large may be reduced. |
049960202 | abstract | A method of restraining a diffusion of tritium and an apparatus for same are disclosed. The method includes the step of disposing a hydrogen-absorbing metal in a tritium passage in a direction of diffusion of tritium so that the hydrogen-absorbing metal absorbs tritium to be diffused. The apparatus includes a hydrogen-absorbing metal surrounding a device of a fast breeder reactor. The tritium to be diffused can be readily captured. An arrangement of capturing the tritium permeating the device is simplified in the present invention. |
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abstract | The present invention provides an electron beam measurement technique for measuring the shapes or sizes of portions of patterns on a sample, or detecting a defect or the like. An electron beam measurement apparatus has a unit for irradiating the patterns delineated on a substrate by a multi-exposure method, and classifying the patterns in an acquired image into multiple groups according to an exposure history record. The exposure history record is obtained based on brightness of the patterns and a difference between white bands of the patterns. |
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050892126 | abstract | A nuclear reactor, such as advanced pressurized water reactor, with controlled power output includes a reactor core having a cross section in a given plane with a center of area and axes of symmetry passing through the center of area. A multiplicity of control elements are disposed in groups, each having at least one of the control elements. The control elements of each of the groups with more than one control element are joined together. The groups are symmetrical to at least two of the axes of symmetry. Drive mechanisms each move a respective one of the groups of control elements. Each of the control elements has a plurality of control rods. A support structure joins the control rods to one another. Fuel assemblies are disposed in groups. Each of the groups of control elements has a given number of control elements being associated with one of the groups of fuel assemblies having the given number of fuel assemblies. Other fuel assemblies with which the control elements are not associated surround the groups of fuel assemblies. |
summary | ||
047599120 | claims | 1. In a nuclear fuel assembly including an outer hollow tubular flow channel for directing the flow of coolant/moderator fluid through the fuel assembly and a hollow water cross extending through said channel and attached along the interior of said channel so as to divide it into plurality of separate compartments and provide a central hollow channel for flow of coolant/moderator fluid through the fuel assembly, a hybrid fuel design comprising: (a) cladded rod-type nuclear fuel located within said compartments spaced from said outer channel and said water cross; and (b) cladded plate-type thermally fissile nuclear fuel attached on the exterior of said water cross for distributing power generation between said cladded rod-type nuclear fuel and said water cross. (a) a plurality of elongated fuel rods located within said compartments between the interior of said outer flow channel and exterior of said radially extending members of said water cross; and (b) a plurality of elongated thermally fissile fuel plates attached on the exterior of said water cross members. an inner sheet of nuclear fuel disposed adjacent the exterior of each of said water cross members; and an outer sheet of cladding disposed adjacent the exterior of said inner sheet and having a periphery attached to said water cross members so as to sealably enclose said inner sheet of nuclear fuel. (a) an array of spaced fuel rods; (b) an outer hollow tubular flow channel surrounding said fuel rods so as to direct flow of coolant/moderator fluid therealong; (c) a water cross extending through said channel and having a plurality of radially extending panels dividing said array of fuel rods into a plurality of separate fuel rod mini-bundles; and (d) a plurality of thermally fissile fuel plates respectively attached on the exterior of said water cross panels. an inner sheet of nuclear fuel disposed adjacent the exterior of each of said water cross panels; and an outer sheet of cladding disposed adjacent the exterior of said inner sheet and having a periphery attached to said water cross panels so as to sealably enclose said inner sheet of nuclear fuel. 2. The fuel assembly as recited in claim 1, wherein said plate-type nuclear fuel is generally coextensive in length with said rod-type nuclear fuel. 3. The fuel assembly as recited in claim 1, wherein said plate-type nuclear fuel is generally coextensive in width with said water cross but shorter in length than said water cross. 4. In a nuclear fuel assembly including an outer hollow tubular flow channel providing an enclosure for directing the flow of coolant/moderator fluid through the fuel assembly and a water cross extending through said channel and having a plurality of radially extending members attached along the interior of said channel and dividing it into a plurality of separate compartments, a hybrid fuel design comprising: 5. The fuel assembly as recited in claim 4, wherein said fuel plates are generally coextensive in length with said fuel rods. 6. The fuel assembly as recited in claim 4, wherein said fuel plates are generally coextensive in width with said water cross members but shorter in length than said water cross members. 7. The fuel assembly as recited in claim 4, wherein each fuel plate includes: 8. In a fuel assembly, the combination comprising: 9. The fuel assembly as recited in claim 8, wherein each fuel plate includes: 10. The fuel assembly as recited in claim 8, wherein said fuel plates are generally coextensive in length with said fuel rods. 11. The fuel assembly as recited in claim 8, wherein said fuel plates are generally coextensive in width with said water cross panels but shorter in length than said water cross panels. 12. The fuel assembly as recited in claim 1 wherein said plate-type thermally fissile nuclear fuel is curved for following the contour of said exterior of said water cross. 13. The fuel assembly as recited in claim 4, wherein said thermally fissile fuel plates are curved for following the contour of said exterior of said water cross members. 14. The fuel assembly as recited in claim 8, wherein said thermally fissile fuel plates are curved for following the contour of said exterior of said water cross panels. |
claims | 1. A fuel assembly, comprising a plurality of fuel rods bundled in grid pattern, a part of the fuel rods containing gadolinium as a burnable poison, wherein at least one of the fuel rods that contain gadolinium comprises gadolinium-155 and gadolinium-157, wherein in the fuel rods that contain gadolinium-155 and gadolinium-157, gadolinium-157 is enriched and is present in an amount more than its isotopic abundance in natural gadolinium, and a ratio of said gadolinium-155 to said gadolinium-157 is 0.1 or less. 2. A fuel assembly, comprising a plurality of fuel rods bundled in grid pattern and containing enriched uranium as a nuclear fuel material, a part of the fuel rods containing gadolinium as a burnable poison, G 0 less than 0.25 xc2x7Pxc2x7M/W wherein at least one of the fuel rods that contain gadolinium comprises gadolinium-155 and gadolinium-157, wherein in the fuel rods that contain gadolinium-155 and gadolinium-157, gadolinium-157 is enriched and is present in an amount more than its isotopic abundance in natural gadolinium, and a ratio of said gadolinium-155 to said gadolinium-157 is 0.1 or less, and wherein an average concentration (wt %) G 0 of an oxide of the gadolinium-155 and an oxide of gadolinium-157 averaged over an entire fuel rod having the gadolinium is set in the range shown by the following expression: wherein in the expression, M denotes a number of month under rated power operation per one cycle of an equilibrium reactor core, P power density of a nuclear reactor (kw/l unit), W a sum of isotopic composition (% unit) of gadolinium-155 and gadolinium-157. 3. A nuclear reactor loaded with the fuel assembly set forth in claim 1 and a fuel assembly having natural gadolinium, respectively, wherein a gadolinia concentration averaged over the fuel assembly having natural gadolinium is larger than that of the fuel assembly set forth in claim 1 . claim 1 claim 1 4. A nuclear reactor loaded with the fuel assembly set forth in claim 2 and a fuel assembly having natural gadolinium, respectively, wherein a gadolinia concentration averaged over the fuel assembly having natural gadolinium is larger than that of the fuel assembly set forth in claim 2 . claim 2 claim 2 |
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summary | ||
claims | 1. A fuel assembly comprising a plurality of fuel rods, each fuel rod containing a plurality of nuclear fuel pellets, wherein at least one fuel pellet in more than 50% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound, wherein the boron-containing compound is boron nitride, and wherein the actinide compound is constituted of one or a multiplicity of actinides compounded with nitrogen, the actinide selected from the group consisting of uranium, plutonium, thorium and a combination thereof. 2. The fuel assembly of claim 1, wherein said boron-containing compound in said at least one fuel pellet comprises enriched boron. 3. The fuel assembly of claim 2, wherein said enriched boron is enriched to a content of 10B greater than natural boron. 4. The fuel assembly of claim 1, wherein said boron-containing compound is present in said fuel pellet in an amount of about 5 ppm to about 5 wt %, based on a total amount of fuel in said fuel pellet. 5. The fuel assembly of claim 4, wherein said boron-containing compound is present in said fuel pellet in an amount of about 10 ppm to about 20,000 ppm, based on the total amount of fuel in said fuel pellet. 6. The fuel assembly of claim 5, wherein said boron-containing compound is present in said fuel pellet in an amount of about 600 ppm to about 20,000 ppm, based on the total amount of fuel in said fuel pellet. 7. The fuel assembly of claim 1, wherein at least one fuel pellet in at least 60% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound. 8. The fuel assembly of claim 1, wherein at least one fuel pellet in at least 80% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound. 9. The fuel assembly of claim 1, wherein at least one fuel pellet in at least 50% and less than 100% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound. 10. A light water reactor having a fuel assembly, the fuel assembly comprising a plurality of fuel rods, each fuel rod containing a plurality of nuclear fuel pellets not requiring coatings, wherein at least one fuel pellet in more than 50% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound, wherein the boron-containing compound is boron nitride, and wherein the actinide compound is constituted of one or a multiplicity of actinides compounded with nitrogen, the actinide selected from the group consisting of uranium, plutonium, thorium, and a combination thereof. 11. The light water reactor of claim 10, wherein said boron-containing compound in said at least one fuel pellet comprises enriched boron. 12. The light water reactor of claim 11, wherein said enriched boron is enriched to a content of 10B greater than natural boron. 13. The light water reactor of claim 10, wherein said boron-containing compound is present in said fuel pellet in an amount of about 5 ppm to about 5 wt percent, based on a total amount of fuel in said fuel pellet. 14. The light water reactor of claim 13, wherein said boron-containing compound is present in said fuel pellet in an amount of about 10 ppm to about 20,000 ppm, based on the total amount of fuel in said fuel pellet. 15. The light water reactor of claim 14, wherein said boron-containing compound is present in said fuel pellet in an amount of about 600 ppm to about 20,000 ppm, based on the total amount of fuel in said fuel pellet. 16. The light water reactor of claim 10, wherein at least one fuel pellet in at least 60% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound. 17. The light water reactor of claim 10, wherein at least one fuel pellet in at least 80% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound. 18. The light water reactor of claim 10, wherein at least one fuel pellet in at least 50% and less than 100% of said fuel rods in said fuel assembly comprises a sintered admixture of an actinide compound and a boron-containing compound. |
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