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048287881 | description | DESCRIPTION OF PREFERRED EMBODIMENT Seen in FIG. 1 is a part of the bottom of the vessel 1 of a fast neutron nuclear reactor cooled by sodium, the vessel being filled with coolant liquid sodium in which the core 2 of the reactor is immersed. The core 2 is constituted by assemblies placed side by side and engaged by their lower part in a bed 3 bearing on a plating 4 constituted by a rigid structure of large size bearing on the bottom of the vessel 1 through a support sleeve 4a which comes in contact with a reinforced part 1a of the vessel. The vessel 1 encloses, in addition to the bed 3 and the plating 4, internal structures, such as the wall 5 which comes to bear on the bed 3 and defines a toric region 6 inside the vessel 1; these internal structures also include cylindrical sleeves having a vertical axis such as the sleeve 7 fixed along an opening in the wall 5 and constituting a hollow shaft for the passage of a pump 8 only the lower part of which is shown in FIG. 1. The pump 8 comprises a vertical body 8a of generally cylindrical shape placed inside the shaft 7 enclosing the active elements of the pump and including intake ports for the liquid sodium in the shaft 7. The upper part of the body 8a of the pump, not shown in FIG. 1, extends through the horizontal slab of the reactor closing the vessel 1 and is fixed to this slab which ensures the suspension of the pump. The region 6 defined inside the vessel 1 by the wall 5 constitutes the cold collector of the reactor in which gathers the cooled liquid sodium after passage through the intermediate heat exchangers extending into the vessel and heating the secondary liquid sodium by using the heat taken off by the primary liquid sodium in the core 2 of the reactor. The pump shaft 7 which communicates with the cold collector is also filled with cooled liquid sodium which the pump 8 re-injects at the base of the core 2. In power fast neutron nuclear reactors at present in service or the construction of which is envisaged, the vessel encloses four indentical pumps associated with eight intermdiate heat exchangers, these elements being placed in an annular region on the periphery of the vessel. Reference will now be made to FIGS. 1, 2 and 3 for describing the lower part of the pump 8 comprising a spherical tank, or discharge sphere 10, into which the lower part of the body 8a of the pump opens. The sphere 10 is open in its upper part in order to constitute a flange on which comes to bear a corresponding support flange 12 of the body of the pump 8. Placed inside the sphere 10 is a separation structure 13 defining two regions in the discharge sphere into each of which opens a discharge pipe 14 through which the liquid sodium cooled by the pump will be discharged. The two regions inside the sphere 10 and the two discharge pipes 14 are perfectly symmetrical relative to a vertical plane containing the axis Z-Z' of the pump 8. The line 15 of this plane of symmetry is seen in FIG. 2. The discharge pipes 14 are engaged in the bed 3 and open out, at their end opposed to their end connected to the sphere 10, into this bed under the core 2. These pipes 14 therefore constitute piping supplying the core with sodium which circulates, from the bed upwardly into the assemblies of the core. The plating 4 comprises, in a radial direction of the vessel 1, an extension 16 which extends outwardly beyond the periphery of the bed 3 supporting the core 2. The suspension of the sphere 10 from the fixed structure of the reactor is ensured by this radial extension 16 of the plating 4. A frusto-conical sleeve 18 is fixed by welding in the region of a connecting member to the sphere 10 along its large base. This frusto-conical sleeve 18 is also integral with a connecting structure 19 fixed along its small base. The frusto-conical sleeve 18 is so disposed that its axis is coincident with the axis Z-Z' of the pump. A substantially planar metal strip 20 is fixed by welding, on one hand, along a horizontal region 21, to the end of the extension 16 of the plating 4 and, on the other hand, along a horizontal region 22, to the connecting structure 19. The regions 21 and 22 extend throughout the width of the strip or plate 20, the region 22 ensuring the connection of the strip 20 to the structure 19 being disposed below the region 21 ensuring the connection of this strip 20 to the fixed structure 4. The sphere 10 is in this way suspended from the plating 4 through the strip 20. As can be seen in FIG. 3, the sleeve 18 is open throughout the height of its lateral wall and constitutes a diametrically extending slot 24 which is upwardly divergent. The strip 20 is disposed along this slot 24 which allows movements of this strip inside the slot 24. The connecting structure 19 comprises two horizontal plates 25a and 25b fixed by welding on each side of the strip 20 and transverse reinforcing lugs 26. The slot 24 extends in a diametrical direction of the sleeve 18 and perpendicularly to the plane of symmetry of the discharge pipes 14, the line 15 of which is seen in FIG. 20. The fixing edge of the strip 20 to the extension 16 is such that this strip 20 is substantially vertical when the sphere 10 is not subjected to stresses of thermal origin. This strip 20 is contained in a vertical plane containing the axis Z-Z' of the pump and perpendicular to the plane of symmetry of the discharge pipes. The strip 20 includes a lower part whose width corresponds to the length of the connecting plates 25a and 25b and an upper part which is wider and located above the connecting region 21. This wider part of the strip is terminated downwardly and upwardly by two edge portions 28 which extend outwardly beyond the lower part of the plate 20. Two plates 30 are vertically fixed to the extension 16 of the plating 4 on each side of this extension vertically below the edge portions 28. As can be seen in FIG. 1a, the plate 20 has a thinner part 20a whose thickness is substantially equal to one half of the thickness of the other parts of the plate. This thinner region 20a extends approximately between the junction regions 21 and 22. Subsequent to an incident or rapid change in the operating conditions, rapid and high-amplitude variations occur in the temperature of the coolant fluid, the sphere 10 is put under stress owing to variations in expansions of the materials, principally in radial directions parallel to the plane of symmetry of the discharge pipes 14. These forces perpendicular to the plane of the strip 20 result in displacements of the sphere 10 which simply bring about a bending of the flexible strip 20 in its thinner part 20a. The assembly constituted by the discharge sphere 10, its frusto-conical connecting sleeve 18 and the discharge pipes 14 can move without producing stresses other than the bending stresses in the strip 20. The bending of the strip 20 is rendered possible owing to the movement allowed by the shape of the slots 24 provided in the suspension sleeve 18. The strip 20 which has a large width and which is subjected to tensile stress may easily ensure the suspension of the discharge sphere whose mass is not extremely large. Further, the vertical distance between the connecting regions 21 and 22 of the strip 20 is such that, even in respect of movements of the sphere of maximum amplitude, the bending of the strip 20 remains small. In the case of a nuclear reactor having an electric power of 1 500 MW having four primary pumps, there will be used, for example, a strip 20 of stainless steel constituted by a plate having a thickness of 40 mm thinned down in a central bending region to a thickness of 20 mm. The width of the plate 20 in its lower part is 1.60 m and the height of the thinned-down region is 0.45 m. In the event that the plate 20, breaks as a result of repeated and non-envisaged fatigues or non-detected internal defects, the discharge sphere 10 would come to bear on the upper part of the strip 20 which bears, through the edge portions 28, on the vertical plates 30. The principal advantages of the support device according to the invention are to permit an extremely reliable suspension,which is nonetheless flexible,of the discharge sphere constituting the lower part of this pump. In the event of rapid variations in the temperature of the coolant fluid of the reactor, the radial expansions are taken up by the bending of the plate 20 with no notable stresses being transmitted to the connection region between the discharge pipes 14 and the bed 3. In a general way, the structures constituting the lower part of the pump or connected to this lower part are subjected to stresses of only very low value in the event of rapid variations in the temperature of the coolant fluid. It is possible to imagine a connecting device between the sphere and the strip of a type different from a split frusto-conical sleeve, a connecting structure different from the structure 19 and a strip of a form other than that described or a plurality of strips. Likewise, the discharge tank may have a shape different from that of a sphere and any number of pipes may be connected thereto. The pump comprising a discharge tank suspended in an elastically yieldable manner from the fixed structure of the reactor according to the invention may be used in nuclear reactors other than fast neutron nuclear reactors cooled with liquid sodium and of the integrated type. Such a pump may be used in non-integrated reactors irrespective of the coolant fluid and, generally, in any device for circulating a thermal exchange fluid in a nuclear reactor employing a pump having a vertical body suspended by its upper part. |
050358405 | abstract | A process for removing metal salts from an H.sub.4 EDTA precipitate by esterification with an esterification reagent to produce an esterification mixture comprising a solid metal salt an EDTA ester and thereafter separating the solid metal salt from the esterification mixture. |
046876286 | abstract | The inner barrel assembly of a pressurized water reactor includes an array of interleaved first and second matrices of respective first and second pluralities of rod guides of corresponding first and second different types, disposed in parallel axial relationship, the bottom ends of the guides being affixed to the top core plate of a lower barrel assembly. A flexible support structure connects the top ends of the rod guides to a lower calandria plate and particularly includes corresponding first and second pluralities of top support plates respectively connected to the rod guides of the first and second different types and having outer peripheries generally corresponding to the respective first and second types of rod guides and inner openings for permitting axial passage therethrough of corresponding, first and second different types of rod clusters. The respective top support plates of the first and second different types have mating, respective exterior and interior vertices for assemblage of same in an array of interdigitated, respective matrices. A flexible linkage is connected between each top plate of the first type and the respective surrounding and contiguous, interdigitated top plates of the second type, thereby interconnecting all of the top plates of the array in a concatenated arrangement. Stop pins received in aligned bores and extending between the interdigitized top plates limit the extent of load that can be applied to the linkage, and the ultimate extent of possible relative movement therebetween. Central recesses in top plates of the second type receive extensions from the lower calandria plate which establish basic alignment of the concatenated and interleaved matrices of top plates. Leaf springs secured to the calandria bottom plate engage and exert a downward force, at their outer ends, on the surfaces of the top plates of the second type for establishing a frictional force opposing lateral movement of the top plates of the second type and, correspondingly, through the concatenated and interleaved arrangement, the top plates of the first type, thereby opposing lateral movement while permitting restrained axial displacement of the individual rod guides. |
abstract | An extreme ultraviolet light source apparatus in which only particles having a high transmittance for EUV light adhere to an EUV collector mirror even if fast ions emitted from plasma collide with a structural member in a vacuum chamber, and thereby, the reflectance thereof is not easily degraded. The apparatus includes: a vacuum chamber; a target supply unit for supplying a target to a predetermined position in the vacuum chamber; a driver laser for applying a laser beam to the target to generate the plasma; a collector mirror for collecting and outputting extreme ultraviolet light emitted from the plasma; a collector mirror holder for supporting the collector mirror; and a shielding member formed of a material having a high transmittance for the extreme ultraviolet light, for shielding the structural member such as the collector mirror holder from the ions generated from the plasma. |
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claims | 1. An assembly comprising:a nuclear reactor;a bundle of target assemblies within the reactor, at least one of the target assemblies comprising a housing defining a volume, wherein Np or Am spheres are within the volume;each of the target assemblies within the bundle being defined by a length along one axis, the length being divided into one or more portions, wherein the Np or Am spheres are contained within at least one of the one or more portions; andwherein two target assemblies that are arranged axially to one another have spheres in the upper portion of one and the lower portion of the other. 2. The reactor assembly of claim 1 wherein the target is circular in at least one cross section. 3. The reactor assembly of claim 1 wherein the bundle is an arrangement of target assemblies, the target assemblies being arranged in substantial circles about a center portion of the bundle. 4. The reactor assembly of claim 3 wherein the circles are substantially concentric defining both inner and outer circles. 5. The reactor assembly of claim 4 wherein each of the assemblies in the inner circle house Np or Am spheres. 6. The reactor assembly of claim 4 wherein each of the assemblies in the inner circle is lined with a ceramic material. 7. The reactor assembly of claim 4 wherein each of the assemblies in the inner circle is capped with a ceramic material. |
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summary | ||
040295451 | summary | BACKGROUND OF THE INVENTION This invention relates broadly to an improvement in nuclear fuel elements for use in the core of nuclear fission reactors, and more particularly to an improved nuclear fuel element having a composite cladding container having an outer layer and two coatings on the inside surface of the outer layer. Nuclear reactors are presently being designed, constructed and operated in which the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, non-reactive, heat conductive container or cladding. The elements are assembled together in a lattice at fixed distances from each other in a coolant flow channel or region forming a fuel assembly, and sufficient fuel assemblies are combined to form the nuclear fission chain reacting assembly or reactor core capable of a self-sustained fission reaction. The core in turn is enclosed within a reactor vessel through which a coolant is passed. The cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a moderator is present, or both if both the coolant and the moderator ar present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator or both if both the coolant and the moderator are present. Common cladding materials are stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others. The failure of the cladding, i.e., a loss of the leak tightness, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation. Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the clad material due to mechanical or chemical reactions of these cladding materials under certain circumstances. Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel claddings since they have low neutron absorption cross sections and at temperatures below about 750.degree. F. (about 398.degree. C.) are strong, ductile, extremely stable and non-reactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators. However, fuel element performance has revealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this undesirable performance is promoted by localized mechanical stresses due to fuel cladding differential expansion (stresses in the cladding are localized at cracks in the nuclear fuel). Corrosive fission products are released from the nuclear fuel and are present at the intersection of the fuel cracks with the cladding surface. Fission products are created in the nuclear fuel during the fission chain reaction during operation of a nuclear reactor. The localized stress is exaggerated by high friction between the fuel and the cladding. Within the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding and the residual water inside the cladding may build up to levels which under certain conditions can result in localized hydriding of the cladding with concurrent local deterioration in the mechanical properties of the cladding. The cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide and carbon dioxide over a wide range of temperatures. The zirconium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor and this occurs in spite of the fact that these gases and fission product elements may not be present in the reactor coolant or moderator, and further may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element. Sintered refractory and ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuel, release measurable quantities of the aforementioned gases and fission products upon heating, such as during fuel element manufacture and further release fission products during irradiation. Particulate refractory and ceramic compositions, such as uranium dioxide powder and other powders used as nuclear fuel, have been known to release even larger quantities of the aforementioned gases during irradiation. These released gases are capable of reacting with the zirconium cladding containing the nuclear fuel. This in light of the foregoing, it has been found desirable to minimize attack of the cladding from water, water vapor and other gases, especially hydrogen, reactive with the cladding from inside the fuel element throughout the time the fuel element is used in the operation of nuclear power plants. One such approach has been to find materials which will chemically react rapidly with the water, water vapor and other gases to eliminate these from the interior of the cladding, and such materials are called getters. Another approach has been to coat the nuclear fuel material with a ceramic to prevent moisture coming in contact with the nuclear fuel material as disclosed in U.S. Pat. No. 3,108,936. U.S. Pat. No. 3,085,059 presents a fuel element including a metal casing containing one or more pellets of fissionable ceramic material and a layer of vitreous material bonded to the ceramic pellets so that the layer is between the casing and the nuclear fuel to assure uniformly good heat conduction from the pellets to the casing. U.S. Pat. No. 2,873,238 presents jacketed fissionable slugs of uranium canned in a metal case in which the protective jackets or coverings for the slugs are a zinc-aluminum bonding layer. U.S. Pat. No. 2,849,387 discloses a jacketed fissionable body comprising a plurality of open-ended jacketed body sections of nuclear fuel which have been dipped into a molten bath of a bonding material giving an effective thermally conductive bond between the uranium body sections and the container (or cladding). The coating is disclosed as any metal alloy having good thermal conduction properties with examples including aluminum-silicon and zinc-aluminum alloys. Japanese Patent Publication No. SHO 47-46559dated Nov. 24, 1972, discloses consolidating discrete nuclear fuel particles into a carbon-containing matrix fuel composite by coating the fuel particles with a high density, smooth carbon-containing coating around the pellets. Still another coating disclosure is Japanese Patent Publication No. SHO 47-14200 in which the coating of one of two groups of pellets is with a layer of silicon carbide and the other group is coated with a layer of pyrocarbon or metal carbide. The coating of a nuclear fuel material introduces reliability problems in that achieving uniform coatings free of faults is difficult. Further, the deterioration of the coating can introduce problems with the long-lived performance of the nuclear fuel material. U.S. Pat. application Ser. No. 330,152 filed Feb. 6, 1973 discloses a method for preventing corrosion of nuclear fuel cladding consisting of the addition of a metal such as niobium to the fuel. The additive can be in the form of a powder, provided the subsequent fuel processing operation does not oxidize the metal, or incorporated into the fuel element as wires, sheets or other forms in, around, or between fuel pellets. Document GEAP-4555 dated Feb. 1964 discloses a composite cladding of a zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy, and the composite cladding is fabricated by use of extrusion of a hollow billet of the zirconium alloy having an inner lining of stainless steel. This cladding has the disadvantage that the stainless steel develops brittle phases, and the stainless steel layer involves a neutron absorption penalty of about ten to fifteen times the penalty for a zirconium alloy layer of the same thickness. U.S. Pat. No. 3,502,549 discloses a method of protecting zirconium and its alloys by the electrolytic deposition of chrome to provide a composite material useful for nuclear reactors. A method for electrolytic deposition of copper on Zircaloy-2 surfaces and subsequent heat treatment for the purpose of obtaining surface diffusion of the electrolytically deposited metal is presented in Energia Nucleare Volume 11, number 9 (Sept. 1964 ) at pages 505-508. In Stability and Compatibility of Hydrogen Barriers Applied to Zirconium Alloys by F. Brossa et al (European Atomic Energy Community, Joint Nuclear Research Center, EUR 4098e 1969 ), methods of deposition of different coatings and their efficiency as hydrogen diffusion barriers are described along with an Al-Si coating as the most promising barrier against hydrogen diffusion. Methods for electroplating nickel on zirconium and zirconium tin alloys and heat treating these alloys to produce alloy-diffusion bonds are disclosed in Electroplating on Zirconium and Zirconium-Tin by W. C. Schickner et al (BMl-757, Technical Information Service, 1952). U.S. Pat. No. 3,625,821 presents a fuel element for a nuclear reactor having a fuel cladding tube with the inner surface of the tube being coated with a retaining metal of low neutron capture cross section such as nickel and having finely dispersed particles of a burnable poison disposed therein. Reactor Development Program Progress Report of August, 1973 (ANL-RDP-19) discloses a chemical getter arrangement of a sacrificial layer of chromium on the inner surface of a stainless steel cladding. Another approach has been to introduce a barrier between the nuclear fuel material and the cladding holding the nuclear fuel material as disclosed in U.S. Pat. No. 3,230,150 (copper foil), German Patent Publication DAS 1,238,115 (titanium layer), U.S. Pat. No. 3,212,988 (sheath of zirconium, aluminum or beryllium), U.S. Pat. No. 3,018,238 (barrier of crystalline carbon between the UO.sub.2 and the zirconium cladding), and U.S. Pat. No. 3,088,893 (stainless steel foil). While the barrier concept proves promising, some of the foregoing references involve incompatible materials with either the nuclear fuel (e.g., carbon can combine with oxygen from the nuclear fuel), or the cladding (e.g., copper and other metals can react with the cladding, altering the properties of the cladding), or the nuclear fission reaction (e.g., by acting as neutron absorbers). None of the listed references disclose solutions to the recently discovered problem of localized chemical-mechanical interactions between the nuclear fuel and the cladding. Further approaches to the barrier concept are disclosed in U.S. Pat. No. 3,969,186 issued July 13, 1976 (refractory metal such as molybdenum, tungsten, rhenium, niobium and alloys thereof in the form of a tube or foil of single or multiple layers or a coating on the internal surface of the cladding), and U.S. Pat. No. 3,925,151, issued Dec. 9, 1975 (liner of zirconium, niobium or alloys thereof between the nuclear fuel and the cladding with a coating of a high lubricity material between the liner and the cladding). Accordingly, it has remained desirable to devleop nuclear fuel elements minimizing the problems discussed above. SUMMARY OF THE INVENTION A particularly effective nuclear fuel element for use in the core of a nucler reactor has a composite cladding container comprised of an outer layer having two coatings on the inside surface with the first coating on the outer layer being a diffusion barrier and the second coating on the first coating being a metal layer. The diffusion barrier is comprised of chromium or a chromiun alloy and the metal layer is selected from the group consisting of copper, nickel, iron and alloys of the foregoing. The diffusion barrier can be physically bonded or metallurgically bonded to the substrate and the metal layer can be physically bonded or metallurgically bonded to the diffusion barrier. The diffusion barrier prevents reaction at very elevated temperatures between the outer layer of the cladding and the metal layer, and the metal layer along with the diffusion barrier forms a shield for the outer layer against fission products and gaseous impurities from the nuclear fuel material held in the container during nuclear fission. The metal layer serves as a preferential reaction site for reaction with gaseous impurities of fission products present inside the nuclear fuel element and in this manner the metal layer, as well as the diffusion barrier, serves to protect the outer layers of the cladding from exposure to and attack by the volatile impurities or fission products. Methods of manufacturing the composite cladding are also presented including (a) sequentially electroplating the diffusion barrier and the metal layer on the outer layer and (b) sequentially electroplating the diffusion barrier and the metal layer on the outer layer and heating the outer layer, diffusion barrier and metal layer to produce diffusion and a metallurgical bond between the outer layer and the diffusion barrier and a metallurgical bond between the diffusion barrier and the metal layer. This invention has the striking advantage that the outer layer of the cladding is protected from contact with fission products, corrosive gases and the like by the metal layer and the diffusion barrier, and the metal layer and the diffusion barrier introduce negligible neutron capture penalties. OBJECTS OF THE INVENTION It is an object of this invention to provide a nuclear fuel element capable of operating in nuclear reactors for extended periods of time without the occurrence of splitting of the cladding, internal corrosion of the cladding, or other fuel failure problems. It is another object of this invention to provide a nuclear fuel element with a composite cladding comprising an outer layer two coatings on the inside surface with the first coating being a diffusion barrier and the second coating being a metal layer with the diffusion barrier preventing the metal layer from reacting with the outer layer at very elevated temperatures such as during a loss of coolant accident. Still another object of the invention is to provide a nuclear fuel element with a composite cladding comprising an outer layer having two coatings on the inside surface with the first coating being a diffusion barrier and the second coating being a metal layer and the metal layer protects the outer layer and reacts with fission products and gaseous impurities. Another object of this invention is to provide an economical process for producing a composite cladding for a nuclear fuel element using either electroplating alone or electroplating with a heating step. The foregoing and other objects of this invention will become apparent to a person skilled in the art from reading the following specification and the appended claims with reference to the accompanying drawings described immediately hereinafter. |
062663860 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows nuclear reactor containment building 10. The containment building 10 includes a floor 12, a plurality of walls 14 made from shielded concrete. The floor 12 is submerged under water 16 containing boric acid. The floor 12, includes a reactor vessel pit 20, and a maintenance bay 40. Both the reactor vessel pit 20 and the maintenance bay 40 are under water 16. The reactor vessel pit 20 has a generally cylindrical side wall 22 and a bottom surface 24. The side wall 22 includes a support means 26 to support a reactor vessel 100 (described below). Maintenance bay 40 includes work stands 42, 44. Maintenance bay 40 may have a floor 41 which is at a lower level than floor 12. Containment building 10 also includes a moveable polar crane 70 above water level 16. Polar crane 70 is moveable between a position above the reactor vessel pit 20 and the maintenance bay 40. Typically, a reactor vessel 100 is disposed within the reactor vessel pit 20. As shown on FIG. 2, the reactor vessel 100 is cylindrical with a hemispherical bottom head 102, a cylindrical body 104, and a flanged removable upper head 106. A plurality of control rod drive mechanisms 107 are mounted on upper head 106. The drive mechanisms 107 are each coupled to a control rod drive shaft 108 which passes through openings 103 in upper head 106. Each control rod drive shaft supports a control rod cluster 109 which may be inserted or removed from the reactor core 220 (described below). The reactor pit bottom surface 24 may include hollow columns 200 which allow instrumentation 204 to pass therethrough. The reactor vessel body 104 has at least two openings 110, 112 which allow inlet nozzle 133 and outlet nozzle 134 (shown on FIG. 3) to pass therethrough. The reactor vessel 100 is supported by the support means 26 located in the reactor vessel pit 20. The support means 26 may include a ledge on the lower surface of openings 110, 112. Within the reactor vessel 100 is an upper internal assembly 120 and a lower internal assembly 170. The upper portion of the reactor vessel 100 forms a support ledge 101 which partially supports the upper and lower internal assemblies 120, 170. The upper internal assembly 120 includes the top support plate 122, support columns 124, control rod guide tubes 126 and the upper core plate 128. The principal function of these structures are to align and locate the upper end of the fuel assemblies 222 (described below), and protect and guide control rod clusters 109 as they are inserted and removed from the reactor core 220. Upper support plate 122 is generally cylindrical having a plurality of openings therethrough 123. A plurality of lower control rod guide tubes 142 are aligned with a portion of the plurality of openings 123 through the upper support plate 122. These lower guide tubes 142 are further aligned with the control rod drive shaft openings 103 passing through the upper head 106. Upper core plate 128 is generally cylindrical having a plurality of openings 129 therethrough. A portion of these openings 129 are aligned with the openings 123 in the upper support plate 122 and allow control rod clusters 109 to be inserted and removed from the reactor core 220. Other openings 129 allow water to pass through the upper core plate 128. As shown in FIGS. 2 and 3, the lower internal assembly 170 includes the core barrel 172, a baffle assembly 171 (shown in FIG. 3), the lower core plate 176, the core support forging 179, a tie plate assembly 180, energy absorbers 194 and a secondary core support 196. The core barrel 172 has an upper end 182 and a lower end 184. As shown in FIG. 2, the core barrel 172 includes one or more outlet ports 132 (only one outlet port is shown) which are each coupled to an outlet nozzle 134. Outlet nozzle 134 is in fluid communication with outlet pipe 135 which communicates with a steam generating vessel (not shown). The upper end 182 of the core barrel 172 includes a flange 173 extending perpendicularly outwardly from the core barrel 172. The flange 173 rests on the internal support ledge 101 of the reactor vessel 100. The core barrel flange 173 has a plurality of threaded openings 174. In the preferred embodiment there are three threaded openings 174. The core barrel lower end 184 includes a radial support 186 which is coupled with the lower core support forging 179. The lower core barrel radial support 186 may be welded to the lower core support forging 179. The core barrel 172 is spaced apart from the reactor vessel 100, forming a plenum 80 therebetween. The lower core plate 176 is generally cylindrical and includes a plurality of openings 190. The lower core plate 176 is spaced above the core support forging 179. Baffle assembly 171 is disposed between upper core plate 128 and lower core plate 176. As shown on FIG. 4, the baffle assembly 171 has a perimeter comprised of flat surfaces connected at 90.degree. angles extending within the cylindrical perimeter of core barrel 172. Lower core support forging 179 is disposed below the lower core plate 176. A plurality of lower core plate columns 177 extend upwardly from the lower core support forging 179, supporting the lower core plate 176. The core support forging 179 also includes a plurality of openings 192 which are aligned with the openings 190 in the lower core plate 176. The core support forging 179 is supported by the core support columns 200 which extend downwardly therefrom and pass through the lower head 102 of the reactor vessel 100. Each core support column 200 includes a coupling 201 to attach the support column to the lower core support forging 179. The core support columns 200 are stabilized by a tie plate assembly 180. The tie plate assembly 180 is generally circular and includes a plurality of openings 181 to allow support columns 200 to pass therethrough. A portion of the core support columns 200 are coupled with a secondary core support 196. The secondary core support 196 is a disk having an arcuate bottom surface which is adapted to match the curvature of the reactor vessel 100. The core support columns 200 are hollow and enclose instrumentation 204. The instrumentation 204 extends through support means 26 and columns 200 from the outer side of the reactor vessel 100 through the lower head 102 and through core support forging 179. The instrumentation 204 is further extended through lower core plate support columns 177 and through lower core plate 176 and extending into the reactor core 220. The nuclear fuel cells 222 are disposed within the core barrel lower portion 176 between the upper core plate 128 and the lower core plate 176 and within the baffle assembly 171. This area is the reactor core 220. A plurality of control rod clusters 109 may be inserted in or removed from the reactor core 220 as required using the control rod drive shaft 108. In operation, water passes through an inlet nozzle 133 into the plenum 80 between the reactor vessel 100 and the core barrel 172. The water is drawn downwards towards the lower end of the reactor vessel 100. As the water passes below the lower core support forging 179 it is drawn upwards through the openings 190, 192 in the lower core forging and lower core plate. The water passes into the reactor core 220 where it is heated by the fuel cells 222. The heated water rises through the reactor core 220 and passes through the plurality of openings in the upper core plate 129. The heated water then exits the reactor vessel 100 through the outlet nozzle 134. Maintenance of the upper internal assembly 120 is well known in the prior art. To reach the upper internal assembly 120, the upper head 106 of the reactor vessel 100 and associated components are removed. The upper internal assembly 120 can also be removed. With the upper internal assembly 120 removed, the fuel cells 222 may be removed from the reactor core 220. (As shown in FIGS. 3 and 4) When this operation is complete, only the lower internal assembly 170, as shown in FIG. 3, remains within the reactor vessel 100. FIGS. 5-9 show an up-ending device according to the present invention. The up-ending device allows the removal of the lower internal assembly 170 from the reactor vessel 100 so that the lower internal assembly 170 may be inverted for maintenance operations. The up-ending device includes a frame assembly 400 (shown in FIGS. 9A-9E) and a spider assembly 500 (shown in FIG. 7D). The spider assembly 500 is, generally, fitted within the core barrel 172 and baffle assembly 171 to support the baffle assembly 171 and core barrel lower portion 172 during the up-ending procedure. The frame 400 is fitted about the core barrel 172 and is attached to the overhead polar crane 70. As shown in FIGS. 7A-7D, the spider assembly 500 includes a central column 510 and at least one baffle support plate assembly 520. In the preferred embodiment, there are three baffle support plate assemblies 520, 522, 524. Central column 510 includes a means to support baffle support plate assemblies 520, 522, 524. In the preferred embodiment the support means is a first, second and third set of intermittent partial flanges 516, 517, 518. The first set of intermittent partial flanges 516 is located on the lower portion of central column 510. A second set of intermittent partial flanges 517 is located on the medial portion of central column 510. The third set of intermittent partial flanges 518 is located at the top of central column 510. Spider assembly central column 510 includes a plurality of lower projections 512 which are sized to engage lower core plate openings 190. The spider assembly central column 510 further includes a lifting bale 514 located at its upper end. As shown on FIG. 4, the baffle assembly 171 has a perimeter comprised of flat surfaces connected at 90.degree. angles extending within the cylindrical perimeter of the core barrel lower portion 172. The baffle support plate assemblies 520, 522, 524 are shaped to fit within the jagged perimeter of the baffle assembly 171, that is, the baffle support plate assemblies 520, 522, 524 have the same cross-sectional shape as the baffle assembly 171. The baffle support plate assemblies 520, 522, 524 further include a medial hole 526 which is sized to fit around a spider assembly central column 510. Each baffle support plate assembly 520, 522, 524 includes a plurality of lifting rings 527 which may be coupled with a lifting device, such as crane 70. The baffle support plate assemblies 520, 522, 524 further include a plurality of plunger assemblies 530 mounted adjacent to the outer perimeter of the baffle support plates 520, 522, 524. The plunger assemblies 530 act as an engaging means to secure the baffle assembly 171. As shown in FIGS. 6A and 6B, the plunger assembly includes a horizontal hollow tubular member 532 attached to a baffle support plate 520, 522, 524. A plunger head 534a is disposed within the hollow tubular member 532. The hollow tubular member 532 further includes an elongated opening slot 533 on at least one side, preferably on two sides adjacent to the plunger head 534a. A rotatable vertical member 536 extends upwardly at a generally 90.degree. angle from a baffle support plate 520 and through tubular member 532. The top of vertical member 536 include a plurality of flat surfaces 542, which may be coupled to a plunger engaging tool 501 (shown in FIG. 7A). The vertical member 536 includes an upper threaded portion 538. A collar 544 having a threaded inner surface is disposed on vertical member 536 engaging threaded portion 538. Collar 544 is coupled to one end of a diagonal member 546. The diagonal member 546 is coupled at the other end to plunger head 534a through opening 533. In operation, as vertical member 536 is rotated, collar 544 moves vertically causing the lower end of diagonal member 546 to move plunger head 534a horizontally. As the plunger head 534a is moved horizontally, it may be biased against baffle assembly 171. The plunger head 534a may be flat, as shown on FIGS. 6A and 6B, or may be contoured, as shown on FIG. 6C. As shown on FIG. 5, the plungers 530 are preferably evenly disposed about baffle support plates 520, 522, 524. Plungers 530 which are disposed adjacent to flat sides of the baffle assembly 171 will have a flat head 534a. A plunger 530 disposed adjacent to corners of baffle assembly 171 and will have contoured heads 534b. As shown in FIGS. 7A-7D, the spider assembly 500 is installed by attaching the lowest baffle support plate assembly 520 to central column 510. Central column 510 is passed through the medial opening on baffle support plate assembly 520, and the baffle support plate assembly 520 is lowered until it rests upon partial flange 516. Cutouts 519, as shown in FIG. 5, allow baffle support plate 520 to pass over partial flanges 517, 518. Using polar crane 70 attached to lifting bale 514, central column 510 and baffle support plate assembly 520 is inserted in the lower internal assembly 170. The central column 510 is lowered until lower projections 512 are mounted within lower core plate openings 190. Once the column 510 and the first baffle support plate 520 are positioned, the plurality of plunger assemblies 530 on are engaged the first baffle support plate 520 by rotating each vertical member 536 until each plunger head 534a, 534b engages baffle assembly 171. As shown in FIG. 7B a second tier baffle support plate 522 is then lowered into place until positioned on the second set of partial flanges 517 in the medial portion of spider assembly central column 510. The baffle support plate assembly 522 may be coupled to the overhead crane 70 by lifting rings 527. Again, once the baffle support plate assembly 522 is positioned, plunger assemblies 530 are engaged with the baffle assembly 171. As shown in FIG. 7C, baffle support plate assembly support columns 560 may be installed on the second tier baffle support plate assembly 522 to support the third tier baffle support plate assembly 524. The third tier baffle support plate assembly 524 is then lowered into the lower internal assembly 170 until the third tier baffle support plate assembly 524 is generally aligned with the top of spider assembly central column 510. Third tier baffle support plate assembly 520 rests upon partial flange 518 located at the top of the spider assembly central column 510. Again, the plunger assemblies 530 are engaged to secure the baffle assembly 171. As shown on FIG. 7D the spider assembly further includes an upper brace assembly 581 which includes a lower baffle support ring 582, a plurality of hollow support columns 584 and an upper lifting plate assembly 590. Hollow columns 584 are disposed below lifting plate assembly 590. Ring 582 rests upon the upper edge of baffle assembly 171. As shown in FIG. 7E, the support columns 584 each have a lower end 585 which includes a threaded opening 586. A height adjustment means, such as a floatable pad 588 coupled to a threaded rod 589, is disposed within threaded opening 586. The support columns 584 are coupled to the lower side 583 of upper lifting plate assembly 590 and are aligned with ring 582. As shown in FIG. 8A lifting plate assembly 590 includes a generally circular planar disk 591 having a diameter approximately equal to the core barrel upper flange 173. Disk 591 includes openings 593 which are aligned with hollow support columns 584. Openings 593 allow access to the interior of the support columns 584 so that a tool may be inserted to rotate threaded rod 589, thereby biasing pad 588 against ring 582. When lifting plate assembly 590 is lowered onto the lower internal assembly 170, the lower end of the support columns are adjacent to ring 582. Lifting plate assembly 590 further includes two parallel cross bars 602, 604 disposed on disk lower surface 583. Cross bars 602, 604 include a plurality of threaded harness holes 610, 612, 614, 616, one each located at each end of the cross bars 602, 604. When the lifting plate assembly 590 is installed, only cross bars 602, 604 contact core barrel upper flange 173. Thus, reducing the possibility of damaging the flange surface. Lifting plate assembly 590 further includes an attaching means and a lifting harness attachment means 620. The preferred embodiment of the attaching means and the lifting harness attachment means 620 are a plurality of threaded fasteners 620 which extend through disk 591. In the most preferred embodiment, there are three fasteners which are threaded into lower internal assembly flange threaded openings 174 (shown on FIG. 3). By installing the threaded fasteners through the lifting plate assembly 590, the lifting plate assembly 590 is attached to the lower internal assembly 170. The threaded fasteners 620 are coupleable to a lifting harness. The fasteners may have, for example, threaded bore holes 620 which a harness 800 (described below) may engage. Thus, when installed the spider assembly 500 supplies a radial force, through plungers 530, to baffle assembly 171 as well as a compressive force, through ring 582. Additionally, lifting plate assembly 590 provides a secure attachment to the lower internal assembly 170. As shown on FIGS. 9A-9E, the frame assembly 400 has two mirror image sides 401, 402 (Side 401 is shown in FIG. 9A) which are spaced apart by cross braces 425, 426, 427, 428 (shown in FIG. 9B). Accordingly, it is understood that certain members, e.g. braces 430, 432 (shown in FIG. 9A) on side 401, have unseen counterparts on side 402. The frame assembly 400 includes first members 410, 412, second members 414, 416, front members 418, 420 and rear members 422, 424. There are additionally cross braces 430, 432, 434, 436 and 438. Lower plate assembly 440 (shown in FIG. 9B) is disposed between second members 414, 416. Shielding plates 450, 452 are disposed adjacent to second members 414, 416. The frame 400 further includes secondary core support saddle 460, tie plate assembly support saddle 464 and core barrel support saddles 476, 477. As shown on FIG. 9A, side 401 (which is mirrored, but not shown on side 402) of the frame 400 is generally rectangular and includes a first member 410 which has a front end 480 and a rear end 481. First member rear end 481 is coupled to one end of end of rear member 422 at a 90.degree. angle. First member front end 480 is coupled to one end of end of front member 418 at a 90.degree. angle. Second member 414 has a front end 484 and a rear end 485. The ends of front member 418 and rear member 422 opposite the connection with first member 410 are coupled to second member 414. Second member rear end 485 is coupled to one end of rear member 422 at a 90.degree. angle. Second member front end 484 is coupled to one end of end of front member 418 at a 90.degree. angle. Each end of rear member 422 includes an arcuate corner portions 453, 454 which each have a outer surface 431, 433. As will be described below, the frame assembly is rotated on outer surfaces, 431, 433. Front member 418 and rear member 422 are also connected by a plurality of braces 430, 432, 434, 436, 438 which are connected to tabs 429 on the front member 418 and tabs 421 on the rear member 422. As shown on FIG. 9B, cross braces 437, 439 are disposed between rear members 422, 424. As shown on FIGS. 9B-9D, sides 401, 402 are spaced apart by braces 425, 426, 427, 428. Brace 425 is disposed between sides 401, 402 adjacent to rear members 422, 424 upper rounded corner 453. Braces 426 and 427 are disposed between sides 401, 402 spaced along the medial portion of rear members 422, 424. Brace 428 is disposed between sides 401, 402 at rear member lower rounded corner 433. In the preferred embodiment, frame assembly second members 414, 416 are I-beams having an upper flange 415, 417. The I-beam upper flange 415, 417 is above the bottom surface of second members 414, 416. Each upper flange 415, 417 has a plurality of threaded fastener holes (not shown). Detachable frame lower plate assembly includes a planar member 445. Frame lower plate planar member 445 is attached by fasteners 441 to upper flange 415, 417. Frame lower plate assembly 440 further includes lifting rings 442 disposed on the outer surface of frame lower plate planar member 445. Frame lower plate assembly lifting rings 442 may be coupled to a lifting means such as crane 70. As shown on FIG. 9B, a plurality of lower internal assembly support columns 443 extend upwardly from the interior surface of frame lower plate planar member 445 terminating in distal ends 447. A lower internal assembly support arc 446 is disposed at the distal ends of the plurality of lower internal assembly support columns 443. As shown on FIG. 9B, core barrel saddles 476, 477 are disposed between rear members 422, 424 adjacent to braces 426, 427. In the preferred embodiment, core barrel saddle 476 is integral to brace 426 and core barrel saddle 477 is integral to brace 427. As shown on FIG. 9D, each core barrel saddles 476, 477 includes an upper arcuate surface 478, 479 which is sized to fit the outer diameter of core barrel 172. Tie plate assembly support saddle 464 includes a support member 465 and a saddle pad 466. Saddle pad 466 includes an arcuate surface 467 which has a curvature matching that of tie plate assembly 180. Secondary core support saddle 460 includes a support member 461 and a secondary core support saddle 462. Secondary core support saddle 462 also includes an arcuate surface 463 which is sized to match the curvature of the secondary core support 196. When the lower internal assembly 170 is installed within Frame 400, secondary core support 196 is adjacent to secondary core support saddle 460. Tie plate assembly 180 is adjacent to tie plate assembly support saddle 464. Core barrel 172 is adjacent to core barrel saddles 476, 477. When the up-ender device is laid horizontally each of these components will rest upon the respective support saddles. Lifting lugs 490 are located at each corner of sides 401, 402. As shown on FIG. 9C, each lifting lug 490 includes a cylindrical base 491 and an outer disk 492. The outer disk 492 has a larger diameter than the cylindrical base 491. In operation, the lower internal assembly 170 fits within sides 401, 402 of frame assembly 400. In the upright orientation, the lower core support forging 179 rests upon lower internal assembly support arc 446. L-shaped retainer 406 prevents the lower internal assembly 170 from tipping out of frame 400. Crane 70 is connected to lifting lugs 490 and is used to rotate the frame 400. When in the horizontal orientation, core barrel saddles 476, 477, tie plate assembly support saddle 464, and secondary core support saddle 460 support the lower internal assembly 170. When in the inverted upright orientation the lower internal assembly 170 rests upon lifting plate assembly 590. FIGS. 11A-11H show the up-ending procedure according to the present invention. The up-ending procedure is accomplished as follows. While the internal assemblies 120, 170 are still with in reactor vessel 100 (not shown), the upper internal assembly 120 and fuel cells 222 are removed as described above. The core support columns 200 are de-coupled from the lower internal assembly 170. As is known in the prior art, the lower internal assembly 170 is then removed from the reactor vessel 100 and placed on storage stands 42, 44. As shown in FIG. 11A, up-ending frame 400 is then positioned on the floor 12 between pit 20 and the maintenance bay 40. As shown in FIGS. 7A-7D, the spider assembly 500 is then installed, as described above, to support the baffle assembly 171 of the lower internal assembly 170. Installing the spider assembly 500 includes coupling the lifting plate 590 to the core barrel upper flange 173. The polar crane 70 is then coupled to the lifting plate 590. In the preferred embodiment, a harness 800 having three lifting rods 801 is coupled to lifting plate assembly 590 through lifting means 620. In the preferred embodiment, the lifting means 620 are threaded fasteners with threaded bore holes 622, and lifting rods 801 include threaded tips that may be engaged with the threaded bore holes 622. Crane 70 may then lift the lower internal assembly 170 off maintenance bay support structures 42, 44. As shown on FIG. 11B, using polar crane 70, the lower internal assembly 170 is then lifted to a position adjacent to the up-ending frame 400. The lower internal assembly 170 is then translated horizontally into the up-ending frame 400. Once positioned within frame 400, the lower internal assembly 170 is lowered until lower core support forging 179 rests on lower internal assembly support arc 446. As shown on FIG. 9C, cross bar 405 is attached between frame first members 410, 412. Cross bar 405 includes an L-shaped retainer 406 disposed on the medial portion of cross bar 405. The L-shaped retainer includes a tab 407 which extends downwardly adjacent to core barrel 172. Tab 407 prevents the lower internals 170 from tipping out of frame 400. As shown in FIG. 11C, the up-ending frame 400, which now holds lower internal assembly 170, is rotated on lower rounded corners 431 into a horizontal orientation. As shown in FIG. 11D, because the floor 12 in the preferred embodiment is at a higher elevation than maintenance bay floor 41. An additional A-frame support member 45 may be installed on maintenance bay storage stands 42, 44. Frame 400 is then translated horizontally until lower members 422, 424 rest on A-frame 45. As shown in FIG. 11E, the up-ending frame 400 is then pivoted about upper rounded corners 431, 433 until the frame 400 is in the inverted, vertical orientation. As shown in FIG. 11F, the up-ending frame 400 is then rotated 180 degrees about a central axis so that the open side of frame 400 is adjacent to the maintenance bay 40. The crane 70 is then detached from the frame assembly 400 and coupled to the lower internal assembly 170. In the preferred embodiment, the crane 70 is coupled to a four legged harness 802. Each leg of the four-legged harness 802 is a rod 804 having a threaded tip 805. The rods 804 pass through one of the harness openings 449 on frame lower plate assembly 440. Each rod 804 is lowered along the outside of the lower internal assembly 170 until each rod tip 805 engages threaded harness attachment holes 610, 612, 614, 616 on cross bars 602, 604. The crane 70 then lifts and translates the lower internal assembly 170 horizontally out of up-ending frame 400 and lowers the lower internal assembly 170 on to maintenance bay storage stands 42, 44. Thus, as shown in FIG. 11H, the procedure terminates with the lower internal assembly 170 being in an inverted orientation on the maintenance storage stands 42, 44. The re-inversion procedure consists of performing the above steps in reverse order. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of invention which is to be given the full breadth of the claims appended and any and all equivalents thereof. |
abstract | A system and method relating to a radiation based imaging are provided. The system may include a radiation source, a detector and a first grid. The detector may include a plurality of detector cells. The first grid may be located between the radiation source and the detector cells and the first grid may include a plurality of radiation transmitting sections. At least one of the plurality of detector cells may include an active area which may be configured to receive radiation from the radiation source that passes through at least one of the plurality of radiation transmitting sections of the first grid. The active area may be adjustable by adjusting the first grid. The radiation source, the first grid and the detectors cells may be operatively coupled for detecting an object. The method may include adjusting the first grid to adjust the active area of the detector. |
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summary | ||
abstract | The invention is intended to increase the number of patients treatable using one wheel having a thickness varied in the rotating direction to change energy of an ion beam passing the wheel. Ion beam delivery equipment for irradiating an ion beam to a patient for treatment comprises a beam generator for producing and accelerating the ion beam, an beam delivery nozzle including a range modulation wheel which has a predetermined thickness distribution in the rotating direction and is rotated on a travel passage of the ion beam generated from the beam generator to control a range of the ion beam, and an irradiation controller for controlling the beam producing and accelerating operation of the beam generator in accordance with the phase of rotation of the range modulation wheel. |
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047132108 | description | Referring to FIG. 1, a control rod R and a control rod drive D are illustrated connected by the driveline L. The control rod is provided with a handle section H which handle section is engaged by a grapple section G. Drive D is conventional and only schematically illustrated. Specifically a drive gear motor 14 revolves a screw drive 16 which is received in a female drive nut 18 on the side of the driveline L. Presuming rod R is engaged at handle H by grapple G, up and down movement of drive nut 18 correspondly moves driveline L to effect withdrawal and insertion of rod R into the reactor. Driveline L consists of two discrete segments. First, there is a tension rod T. Second there is an outer cylinder C. As will hereinafter become more apparent by relative downward movement of tension rod T relative to cylinder C, release of grapple mechanism G to drop handle H can occur. Such movement can occur by one of two expedients. First, it will be seen that tension rod T includes an upper flange 20 for abutment to an electromagnet 22. Providing magnet 22 is energized, tension rod T will remain in the upper position with flange 20 abutted to electromagnet 22. Grapple G will remain closed with handle H therein. If magnet 22 has its power interrupted, the rod is dropped. Secondly, it will be understood that tension rod T is made of a material having a high index of thermal expansion and cylinder C is made of a material having a low index of thermal expansion. Upon thermal heating of the drive line L, tension rod T will move downwardly with respect to cylinder C. As will be hereinafter be set forth, the grapple mechanism G will open. Handle H will be released. Rod R will drop interiorly of the reactor. It will be understood that the control rod mechanism is preferably used interior of a sodium cooled fast breeder reactor. Accordingly, sodium and an inert cover gas 26 are maintained at the top of the reactor. In order to maintain isolation of the inert cover gas from the atmosphere, a first exterior bellows 28 surrounds the driveline L and is connected between the reactor at 30 and the driveline at 32. A second bellows mechanism is connected between the interior of cylinder C at 34 and the tension rod at 36. The reader will understand that a fluid type seal exists within the driveline L to prevent contamination with atmospheric gas. Having set forth the overall construction of the drive mechanism, attention will now be devoted to FIGS. 3A, 3B, 4A, 4B and 4C. The handle segment H at the upper end of the rod R will first be discussed with reference to FIG. 3A. Thereafter, the grapple segments will be set forth with reference to FIGS. 3A, 3B, 4A, 4B, and 4C. Rod R is typically connected by a cylindrical rod handle 40. As shown here, handle 40 flares outwardly in an inverted frustum to expanded cylindrical portion 43. Thereafter, the handle portion again decreases in section at contracting frustum portion 44. An upward flange portion 45 terminates the upper handle surface. A cylindrical centering pin 47 protrudes upwardly from handle surface 45. It terminates in a gathering pin point 49. Handle H forms essentially a male member. This male member is received in the female concavity of the grapple G. Centering pin 47 is received in the lower portion of the tension rod T, which lower portion will now be described. Tension rod T at lower end 50 defines a flange 52. Flange 52 includes a central pin receiving aperture 54. Aperture 54 includes lowered gathering surface 56. It will therefore be understood that when tension rod T is lowered towards handle H, centering of the pin 47 within the aperture 54 will occur. Turning now to FIG. 4A, a grapple segment S is illustrated. Seeing FIG. 4B it can be seen that each grapple segment S comprises 60.degree. (degrees) of a total grapple mechanism G. Grapple segment S will first be described with respect to that section which confronts the handle H. Thereafter, the grapple segment will be described with respect to the exterior portion which confronts the interior lower portion of cylinder C. Each grapple segment is rounded. Since six grapple segments define a circular grapple, the segments each include 60.degree. (degrees) of curvature. For simplicity, the following description will include a description of the profile of a longitudinal section. The reader's understanding of the curvature of all the segments will be assumed. There are two exceptions to this curvature. These exceptions are surfaces 72 and 74. As will hereinafter become apparent, these are surfaces on which the segments rock. A straight surface is preferred. Grapple segment S includes a lower gathering surface 60. This surface typically bears against frustum 44 on handle H when handle H is received causing the segment S to pivot outwardly. (See FIG. 3A.) A constriction point 64 includes an upper beveled segment 62. When six such segments are combined, the respective combined segments provide a female cavity on which frustum 42 of handle H rests to effect engagement between the grapple G and the handle H. (See FIG. 3A again.) The interior of the segments S must contain in the entirety the handle H. Therefore when frustum 42 is contained at segment 62, cylindrical portion 43 abuts indentation 63. Upper member 68 accommodates the profile of frustum portion 44. Segment S is designed for pivot about flange 52 on the lower end of tension rod T. Accordingly, there is a flange receiving indentation 70. Indentation 70 includes upper bearing surface 72 for bearing on the top surface of flange 52 and lower bearing surface 74 for bearing on the bottom surface of flange 52. Pivotal interaction between the top surface 45 of the enlarged portion of the handle H and each segment is required. Therefore an inwardly extending segment member 76 is provided with a lower bearing surface 78. As will hereinafter be emphasized with respect to the view of FIG. 3A, the interaction between surface 78 and surface 72 of flange receiving portion 70 causes the grapple segment to move inwardly relative to the other grapple segments to cause capture of the handle H within grapple G. The outer portion of the grapple segments S is simpler of detail. Three major portions of this outer segment are of concern. First, opposite flange receiving portion 70 the outer segment S is given at section 80 a thickness so that the inner portion of the cylinder C is loosely abutted. This thickness makes sure that the assembled grapple G is maintained firmly about the lower flange 52 on the tension rod T. Secondly, each segment is provided with a lower protruding flange portion 82. Lower protruding flange portion 82 restricts opening of each segment S of the grapple to close the grapple G itself. Specifically, when segment 82 is within the cylinder C, the grapple is closed and the handle H may not enter or leave. Finally, and between shoulder 80 and annulus 82, there is a tapered portion 84 of the segment. Tapered portion 84 gradually constricts the overall dimension of the grapple G in a downward flaring frustum shape. This shape ends at annulus 82. This portion 84 ensures that when annulus 82 clears the bottom of cylinder C, the grapple segment may pivot outwardly so as to release and/or receive handle H. Having completely described the handle H, the single grapple segment S, the assembled female configuration of the grapple G by confrontation of six of the segments S may be understood. First, the assembled segments S form a flange receiving portion 170. Secondly, the assembled segments form a gathering portion 160. Finally, the assembled segments form a handle capturing portion 163. Bearing in mind these respective portions, attention will now be directed to the cartoon series of FIG. 2A together with the sections of FIGS. 3A and 4A to describe operation. First, assembly to the configuration of FIG. 2A can easily be understood. Six segments S are assembled around a tension rod T at flange 52. The tension rod T and flange 52 are moved interior of the cylinder C. Cylinder C captures the segments between flange portion 70 and shoulder 80. Grapple segments S are constricted as a unitary body about the lower part of the tension rod T. To initially engage a rod R at handle H with respect to FIG. 2A, tension rod T is released at flange 20 by electromagnet 22. Flange 82 of the grapple segments S extends beyond the lower portion of cylinder C. The grapple segments S can all open. The cavity interior of the grapple G is open to receive the rod R at handle H. This occurs because gathering surfaces 60 tend to pivot the grapple segments S outwardly. In the view of FIG. 2B, drive D (not shown in the view of FIG. 2) has been lowered. Grapple G has come into contact with handle H. Even though the flange 82 is not within the cylinder C, closure of the grapple segments S about the handle H occurs. The action by which this occurs can best be explained with references to FIGS. 3A and 4A. Upon downward movement of the tension rod T, shoulders 74 (see FIG. 4A) and 78 bear respectively against the lower portion of the tension rod T and the flat handle surface 45. Since the respective shoulders 72 and 78 are separated by a lever arm, the particular grapple segment in FIG. 4A attempts to pivot counterclockwise as shown in the view of FIG. 4A. In such counterclockwise pivot, the grapple segments 64 all move to and towards one another. Consequently, handle H is captured at frustum 42 by the constriction 64. Just as a single segment S moves singularly, all segments S move inwardly collectively. This being the case, grapple G is configured so that when cylinder C moves downwardly relative to the grapple G, all the segments S are confined within it (see FIG. 2C). When the driveline L begins to move upwardly, handle H falls away interior of the female segment defined by the grapple G (see FIG. 2D). However, release cannot occur if electromagnetic 20 engages flange 22. With this engagement, rod R is fully coupled to driveline L. Referring to FIG. 2E, it is easy to understand how full withdrawal of rod R can occur. Assuming full withdrawal of rod R occurs, all that remains to be explained is the thermal release of FIG. 2F and the electromagnetic release of FIG. 2D. It will be understood that any relative movement sufficient to clear flange 82 of the lower portion of cylinder C will cause release of the handle H. Referring to FIG. 2F, it will be remembered that tension rod T is constructed of a material having a relatively high coefficient of expansion. Cylinder C is constructed of a material having a relatively low coefficient of expansion. Further, it is well known to design such bimetallic parts to have precise movement responsive to overall temperature conditions. Specifically, when tension rod T expands, a large amount and cylinder C expands only a small amount, flange 82 clears the lower portion of cylinder C. Grapple segments S open and handle H is released. Rod R falls and causes SCRAM responsive to its penetration within the reactor core (not shown). Finally, and with respect to FIG. 2G, control circuits can cause a release of the current to the electromagnetic portion 22. Flange 20 at the upper end of tension rod T is released. When the tension rod is released again, flange 82 clears the lower portion of cylinder C. Handle H is released with the result that rod R effects full core penetration and responsive SCRAM. The reader will understand that we have illustrated only a single reactor rod. In actual fact many will be used for control of a reactor. It will be further understood that the particular grapple mechanism here illustrated is exemplary only. For example, the grapple segments shown could be virtually any shape which would co-act with the lower and cylindrical portion of the cylinder to restrict a handle captured in the interior of the device. It will be apparent also that the handle H does not have to have the particular preferred shape here illustrated. The handle could be spherical in shape. Likewise it could be given any imparted shape which could be received and restricted in the lower end of the driveline L. It will be also understood that the disclosed driveline and grapple can be used with many alternate drives. Examples of some drives include rack and pinion, hydraulic, pneumatic and other equivalent mechanical and electromechanical expedients. Likewise, magnet 20 and flange 22 may have equivalent devices substituted, such as pneumatic, hydraulic and other mechanical and electromechanical expedients. Likewise, the manner in which the grapple segments attach to the rod can vary. For example, a pivotal attachment between the lower end of the tension rod and the grapple segments may be used. Additionally, the number of grapple segments may vary, although the illustrated six segments are preferred. |
046559914 | description | DETAILED DESCRIPTION OF THE INVENTION With reference now to the drawings, FIG. 1 illustrates a linear matrix array 82 of helical springs 90, such as may be used in a nuclear fuel assembly. The fuel assembly typically includes a matrix array of fuel rods wherein each fuel rod comprises an elongate tube containing a fissionable fuel material. Each tube is sealed at opposite ends by means of end plugs, e.g. as shown in U.S. Pat. No. 4,022,661, which engage corresponding tie plates. Helical springs 90 are fitted around shanks 92 of upper end plugs 88 to allow expansion of the fuel rods and to insure firm seating of the latter. In row B--B of matrix array 82, shank 92' of end plug 88' is shown with its helical spring missing. The location of shank 92' within array 82 is such as to be inaccessible to visual inspection without removing the upper tie plate. Probe 10, which constitutes the subject matter of the present invention, includes an elongate arm 12 formed of a rigid material such that the arm will not warp and lose its linearity with repeated use. Arm 12 has a substantially rectangular cross section, having a width defined by first and second parallel sides 16 and 18 enabling it to be inserted between adjacent rows of helical springs 90 and a height less than the height of the springs. The length of arm 12 is sufficient to probe the full length of a row of helical springs, e.g. the length of rows A--A and B--B in linear matrix array 82. Arm 12 includes a forward end 24 and a holding end 26. A rectangular aperture 28 is formed near the forward end of the arm. The height of aperture 28 is defined by surfaces 30 and 32, and its length by surfaces 34 and 36. A pivot pin 42 is transversely positioned in aperture 28, parallel to sides 16 and 18, and with opposite pin ends embedded in surfaces 30 and 32. First and second substantially identical pawls 48 and 50 are mounted in superposed relationship on pivot pin 42 within aperture 28. Each pawl is capable of rotational movement about pivot pin 42, between a retracted and an extended pawl position and each is constrained from rotating beyond substantially a ninety degree arc with respect to arm 12. Each pawl includes first and second pawl ends 52 and 54 respectively, a forward pawl surface 60 which has a rounded surface portion near pawl end 54, and a flat rear pawl surface 62. The angle of surfaces 60 and 62 relative to each other is such as to define a surface discontinuity at pawl end 54. In FIG. 1, the respective pawl portions are designated with letter subscripts to permit separate reference to each. The dimensions of the aperture and the pawls are such that the pawls are completely contained in the aperture in their retracted positions e.g. as shown by pawl 48 in FIG. 1. Both pawls are notched near their respective first ends 52 to receive a torsion spring 68 which is coaxially disposed on pivot pin 42. Spring 68 applies a bias to each pawl, urging end 54a of pawl 48 out of aperture 28 beyond side 18 of arm 12 and urging end 54b of pawl 50 out beyond side 16. A handle 74 is affixed to arm 12 near holding end 26 by means of suitable fasteners 76 and 78 such as screws or rivets and provides a hand grip for the manual operation of probe 10. Handle surface 80 provides a stop against outer structure 94 of the fuel bundle in which the fuel rods are located. This prevents the insertion of probe 10 into linear matrix array 82 to a point beyond the last row of elements C--C in array 82. In operation, in order to detect the absence of any helical springs missing in rows A--A and B--B of the array, probe 10 is inserted horizontally into the space between these rows by way of its forward end 24. As the probe is advanced into linear matrix array 82, contact between pawl surfaces 60b and 60a and the helical springs in rows B--B and A--A, urges pawls 48 and 50 toward arm 12 and thus toward their retracted positions within aperture 28, against the force of the torsion spring. Probe 10 is advanced into array 82 until handle surface 80 abuts outer structure 94. The length of arm 12 is chosen so that the absence of a helical spring in row C--C will allow the appropriate pawl to pivot to its extended position, but not so long as to allow the pawls to assume their extended position when helical springs 90a and 90b are present. The withdrawal of probe 10 proceeds smoothly by virtue of the rounded surface portions of surfaces 60a and 60b near pawl ends 54a and 54b respectively, which make sliding contact with springs 90 in both directions of probe movement. Thus, as probe 10 is withdrawn from the space between rows A--A and B--B, helical springs 90 on the shanks of end plugs 88 continue to urge pawls 48 and 50 toward their respective retracted positions. Hence, as long as all springs 90 are present, the pawls are prevented from assuming their respective extended positions. Testing of the array for absent springs 90 continues by insertion of the probe between subsequent row pairs, until the matrix has been completely checked out. If a helical spring 90 is missing, as indicated in FIG. 2 with respect to shank 92' in row B--B, pawl 50 will pivot to its extended position at the location in question, as urged by torsion spring 68. Upon attempted withdrawal of the probe, the extended pawl becomes lodged against springless shank 92'. Since the pawl is constrained from pivoting more than 90.degree. and the construction of the fuel bundle prevents lifting the probe out, the probe is effectively locked against withdrawal from the array. To retrieve the probe, the upper tie plate of the fuel bundle must be removed. The position of the locked-in probe then indicates to the operator the location of the missing spring, which is replaced before reassembly. The present invention is not limited to use in a linear matrix array. With a suitably curved arm, the probe may also be used where the elements tested for are arranged in curved rows. Also, the probe may carry only a single pawl for use in ascertaining the absence of an element in a single row of elements. It will also be clear that the use of the probe with linear matrix array 82 is not limited to testing rows of elements parallel to rows A--A and B--B, but that it may also be used along rows parallel to row C--C. In a broader sense, the present invention is not limited to testing for the absence of helical springs, but it is applicable to detect the absence of an element from any suitable array of dimensionally similar elements which are regularly spaced from each other. Further, the invention is not limited to the particular embodiment illustrated and described. Numerous variations, changes, modifications, substitutions and equivalents will now occur to those skilled in the art, all falling within the true spirit and scope of the invention. Accordingly, the invention is intended to be limited only by the scope of the appended claims. |
abstract | Disclosed is a coolant with dispersed neutron poison micro-particles, used in a supercritical water-cooled reactor (SCWR) emergency core cooling system. Since the neutron poison micro-particles are uniformly dispersed in the coolant of the emergency core cooling system for a long period time, their fluidity is not lowered even though the polarity of water is changed in a supercritical state. Therefore, the neutron poison micro-particles absorb neutrons produced from nuclear fission in a nuclear reactor core. Accordingly, the neutron poison micro-particles can be appropriately used as a means for controlling neutrons and stopping a nuclear reactor in the SCWR emergency core cooling system. |
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summary | ||
048658006 | claims | 1. In a fuel assembly grid inspection method, the combination comprising the steps of: (a) defining an illuminated inspection field of view; (b) supporting a fuel assembly grid of a design having known standard measurements within the inspection field of view by using a fixture whose portions which project into the field of view to support the grid therein are substantially transparent to the field of view; (c) viewing the grid within the field of view; and (d) recording an image thereof to provide information about the grid from which actual measurements can be calculated and compared to the known standard measurements for the particular grid design. maintaining a video camera and lens system pointed toward the field of view and the grid supported therein; and moving at least one of the grid and the video camera and lens system relative to the other for viewing the grid. (a) defining an illuminated inspection field of view; (b) supporting within the field of view a fuel assembly grid having at least a pair of fuel rod contacting dimples disposed in each cell of a plurality thereof defined in the grid which are adapted to receive fuel rods therethrough; (c) inspecting the pair of dimples for perpendicularity with respect to one another by viewing the dimples at separate instances from the same location within the field of view; and (d) recording a separate image of each dimple to provide information from which actual measurements of any offset in X and Y directions of one dimple with the other can be calculated. (a) defining an illuminated inspection field of view; (b) positioning a fuel assembly grid support fixture about the perimeter of the field of view with portions of the fixture which project into the field of view to support a grid therein being substantially transparent to the field of view; (c) supporting on the fixture within the field of view a fuel assembly grid of a design having known standard measurements; (d) sensing the position of the fixture for providing information to determine the correctness thereof before proceeding with inspection of the grid; and (e) inspecting the grid within the field of view. viewing the grid within the field of view; and recording at least one image thereof to provide information about the grid from which actual measurements can be calculated and compared to the known standard measurements for the particular grid design. 2. The method as recited in claim 1, wherein said supporting step includes providing a fixture being universally adapted for supporting within the inspection field of view any one of a plurality of fuel assembly grids of different designs having different known standard measurements. 3. The method as recited in claim 1, wherein said viewing step includes: 4. In a fuel assembly grid inspection method, the combination comprising the steps of: 5. In a fuel assembly grid inspection method, the combination comprising the steps of: 6. The method as recited in claim 5, wherein said inspecting step includes: |
061907251 | abstract | The present invention relates to a coating method for the preparation of a coated nuclear fuel. Particularly, the present invention relates to the coating method of nuclear fuel surface with more than two coated layers of carbides, borides or nitrides and their compounds comprising deposition or permeation steps of i) elements or mixture that can form carbides, borides or nitrides and ii) a layer of pyrolytic carbon or boron prepared by chemical vapor deposition(CVD) or sputtering in sequence or in reverse sequence, or nitrogen prepared by gas permeation in sequence, on the nuclear fuel surface. The coated layers are formed with carbides, borides, nitrides or their mixture at high temperature and pressure by a combustion synthesis. The coating method of this invention can be applied to various types of nuclear fuels either in particle or in pellet type and control and preserve fine crystal structure without phase transition since the surface of nuclear fuel coated with pyrolytic carbon and silicon is heated only for several seconds by heat source such as laser beam, arc or microwave. Thus, the present invention is excellent method of coating nuclear fuel surface not only for particle type fuel which used in High Temperature Gas-cooled Reactor (HTGR) but also for pellet type fuel used for Water-cooled Reactors. |
claims | 1. Apparatus for X-ray analysis of a sample, comprising: an X-ray source, which irradiates the sample; and an X-ray detector device, which receives X-rays from the sample responsive to the irradiation, the device comprising: an array of radiation-sensitive detectors, which generate electrical signals responsive to radiation photons incident thereon; and processing circuitry comprising a plurality of signal processing channels, each coupled to process the signals from a respective one of the detectors so as to generate an output dependent upon a rate of incidence of the photons on the respective detector and upon a distribution of the energy of the incident photons. 2. Apparatus according to claim 1 , wherein the array of detectors comprises an array of radiation-sensitive diodes. claim 1 3. Apparatus according to claim 2 , wherein the diodes comprise silicon diode detectors. claim 2 4. Apparatus according to claim 1 , wherein each of the plurality of signal processing channels comprises an integrated circuit disposed on a common substrate with the respective detector. claim 1 5. Apparatus according to claim 4 , wherein the common substrate comprises a semiconductor chip including integrated circuits belonging to a multiplicity of the signal processing channels. claim 4 6. Apparatus according to claim 1 , wherein the signal processing channels process the signals in accordance with adjustable processing parameters. claim 1 7. Apparatus according to claim 6 , wherein the processing parameters are adjusted independently for different ones of the channels responsive to different incidence rates of the photons at the respective detectors. claim 6 8. Apparatus according to claim 1 , wherein the signal processing channels comprise discriminators, which reject signals corresponding to photons outside a predetermined energy range. claim 1 9. Apparatus according to claim 8 , wherein the processing circuitry comprises a threshold control circuit, which adjusts the predetermined energy range of the discriminators. claim 8 10. Apparatus according to claim 1 , wherein the signal processing channels comprise counters, which count the number of photons incident on the respective detectors responsive to the energy of the photons, and wherein the processing circuitry comprises a bus common to a multiplicity of the channels, which receives and outputs respective photon counts from the channels in turn. claim 1 11. Apparatus according to claim 1 , wherein the X-ray detector device receives X-rays reflected from the sample. claim 1 12. Apparatus according to claim 1 , wherein the X-ray detector device receives fluorescent X-rays emitted by the sample. claim 1 13. Apparatus according to claim 1 , wherein the X-ray source comprises a monochromator, such that the sample is irradiated with substantially monochromatic X-rays at a predetermined energy. claim 1 14. Apparatus according to claim 13 , wherein the signal processing channels comprise discriminators, which are adjusted to reject signals corresponding to photons outside an energy range including the predetermined energy of the monochromatic X-rays. claim 13 15. A method for X-ray analysis of a sample, comprising: irradiating the sample with X-rays; receiving X-rays from the sample, responsive to the irradiation, at an array of detectors in respective, predetermined locations, which detectors generate electrical signals responsive to X-ray photons incident thereon; and processing the signals from the array of detectors in respective processing channels, so as to generate an output indicative of a rate of arrival of the photons incident at the respective locations and dependent upon a distribution of the energy of the incident photons. 16. A method according to claim 15 , wherein processing the signals comprises providing a plurality of channels each comprising an integrated circuit disposed on a common substrate with the respective detector for processing signals generated by the detector. claim 15 17. A method according to claim 15 , wherein processing the signals comprises processing signals in accordance with processing parameters, which are independently adjustable for different ones of the channels. claim 15 18. A method according to claim 17 , wherein processing the signals comprises adjusting the processing parameters in the channels responsive to an incidence rate of the photons on the detectors. claim 17 19. A method according to claim 15 , wherein processing the signals comprises discriminating signal levels so as to reject signals corresponding to photons outside a predetermined energy range. claim 15 20. A method according to claim 19 , wherein processing the signals comprises counting the number of photons incident at each of the locations within the predetermined energy range. claim 19 21. A method according to claim 20 , wherein irradiating the sample comprises irradiating the sample with substantially monochromatic X-rays at a selected energy, and wherein discriminating the signal levels comprises rejecting signals corresponding to photons outside an energy range including the selected energy of the monochromatic X-rays. claim 20 22. A method according to claim 15 , wherein receiving the X-rays comprises receiving X-rays reflected from the sample. claim 15 23. A method according to claim 15 , wherein receiving the X-rays comprises receiving fluorescent X-rays emitted by the sample. claim 15 24. Radiation detection apparatus, comprising: an array of radiation-sensitive detectors, which generate electrical signals responsive to radiation photons incident thereon; and processing circuitry comprising: a plurality of signal processing channels, each channel coupled to process the signals from a respective one of the detectors and comprising a counter, which counts the number of photons incident on the respective detector, wherein each of the channels comprises an integrated circuit disposed on a common substrate with the respective detector; and a bus common to a multiplicity of the channels, which receives and outputs respective photon counts from the channels in turn. 25. Apparatus according to claim 24 , wherein the signal processing channels comprise discriminators, which reject signals corresponding to photons outside a predetermined energy range, so that the counters count only photons within the predetermined energy range. claim 24 26. Apparatus according to claim 25 , wherein the processing circuitry comprises a threshold control circuit, which adjusts the predetermined energy range of the discriminators. claim 25 27. Apparatus according to claim 24 , wherein the array of detectors comprises an array of radiation-sensitive diodes. claim 24 28. Apparatus according to claim 27 , wherein the diodes comprise silicon diode detectors. claim 27 29. Apparatus according to claim 24 , wherein the common substrate comprises a semiconductor chip including integrated circuits belonging to a multiplicity of the signal processing channels. claim 24 |
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056129866 | abstract | Methods for forming X-ray images having 0.25 .mu.m minimum line widths on X-ray sensitive material are presented. A holgraphic image of a desired circuit pattern is projected onto a wafer or other image-receiving substrate to allow recording of the desired image in photoresist material. In one embodiment, the method uses on-axis transmission and provides a high flux X-ray source having modest monochromaticity and coherence requirements. A layer of light-sensitive photoresist material on a wafer with a selected surface is provided to receive the image(s). The hologram has variable optical thickness and variable associated optical phase angle and amplitude attenuation for transmission of the X-rays. A second embodiment uses off-axis holography. The wafer receives the holographic image by grazing incidence reflection from a hologram printed on a flat metal or other highly reflecting surface or substrate. In this second embodiment, an X-ray beam with a high degree of monochromaticity and spatial coherence is required. |
abstract | A spacer grid can be applied to close-spaced nuclear fuel rods. The spacer grid is directed to solve the problem in which, as the outer diameter of each nuclear fuel rod increases due to the use of dual-cooled nuclear fuel rods for improving cooling performance and obtaining high combustion and high output power, the gap between the neighboring nuclear fuel rods is narrowed to thus make it impossible to use an existing spacer grid. The spacer grid is a combination of unit grid straps, each of which has supports for supporting each of the nuclear fuel rods set in a narrow array and has a sheet shape, which are combined with each other. The supports are located at positions shifted from the longitudinal central line of each unit grid strap toward sub-channels. |
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description | A CR/FS grapple according to an embodiment of the present invention will be explained in detail with reference to the accompanying drawings hereinafter. The CR/FS grapple according to one embodiment of the resent invention is employed such that, when a periodical inspection of the BWR, etc. is conducted, the CR/FS grapple is hung down into a reactor pressure vessel 1 (see FIG. 8) so as to release control rods 4 and a fuel support 6 from a core and, in turn, install the control rods 4 and the fuel support 6 from a fuel storage pool outside the reactor pressure vessel into the core. FIGS. 1 and 2 are vertical sectional views showing a schematic configuration of the CR/FS grapple according to the embodiment of the present invention. FIG. 1 is a front vertical sectional view while FIG. 2 is a side vertical sectional view. As shown in FIGS. 1 and 2, the CR/FS grapple according to the embodiment of the present invention comprises a main body frame 24 to which a separating frame 18 is fitted attachably/detachably by fitting bolts 31. A guide member 23 is provided at a lower end of the separating frame 18 to guide a hoist handle 4d (see FIG. 8) of the control rod 4. A square cylinder member 15 is provided in an inside of the separating frame 18 to be moved vertically by a predetermined width, but an upward movement of the square cylinder member 15 can be limited by a stopper 32 which is projected from an inner surface of the separating frame 18. A stud 14 is secured to an top end of the square cylinder member 15. The stud 14 is formed to have a square cylinder shape because a grid guide (not shown) which can guide the CR/FS grapple in passing through the upper grid 5 (see FIG. 8) can be attached to the stud 14. The stud 14 is connected to a hoist rope, etc. which is wound on an auxiliary hoist of a refueling machine. The CR/FS grapple can be picked up by the hoist rope, etc. and then inserted into the inside of the reactor pressure vessel 1 (see FIG. 8). A control rod holding air cylinder 17 is pivotally coupled to the square cylinder member 15 via a hinged coupling. A hook 16 which is formed of a hook-shaped member is pivotally coupled to a piston rod 17a of the air cylinder 17 via the hinged coupling. A bearing member 39 having a longitudinal hole 38 therein is provided on the separating frame 18. A pivot axis 40 is inserted into the longitudinal hole 38 and is movable vertically. Then, the hook 16 is fitted to the bearing member 39 via the pivot axis 40 such that such hook 16 can be pivoted and moved vertically. A limit switch 51 is provided in the close vicinity of the air cylinder 17. An operation state of the air cylinder 17 can be detected by the limit switch 51. The limit switch 51 is electrically connected to an indicator lamp 29 which is energized by a battery 28. The result detected by the limit switch 51 can be displayed by the indicator lamp 29. Therefore, the operator can check by eye whether or not the hook 16 is located in a hooking position. A control rod holding means 50 consists of the square cylinder member 15, the hook 16, and the air cylinder 17. Since the square cylinder member 15 can be moved up and down relative to the separating frame 18 by a predetermined width, the control rod holding means 50 can also be moved up and down relative to the separating frame 18. At that time, because the separating frame 18 is fixed to a main body frame 24, the control rod holding means 50 can be moved vertically relative to the main body frame 24 by a predetermined width, i.e., can be displaced relatively along a longitudinal direction of the control rod 4. A hoisting stroke of the control rod holding means 50 with respect to the separating frame 18 is designed to exceed a height of a projected portion of the hook 16 such that a mechanical lock mechanism can operate while keeping an engagement of the projection portion formed at the top end of the hook 16 with the hoist handle 4d (see FIG. 8) of the control rod 4. As shown in FIG. 2, a coupling releasing air cylinder 22 is fixed to the separating frame 18. A piston rod 22a of the coupling releasing air cylinder 22 is connected to a coupling releasing link mechanism 20 via an actuating rod 21. In this manner, the coupling releasing link mechanism 20 can be driven by the coupling releasing air cylinder 22. Also, an arm 19 is provided to extend along the actuating rod 21. The arm 19 is also provided such that it can be moved vertically relative to the separating frame 18. As shown in FIG. 3, during the exchange operation of the control rod 4, the arm 19 is inserted into a clearance between the control rod 4 and the cruciform through hole 4a of the fuel support 6 from the upper side. A coupling releasing means 60, which can uncouple the control rod 4 and the control rod drive mechanism 8 by virtue of a spud coupling, is composed of the coupling releasing link mechanism 20, the actuating rod 21, the coupling releasing air cylinder 22. As shown in FIG. 2, a damper mechanism 30 is provided to a top end of the air cylinder 22 to apply resistance to a piston rod 22a of the coupling releasing air cylinder 22 in its operation. The damper mechanism 30 constitutes an operational timing control mechanism which can uncouple the control rod 4 and control rod drive mechanism 8 by using the coupling releasing means 60 after the control rod holding means 50 has held the hoist handle 4d (see FIG. 8). A damper drive axis 30a of the damper mechanism 30 is connected to the piston rod 22a of the air cylinder 22. FIG. 4A, FIG. 4B, and FIG. 4C show an operation of the coupling releasing link mechanism 20 in the CR/FS grapple. As shown in FIG. 4A, FIG. 4B, and FIG. 4C, the coupling releasing link mechanism 20 includes a first arm 20a, a second arm 20b, and a third arm 20c. A top end of the first arm 20a is pivotally coupled to a bottom end of the actuating rod 21, a bottom end of the first arm 20a is pivotally coupled to one end of the second arm 20b, and the other end of the second arm 20b is pivotally coupled to a middle portion of the third arm 20c. In addition, one end of the third arm 20c is pivotally coupled to the arm 19. As shown in FIG. 4A, FIG. 4B, and FIG. 4C in sequence, when the piston rod 22a is lifted upward by driving the air cylinder 22, the first arm 20a and the second arm 20b can be pulled up and also the third arm 20c can be pulled up until a horizontal position while being pivoted. Further, a bend-shaped switching member 20d is swingably and pivotally coupled to the third arm 20c. A limit switch 33 is provided over the switching member 20d.The limit switch 33 is electrically connected to an indicator lamp 29 shown in FIG. 1. On/off states of the limit switch 33 can be displayed by the indicator lamp 29. As shown in FIG. 1 and FIG. 2, a pair of fuel support holding plungers 25 are provided on a bottom surface of the main body frame 24. A pair of fuel support holding air cylinders 26 for driving these holding plungers 25 are provided on the fuel support holding plungers 25. Limit switches 27 are provided in the neighborhood of the air cylinders 26 respectively. Operation states of the air cylinders 26 can be detected by the limit switches 27 and then detected results can be displayed by the indicator lamp 29. Then, a fuel support holding means 70 which can hold the fuel support 6 (see FIG. 11) is composed of the fuel support holding plungers 25 and the fuel support holding air cylinders 26. The fuel support holding plungers 25 have contact pieces 34 respectively. As shown in FIG. 5, the contact pieces 34 can be moved back and forth by a fuel support holding link mechanism 35. A pair of contact pieces 34 are arranged at positions corresponding to a pair of opposing orifices 6e (f, g, h) of the fuel support 6 shown in FIG. 11. Stepped portions 34a on which the upper portions of the orifices 6e (f, g, h) are placed are formed on the contact pieces 34 respectively. The fuel support holding link mechanism 35 has a first arm 35a and a second arm 35b. One end of the first arm 35a is pivotally coupled to a rear end of the contact piece 34, and the other end of the first end 35a is pivotally coupled to one end of the second arm 35b, and the other end of the second arm 35b is pivotally coupled to outer peripheral wall 25a (see FIG. 1) of the fuel support holding plunger 25. In addition, a bottom end of the actuating rod 36 which is coupled to the piston rod of the air cylinder 26 is pivotally coupled to a pivotable portion between the first arm 35a and the second arm 35b. FIG. 6 is a schematic system diagram showing a piping system for supplying a working air to a control rod holding air cylinder 17 and a coupling releasing air cylinder 22 in the CR/FS grapple. As can be seen from FIG. 6, the control rod holding air cylinder 17 and the coupling releasing air cylinder 22 employ commonly a set of low and high pressure working air sources. In general, the working air sources which can be employed in the nuclear power plant consist of two sets of low and high pressure working air sources. Hence, a set of working air sources are commonly used for the control rod holding air cylinder 17 and the coupling releasing air cylinder 22, while a set of remaining working air sources can be used to operate the fuel support holding air cylinders 26. Therefore, there is no case where the site has lack of the working air sources. FIG. 7 shows a modification in which a flow restrict mechanism 37 is provided in place of the damper mechanism 30 as an operational timing control mechanism in the CR/FS grapple. In this modification, a flow restrict mechanism 37 is provided in the middle of a low pressure side working fluid pipe 41 connected to the coupling releasing air cylinder 22. Next, referring to FIGS. 1 and 8, when the control rods 4 and the fuel support 6 are picked up from the inside of the water-filled reactor pressure vessel 1, operations performed during the periodical inspection of the BWR by using the CR/FS grapple according to the present embodiment will be explained. At the time when the lifting operation of the control rods 4 and the fuel support 6 is to be carried out, the fuel assemblies 3 fitted in predetermined grids have already been taken out from the inside of the reactor pressure vessel 1 by the refueling machine, etc. and the control rods 4 have been descended to their full pull-out states. At first, a hoist rope wound on an auxiliary hoist, etc. of the refueling machine (not shown) is connected to the stud 14 of the CR/FS grapple, and then the CR/FS grapple is hung down inside of the reactor pressure vessel 1 to be inserted into the preselected grid. Thus, the separating frame 18 and the main body frame 24 are seated on the control rod 4 and the fuel support 6 respectively. Then, a compressed air is supplied from an air system (working air source) of the refueling machine to the control rod holding air cylinder 17, the coupling releasing air cylinder 22, and the fuel support holding air cylinders 26 respectively. At that time, since the operational timing control mechanism composed of the damper mechanism 30 or the flow restrict mechanism 37 is provided to the coupling releasing air cylinder 22, movement of the coupling releasing air cylinder 22 is delayed relative to that of the control rod holding air cylinder 17 in operation. Therefore, at first the hook 16 of the control rod holding means 50 holds the hoist handle 4d of the control rod 4 and then the third arm 20c of the coupling releasing link mechanism 20 of the coupling releasing means 60 pulls up the release handle 4e of the control rod 4, whereby the coupled state of the control rod 4 and the control rod drive mechanism 8 by using the spud coupling can be released. Like the above, since the operational timing control mechanism is provided, a holding operation of the control rod 4 can be effected by the control rod holding means 50 before the coupled state is released by the coupling releasing means 60. Since the time difference in the operations of the control rod holding air cylinder 17 and the coupling releasing air cylinder 22 is caused by the operational timing control mechanism composed of the damper mechanism 30 or the flow restrict mechanism 37, a time difference can be generated in their respective operations even though the common working air source is employed for the control rod holding air cylinder 17 and the coupling releasing air cylinder 22. In addition, since the common working air source is employed for both air cylinders 17, 22, a time difference can be caused without fail in their operation even when a pressure of the supplied air from the working air source is varied. The operation when the hoist handle 4d of the control rod 4 is held by the control rod holding means 50 will be explained hereunder. After the separating frame 18 is seated on the control rod 4, the piston rod 17a is drawn in by driving the control rod holding air cylinder 17 to pivot the hook 16 such that the hook 16 is shifted to the hooked position shown in FIG. 1. At that time, an operation state of the control rod holding air cylinder 17 can be detected by the limit switch 51 and then the detection result can be displayed on the indicator lamp 29. Accordingly, the operator can check by eye whether or not the hook 16 is located in its hooking position. Next, when the stud 14 is lifted by winding the hoist rope of the auxiliary hoist in this state, only the control rod holding means 50 which consists of the hook 16, the control rod holding air cylinder 17, and the square cylinder member 15 can be lifted up together with the stud 14 while the separating frame 18 is still seated on the control rod 4. Then, the hook 16 grasps the hoist handle 4d of the control rod 4 and lifts it. At that time, it can be detected by sensing a weight applied to the hoist rope whether or not the hook 16 of the control rod holding means 50 has held the hoist handle 4d of the control rod 4. In this way, prior to lifting-up of an entire CR/FS grapple, it can be checked or confirmed whether the control rod 4 is held by the control rod holding means 50, or not. Also, under the condition that the hoist handle 4d is grasped and then lifted up by the hook 16, a hooked state by the hook 4 can be held by a self-weight of the control rod 4 since the hook 16 is formed like a hook-shape. As a result, the control rod 4 is prevented from dropping down and also safety can be maintained even if either supply of the compressed air to the control rod holding air cylinder 17 has been lost or operations have been performed wrong. Subsequently, an operation performed when the coupled state of the control rod 4 and the control rod drive mechanism 8 by the spud coupling is released by the coupling releasing means 60 will be explained hereunder. First of all, when the piston rod 22a is lifted up by driving the coupling releasing air cylinder 22, the first arm 20a and the second arm 20b can be pulled up and also the third arm 20c can be pulled up to its horizontal position while it is being pivoted, as shown in FIG. 4A, FIG. 4B, and FIG. 4C in order. Next, the overall coupling releasing link mechanism 20 as well as the arm 19 can be lifted up by rising up the piston rod 22a further from the state shown in FIG. 4C. Then, the release handle 4e of the control rod 4 can be pulled up by the third arm 20c, so that the coupling of the control rod 4 and the control rod drive mechanism 8 can be released. In addition, the switching member 20d can be pivoted when the third arm 20c is engaging with the release handle 4e and thus the limit switch 33 can be pushed up by one end of the switching member 20d. Then, the limit switch 33 is operated to switch its on/off state and as a result the displaying state of the indicator lamp 29 is changed, whereby the operator can check or confirm by eye that the release handle 4e of the control rod 4 has been actuated. Operations effected when the fuel support 6 is held by the fuel support holding means 70 are explained hereinbelow. Prior to starting the holding operation of the fuel support 6, the contact pieces 34 and the fuel support holding link mechanism 35 are positioned at locations shown by chain double-dashed lines in FIG. 5. The actuating rod 36 is moved upward from this location by driving the fuel support holding air cylinders 26. Thus, the contact pieces 34 are caused to advance to the orifices 6e (f, g, h) of the fuel support 6 and thus advance further more than the position indicated by solid lines in FIG. 5. Forward movements of the contact pieces 34 are continued until the first arm 35a and the second arm 35b are positioned from their downward-convex alignment to their linear alignment, and then the contact pieces 34 are switched to their backward movements when the first arm 35a and the second arm 35b are shifted from their linear alignment to their upward-convex alignment. Then, at the time point when the piston rods of the fuel support holding air cylinders 26 and the actuating rod 36 reach their upper limit positions of lifting strokes, the backward movements of the contact pieces 34 are stopped and therefore the contact pieces 34 are positioned, as shown by the solid lines in FIG. 5. At this time, operation states of the fuel support holding air cylinders 26 can be detected by the limit switches 27 and the detected results can then be displayed by the indicator lamps 29 respectively. Accordingly, the operator can check or confirm by eye whether or not the contact pieces 34 are in their held positions. Then, in the situation indicated by the solid lines in FIG. 5, when the CR/FS grapple is pulled up by winding up the hoist rope of the auxiliary hoist, the upper portions of the orifices 6e (f, g, h) of the fuel support 6 are put on the stepped portions 34a of the contact pieces 34 so that the CR/FS grapple as well as the fuel support 6 can be pulled up together. Since both forward and backward movements of the contact pieces 34 can be prevented in the situation that the upper portions of the orifices 6e (f, g, h) of the fuel support 6 are put on the stepped portions 34a of the contact pieces 34, the fuel support 6 can be prevented from dropping down to thus maintain safety even if either supply of the compressed air to the control rod holding air cylinder 17 has been lost at the worst or wrong operations have been effected. As described above, the CR/FS grapple can be lifted upward after the control rods 4 have been held by the control rod holding means 50 and also the fuel support 6 has been held by the fuel support holding means 70, so that the control rods 4 and the fuel support 6 can be lifted up simultaneously and carried out together from the reactor pressure vessel 1. The CR/FS grapple according to the embodiment of the present invention may be employed when the control rods 4 and the fuel support 6 are carried into the inside of the reactor pressure vessel 1 and then installed therein. In this case, the control rods 4 and the fuel support 6 maybe lifted up and installed simultaneously. In the CR/FS grapple according to the embodiment of the present invention, since the separating frame 18 may be detached from the main body frame 24 by releasing the fitting bolts 31 (see FIG. 2), the fuel support holding means 70 and an assembly consisting of the control rod holding means 50 and the coupling releasing means 60 can be employed independently respectively as separate bodies. As a consequence, in a case that the control rods 4 cannot be pulled out until their full pull-out states due to a failure of the control rod drive mechanism 8, at first only the fuel support 6 can be lifted up by using the fuel support holding means 70 to be picked out from the reactor pressure vessel 1, and then the control rods 4 are hoisted by the control rod holding means 50 and the coupling releasing means 60 to be taken out from the reactor pressure vessel 1. As described above, according to the CR/FS grapple of the embodiment of the present invention, since both the control rods 4 and the fuel support 6 can be held by the control rod holding means 50 and the fuel support holding means 70 and also the coupled state of the control rods 4 and the control rod drive mechanism 8 by using the spud coupling can be released by the coupling releasing means 60, there is no necessity of executing the coupling releasing operation from the pedestal side of the bottom of the reactor, unlike the aforesaid related art. Therefore, an efficiency of the exchange operation of the control rods 4 can be improved by reducing the term of the periodical inspection and the exposure of the operator. More particularly, in contrast to the prior operation, it is possible to reduce the operation time by about thirty minutes per control rod 4. The typical operation time of the prior operation was about 55 minutes per control rod 4. Therefore, about 55% of the operation time can be reduced by using the present CR/FS grapple. |
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abstract | A rotatable contamination barrier for use with an EUV radiation system is disclosed. The contamination barrier has a blade structure configured to trap contaminant material coming from a radiation source, a bearing structure, coupled to a static frame, configured to rotatably bear the blade structure, and an eccentric mass element displaced relative to a central axis of rotation to balance the blade structure in the bearing structure. |
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abstract | An x-ray tube for generating a sweeping x-ray beam. A cathode is disposed within a vacuum enclosure and emits a beam of electrons attracted toward an anode. The anode is adapted for rotation with respect to the vacuum enclosure about an axis of rotation. At least one collimator opening corotates with the anode within the vacuum enclosure, such that a swept x-ray beam is emitted. |
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060375973 | claims | 1. A laminate device for self-contained delivery of epithermal neutrons, comprising: a first polymer layer having embedded therein a chemical source of alpha-radiation; a second polymer layer, the second polymer layer being opaque to alpha-radiation and having defined therein a diffraction opening for passing alpha-radiation from said chemical source of alpha-radiation; a third polymer layer having embedded therein a source of epithermal neutrons, said third polymer layer being positioned relative to said second polymer layer such that alpha-radiation from said chemical source of alpha-radiation impinges upon the source of neutrons, thereby generating epithermal neutrons; and a fourth polymer layer, said fourth polymer layer being impermeable to epithermal neutrons and defining a diffraction opening for passing said generated epithermal neutrons. a non-conductive substrate opaque to epithermal neutrons; a pattern of electrically conductive material imprinted upon said non-conductive substrate, said pattern providing a plurality of leads for electrical connection of radiation detectors; a plurality of radiation detectors affixed to said substrate and electrically connected to said plurality of leads; and a plurality of self-contained radiation sources affixed to said substrate. positioning said object adjacent a source/detector array device for non-destructive interrogation, the device including: a non-conductive substrate; a pattern of electrically conductive material imprinted upon said non-conductive substrate, said pattern providing a plurality of leads for electrical connection of radiation detectors; a plurality of radiation detectors affixed to said substrate and electrically connected to said plurality of leads; and a plurality of self-contained radiation sources affixed to said substrate; detecting said characteristic radiation signal with said detectors. a first polymer layer having embedded therein a chemical source of alpha-radiation; a second polymer layer, the second polymer layer being opaque to alpha-radiation and having defined therein a diffraction opening for passing alpha-radiation from said chemical source of alpha-radiation; a third polymer layer having embedded therein a chemical source of epithermal neutrons, said third polymer layer being positioned relative to said second polymer layer such that alpha-radiation from said source of alpha-radiation impinges upon the source of neutrons, thereby generating epithermal neutrons; and a fourth polymer layer, said fourth polymer layer being opaque to epithermal neutrons and defining a diffraction opening for passing said generated epithermal neutrons. 2. The device of claim 1, which also includes a fifth polymer layer having embedded therein a sealed article containing liquid nitrogen, for cooling the device. 3. A radiation source/detector array device for non-destructive interrogation, comprising: 4. The device of claim 3 wherein said plurality of self-contained radiation sources includes devices according to claim 1. 5. The device of claim 4 which also includes a plurality of x-ray sources, and a plurality of x-ray detectors. 6. The device of claim 5, which also includes a plurality of neutron detectors. 7. A method for non-destructively interrogating an object, comprising: 8. The method of claim 7 wherein said radiation sources include thermal neutron sources. 9. The method of claim 8 wherein the epithermal neutron sources include: |
052876764 | abstract | A device for handling liquid radioactive waste includes a heater for heating and drying liquid radioactive waste being poured into a container. A pallet, which is preferably formed of metal, receives the container. A ground vehicle transports the pallet. A supplementary heater is part of the pallet. |
claims | 1. A scattered radiation shielding grid comprising a radiation absorbing material representing a pattern corresponding to a combined motif of a plurality of tiled prototiles, each prototile comprising a width W(p), a length and a motif solely within the prototile, wherein the prototile width W(p)=W/(Ixc2x10.05I) and W(p)xe2x89xa0W+D, where W is a radiation sensitive area width of a radiation sensor of a radiation detection panel comprising a plurality of equal size radiation sensors separated by interstitial spaces having a width D, over which said grid is positioned, and I is an integer. 2. The scattered radiation grid according to claim 1 wherein W(p)=W/I. claim 1 3. A scattered radiation shielding grid comprising a radiation absorbing material, and a radiation detection panel over which said grid is positioned comprising a plurality of equal size radiation sensors having a radiation sensitive area width W, separated by radiation insensitive interstitial spaces having a width D, and wherein said grid absorbing material forms a pattern representing a combined motif of a tiled plurality of substantially identical prototiles, each prototile comprising: (a) a width W(p)=W/I, wherein I is an integer; (b) a length; and (c) a motif contained solely within the prototile. 4. The scattered radiation grid and detection panel according to claim 3 further comprising a gain correction circuit associated with said detection panel and wherein W(p)=W/(Ixc2x10.05I) and W(p)xe2x89xa0W+D. claim 3 5. The scattered radiation grid and detection panel according to claim 4 further comprising a radiation source and said grid is positioned between said panel and said radiation source at a fixed, known distance from said panel, wherein said prototile width W(p) is a projected prototile width on said panel. claim 4 6. A method for designing a pattern for absorption material for a scattered radiation shielding grid for a radiation detection panel comprising an array of a plurality of sensors each having a radiation sensitive area having a width W and a length, the sensors arrayed so that each radiation sensitive area is separated by each adjacent radiation sensitive area by an interstitial space having a width D, the method comprising: a) determining a sensor width corresponding to the width of the radiation sensitive area of the sensor b) creating a prototile having a width W(p)=W/I wherein I is an integer; c) producing within said prototile a motif and d) tiling a plurality of said prototiles to produce the pattern, said pattern consisting of the combined motif of the tiled prototiles. 7. A method for manufacturing a scattered radiation shielding grid comprising a pattern of radiation absorbing material for a radiation detection panel comprising an array of a plurality of sensors each having a radiation sensitive area having a width W and a length, the sensors arrayed so that each radiation sensitive area is separated by each adjacent radiation sensitive area by an interstitial space having a width D, the method comprising: a) determining a sensor width W corresponding to the width of the radiation sensitive area of the sensor b) creating a prototile having a width W(p)=W/(Ixc2x10.05I), W(p)xe2x89xa0W+D and wherein I is an integer; c) producing within said prototile a motif; d) tiling a plurality of said prototiles to produce a pattern consisting of the combined motif of the tiled prototiles; e) forming said radiation absorbing material in said grid in the shape of said combined motif. 8. The method according to claim 7 wherein in step (b) the prototile width: W(p)=W/I. claim 7 9. A method for generating a radiogram with an exposure system comprising radiation source, and a radiation detection panel, wherein said radiation detection panel comprises an array of a plurality of sensors each having a radiation sensitive area having a width W and a length, the sensors arrayed so that each radiation sensitive area is separated by each adjacent radiation sensitive area by an interstitial space having a width D, the method comprising: positioning between said radiation source and said panel a grid comprising a radiation absorbing material formed in a pattern representing a combined motif of a plurality of substantially identical tiled prototiles, each prototile comprising a width W(p), a length and said motif, said motif contained solely within the prototile, wherein the prototile width W(p)=W/I where I is an integer. 10. The method of producing a radiogram according to claim 9 wherein said system further comprises a gain correction circuit, said prototile width W(p)=W/(Ixc2x10.05I), W(p)xe2x89xa0W+D and wherein after positioning the grid between said source and said panel there is performed a calibration step comprising exposing the panel to radiation through said grid and adjusting said gain correction circuit to produce a uniform output from all sensors in said panel. claim 9 |
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description | The disclosure herein relates to X-ray detectors, particularly relates to semiconductor X-ray detectors. X-ray detectors may be devices used to measure the flux, spatial distribution, spectrum or other properties of X-rays. X-ray detectors may be used for many applications. One important application is imaging. X-ray imaging is a radiography technique and can be used to reveal the internal structure of a non-uniformly composed and opaque object such as the human body. Early X-ray detectors for imaging include photographic plates and photographic films. A photographic plate may be a glass plate with a coating of light-sensitive emulsion. Although photographic plates were replaced by photographic films, they may still be used in special situations due to the superior quality they offer and their extreme stability. A photographic film may be a plastic film (e.g., a strip or sheet) with a coating of light-sensitive emulsion. In the 1980s, photostimulable phosphor plates (PSP plates) became available. A PSP plate may contain a phosphor material with color centers in its lattice. When the PSP plate is exposed to X-ray, electrons excited by X-ray are trapped in the color centers until they are stimulated by a laser beam scanning over the plate surface. As the plate is scanned by laser, trapped excited electrons give off light, which is collected by a photomultiplier tube. The collected light is converted into a digital image. In contrast to photographic plates and photographic films, PSP plates can be reused. Another kind of X-ray detectors are X-ray image intensifiers. Components of an X-ray image intensifier are usually sealed in a vacuum. In contrast to photographic plates, photographic films, and PSP plates, X-ray image intensifiers may produce real-time images, i.e., do not require post-exposure processing to produce images. X-ray first hits an input phosphor (e.g., cesium iodide) and is converted to visible light. The visible light then hits a photocathode (e.g., a thin metal layer containing cesium and antimony compounds) and causes emission of electrons. The number of emitted electrons is proportional to the intensity of the incident X-ray. The emitted electrons are projected, through electron optics, onto an output phosphor and cause the output phosphor to produce a visible-light image. Scintillators operate somewhat similarly to X-ray image intensifiers in that scintillators (e.g., sodium iodide) absorb X-ray and emit visible light, which can then be detected by a suitable image sensor for visible light. In scintillators, the visible light spreads and scatters in all directions and thus reduces spatial resolution. Reducing the scintillator thickness helps to improve the spatial resolution but also reduces absorption of X-ray. A scintillator thus has to strike a compromise between absorption efficiency and resolution. Semiconductor X-ray detectors largely overcome this problem by direct conversion of X-ray into electric signals. A semiconductor X-ray detector may include a semiconductor layer that absorbs X-ray in wavelengths of interest. When an X-ray photon is absorbed in the semiconductor layer, multiple charge carriers (e.g., electrons and holes) are generated and swept under an electric field towards electrical contacts on the semiconductor layer. Cumbersome heat management required in currently available semiconductor X-ray detectors (e.g., Medipix) can make a detector with a large area and a large number of pixels difficult or impossible to produce. Disclosed herein is an apparatus suitable for detecting x-ray, comprising: an X-ray absorption layer comprising an electrode; a first voltage comparator configured to compare a voltage of the electrode to a first threshold; a second voltage comparator configured to compare the voltage to a second threshold; a counter configured to register a number of X-ray photons absorbed by the X-ray absorption layer; a controller; wherein the controller is configured to start a time delay from a time at which the first voltage comparator determines that an absolute value of the voltage equals or exceeds an absolute value of the first threshold; wherein the controller is configured to activate the second voltage comparator during (including the beginning and the expiration) the time delay; wherein the controller is configured to cause the number registered by the counter to increase by one, if the second voltage comparator determines that an absolute value of the voltage equals or exceeds an absolute value of the second threshold. The first voltage comparator and the second voltage comparator may be the same component. When a voltage comparator determines whether an absolute value of a voltage equals or exceeds an absolute value of a threshold, the voltage comparator does not necessarily compare the absolute values. Instead, when the voltage and the threshold are both negative, the voltage comparator may compare the actual values of the voltage and the threshold; when the voltage is equally or more negative than the threshold, the absolute value of voltage equals or exceeds the absolute value of the threshold. According to an embodiment, the apparatus further comprises a capacitor module electrically connected to the electrode, wherein the capacitor module is configured to collect charge carriers from the electrode. According to an embodiment, the controller is configured to activate the second voltage comparator at a beginning or expiration of the time delay. According to an embodiment, the controller is configured to deactivate the first voltage comparator at the beginning of, or during the time delay. According to an embodiment, the controller is configured to deactivate the second voltage comparator at the expiration of the time delay or at the time when the second voltage comparator determines that the absolute value of the voltage equals or exceeds the absolute value of the second threshold. According to an embodiment, the apparatus further comprises a voltmeter and the controller is configured to cause the voltmeter to measure the voltage upon expiration of the time delay. According to an embodiment, the controller is configured to determine an X-ray photon energy based on a value of the voltage measured upon expiration of the time delay. According to an embodiment, the controller is configured to connect the electrode to an electrical ground. The electrical ground may be a virtual ground. A virtual ground (also known as a “virtual earth”) is a node of a circuit that is maintained at a steady reference potential, without being connected directly to the reference potential. According to an embodiment, a rate of change of the voltage is substantially zero at expiration of the time delay. According to an embodiment, a rate of change of the voltage is substantially non-zero at expiration of the time delay. According to an embodiment, the X-ray absorption layer comprises a diode. According to an embodiment, the X-ray absorption layer comprises silicon, germanium, GaAs, CdTe, CdZnTe, or a combination thereof. According to an embodiment, the apparatus does not comprise a scintillator. According to an embodiment, the apparatus comprises an array of pixels. Disclosed herein is a system comprising the apparatus described above and an X-ray source, wherein the system is configured to perform X-ray radiography on human chest or abdomen. According to an embodiment, the system comprises the apparatus described above and an X-ray source, wherein the system is configured to perform X-ray radiography on human mouth. Disclosed herein is a cargo scanning or non-intrusive inspection (NII) system, comprising the apparatus described above and an X-ray source, wherein the cargo scanning or non-intrusive inspection (NII) system is configured to form an image using backscattered X-ray. Disclosed herein is a cargo scanning or non-intrusive inspection (NII) system, comprising the apparatus described above and an X-ray source, wherein the cargo scanning or non-intrusive inspection (NII) system is configured to form an image using X-ray transmitted through an object inspected. Disclosed herein is a full-body scanner system comprising the apparatus described above and an X-ray source. Disclosed herein is an X-ray computed tomography (X-ray CT) system comprising the apparatus described above and an X-ray source. Disclosed herein is an electron microscope comprising the apparatus described above, an electron source and an electronic optical system. Disclosed herein is a system comprising the apparatus described above, wherein the system is an X-ray telescope, or an X-ray microscopy, or wherein the system is configured to perform mammography, industrial defect detection, microradiography, casting inspection, weld inspection, or digital subtraction angiography. Disclosed herein is a method comprising: starting a time delay from a time at which an absolute value of a voltage of an electrode of an X-ray absorption layer equals or exceeds an absolute value of a first threshold; activating a second circuit during (including the beginning and expiration of) the time delay; if an absolute value of the voltage equals or exceeds an absolute value of a second threshold, increasing a count of X-ray photon incident on the X-ray absorption layer by one. According to an embodiment, the method further comprises connecting the electrode to an electrical ground. According to an embodiment, the method further comprises measuring the voltage upon expiration of the time delay. According to an embodiment, the method further comprises determining an X-ray photon energy based on a value of the voltage at expiration of the time delay. According to an embodiment, a rate of change of the voltage is substantially zero at expiration of the time delay. According to an embodiment, a rate of change of the voltage is substantially non-zero at expiration of the time delay. According to an embodiment, activating the second circuit is at a beginning or expiration of the time delay. According to an embodiment, the second circuit is configured to compare the absolute value of the voltage to the absolute value of the second threshold. According to an embodiment, the method further comprises deactivating a first circuit at a beginning the time delay. According to an embodiment, the first circuit is configured to compare the absolute value of the voltage to the absolute value of the first threshold. The first circuit and the second circuit may be the same circuit. Disclosed herein is a system suitable for phase-contrast X-ray imaging (PCI), the system comprising: the apparatus described above, a second X-ray detector, a spacer, wherein the apparatus and the second X-ray detector are spaced apart by the spacer. According to an embodiment, the apparatus and the second X-ray detector are configured to respectively capture an image of an object simultaneously. According to an embodiment, the second X-ray detector is identical to the apparatus. Disclosed herein is a system suitable for phase-contrast X-ray imaging (PCI), the system comprising: the apparatus described above, wherein the apparatus is configured to move to and capture images of an object exposed to incident X-ray at different distances from the object. FIG. 1A schematically shows a semiconductor X-ray detector 100, according to an embodiment. The semiconductor X-ray detector 100 may include an X-ray absorption layer 110 and an electronics layer 120 (e.g., an ASIC) for processing or analyzing electrical signals incident X-ray generates in the X-ray absorption layer 110. In an embodiment, the semiconductor X-ray detector 100 does not comprise a scintillator. The X-ray absorption layer 110 may include a semiconductor material such as, silicon, germanium, GaAs, CdTe, CdZnTe, or a combination thereof. The semiconductor may have a high mass attenuation coefficient for the X-ray energy of interest. The X-ray absorption layer 110 may include one or more diodes (e.g., p-i-n or p-n) formed by a first doped region 111, one or more discrete regions 114 of a second doped region 113. The second doped region 113 may be separated from the first doped region 111 by an optional the intrinsic region 112. The discrete portions 114 are separated from one another by the first doped region 111 or the intrinsic region 112. The first doped region 111 and the second doped region 113 have opposite types of doping (e.g., region 111 is p-type and region 113 is n-type, or region 111 is n-type and region 113 is p-type). In the example in FIG. 1A, each of the discrete regions 114 of the second doped region 113 forms a diode with the first doped region 111 and the optional intrinsic region 112. Namely, in the example in FIG. 1A, the X-ray absorption layer 110 has a plurality of diodes having the first doped region 111 as a shared electrode. The first doped region 111 may also have discrete portions. FIG. 1B shows a semiconductor X-ray detector 100, according to an embodiment. The semiconductor X-ray detector 100 may include an X-ray absorption layer 110 and an electronics layer 120 (e.g., an ASIC) for processing or analyzing electrical signals incident X-ray generates in the X-ray absorption layer 110. In an embodiment, the semiconductor X-ray detector 100 does not comprise a scintillator. The X-ray absorption layer 110 may include a semiconductor material such as, silicon, germanium, GaAs, CdTe, CdZnTe, or a combination thereof. The semiconductor may have a high mass attenuation coefficient for the X-ray energy of interest. The X-ray absorption layer 110 may not include a diode but includes a resistor. When an X-ray photon hits the X-ray absorption layer 110 including diodes, it may be absorbed and generate one or more charge carriers by a number of mechanisms. An X-ray photon may generate 10 to 100000 charge carriers. The charge carriers may drift to the electrodes of one of the diodes under an electric field. The field may be an external electric field. The electrical contact 119B may include discrete portions each of which is in electrical contact with the discrete regions 114. In an embodiment, the charge carriers may drift in directions such that the charge carriers generated by a single X-ray photon are not substantially shared by two different discrete regions 114 (“not substantially shared” here means less than 5%, less than 2% or less than 1% of these charge carriers flow to a different one of the discrete regions 114 than the rest of the charge carriers). In an embodiment, the charge carriers generated by a single X-ray photon can be shared by two different discrete regions 114. FIG. 2 shows an exemplary top view of a portion of the device 100 with a 4-by-4 array of discrete regions 114. Charge carriers generated by an X-ray photon incident around the footprint of one of these discrete regions 114 are not substantially shared with another of these discrete regions 114. The area around a discrete region 114 in which substantially all (more than 95%, more than 98% or more than 99% of) charge carriers generated by an X-ray photon incident therein flow to the discrete region 114 is called a pixel associated with that discrete region 114. Namely, less than 5%, less than 2% or less than 1% of these charge carriers flow beyond the pixel. By measuring the drift current flowing into each of the discrete regions 114, or the rate of change of the voltage of each of the discrete regions 114, the number of X-ray photons absorbed (which relates to the incident X-ray intensity) and/or the energies thereof in the pixels associated with the discrete regions 114 may be determined. Thus, the spatial distribution (e.g., an image) of incident X-ray intensity may be determined by individually measuring the drift current into each one of an array of discrete regions 114 or measuring the rate of change of the voltage of each one of an array of discrete regions 114. The pixels may be organized in any suitable array, such as, a square array, a triangular array and a honeycomb array. The pixels may have any suitable shape, such as, circular, triangular, square, rectangular, and hexangular. The pixels may be individually addressable. When an X-ray photon hits the X-ray absorption layer 110 including a resistor but not diodes, it may be absorbed and generate one or more charge carriers by a number of mechanisms. An X-ray photon may generate 10 to 100000 charge carriers. The charge carriers may drift to the electrical contacts 119A and 119B under an electric field. The field may be an external electric field. The electrical contact 119B includes discrete portions. In an embodiment, the charge carriers may drift in directions such that the charge carriers generated by a single X-ray photon are not substantially shared by two different discrete portions of the electrical contact 119B (“not substantially shared” here means less than 5%, less than 2% or less than 1% of these charge carriers flow to a different one of the discrete portions than the rest of the charge carriers). In an embodiment, the charge carriers generated by a single X-ray photon can be shared by two different discrete portions of the electrical contact 119B. Charge carriers generated by an X-ray photon incident around the footprint of one of these discrete portions of the electrical contact 119B are not substantially shared with another of these discrete portions of the electrical contact 119B. The area around a discrete portion of the electrical contact 119B in which substantially all (more than 95%, more than 98% or more than 99% of) charge carriers generated by an X-ray photon incident therein flow to the discrete portion of the electrical contact 119B is called a pixel associated with the discrete portion of the electrical contact 119B. Namely, less than 5%, less than 2% or less than 1% of these charge carriers flow beyond the pixel associated with the one discrete portion of the electrical contact 119B. By measuring the drift current flowing into each of the discrete portion of the electrical contact 119B, or the rate of change of the voltage of each of the discrete portions of the electrical contact 119B, the number of X-ray photons absorbed (which relates to the incident X-ray intensity) and/or the energies thereof in the pixels associated with the discrete portions of the electrical contact 119B may be determined. Thus, the spatial distribution (e.g., an image) of incident X-ray intensity may be determined by individually measuring the drift current into each one of an array of discrete portions of the electrical contact 119B or measuring the rate of change of the voltage of each one of an array of discrete portions of the electrical contact 119B. The pixels may be organized in any suitable array, such as, a square array, a triangular array and a honeycomb array. The pixels may have any suitable shape, such as, circular, triangular, square, rectangular, and hexangular. The pixels may be individually addressable. The electronics layer 120 may include an electronic system 121 suitable for processing or interpreting signals generated by X-ray photons incident on the X-ray absorption layer 110. The electronic system 121 may include an analog circuitry such as a filter network, amplifiers, integrators, and comparators, or a digital circuitry such as a microprocessors, and memory. The electronic system 121 may include components shared by the pixels or components dedicated to a single pixel. For example, the electronic system 121 may include an amplifier dedicated to each pixel and a microprocessor shared among all the pixels. The electronic system 121 may be electrically connected to the pixels by vias 131. Space among the vias may be filled with a filler material 130, which may increase the mechanical stability of the connection of the electronics layer 120 to the X-ray absorption layer 110. Other bonding techniques are possible to connect the electronic system 121 to the pixels without using vias. FIG. 3A and FIG. 3B each show a component diagram of the electronic system 121, according to an embodiment. The electronic system 121 may include a first voltage comparator 301, a second voltage comparator 302, a counter 320, a switch 305, a voltmeter 306 and a controller 310. The first voltage comparator 301 is configured to compare the voltage of an electrode of a diode 300 to a first threshold. The diode may be a diode formed by the first doped region 111, one of the discrete regions 114 of the second doped region 113, and the optional intrinsic region 112. Alternatively, the first voltage comparator 301 is configured to compare the voltage of an electrical contact (e.g., a discrete portion of electrical contact 119B) to a first threshold. The first voltage comparator 301 may be configured to monitor the voltage directly, or calculate the voltage by integrating an electric current flowing through the diode or electrical contact over a period of time. The first voltage comparator 301 may be controllably activated or deactivated by the controller 310. The first voltage comparator 301 may be a continuous comparator. Namely, the first voltage comparator 301 may be configured to be activated continuously, and monitor the voltage continuously. The first voltage comparator 301 configured as a continuous comparator reduces the chance that the system 121 misses signals generated by an incident X-ray photon. The first voltage comparator 301 configured as a continuous comparator is especially suitable when the incident X-ray intensity is relatively high. The first voltage comparator 301 may be a clocked comparator, which has the benefit of lower power consumption. The first voltage comparator 301 configured as a clocked comparator may cause the system 121 to miss signals generated by some incident X-ray photons. When the incident X-ray intensity is low, the chance of missing an incident X-ray photon is low because the time interval between two successive photons is relatively long. Therefore, the first voltage comparator 301 configured as a clocked comparator is especially suitable when the incident X-ray intensity is relatively low. The first threshold may be 5-10%, 10%-20%, 20-30%, 30-40% or 40-50% of the maximum voltage one incident X-ray photon may generate in the diode or the resistor. The maximum voltage may depend on the energy of the incident X-ray photon (i.e., the wavelength of the incident X-ray), the material of the X-ray absorption layer 110, and other factors. For example, the first threshold may be 50 mV, 100 mV, 150 mV, or 200 mV. The second voltage comparator 302 is configured to compare the voltage to a second threshold. The second voltage comparator 302 may be configured to monitor the voltage directly, or calculate the voltage by integrating an electric current flowing through the diode or the electrical contact over a period of time. The second voltage comparator 302 may be a continuous comparator. The second voltage comparator 302 may be controllably activate or deactivated by the controller 310. When the second voltage comparator 302 is deactivated, the power consumption of the second voltage comparator 302 may be less than 1%, less than 5%, less than 10% or less than 20% of the power consumption when the second voltage comparator 302 is activated. The absolute value of the second threshold is greater than the absolute value of the first threshold. As used herein, the term “absolute value” or “modulus” |x| of a real number x is the non-negative value of x without regard to its sign. Namely, x = { x , if x ≥ 0 - x , if x ≤ 0 . The second threshold may be 200%-300% of the first threshold. The second threshold may be at least 50% of the maximum voltage one incident X-ray photon may generate in the diode or resistor. For example, the second threshold may be 100 mV, 150 mV, 200 mV, 250 mV or 300 mV. The second voltage comparator 302 and the first voltage comparator 310 may be the same component. Namely, the system 121 may have one voltage comparator that can compare a voltage with two different thresholds at different times. The first voltage comparator 301 or the second voltage comparator 302 may include one or more op-amps or any other suitable circuitry. The first voltage comparator 301 or the second voltage comparator 302 may have a high speed to allow the system 121 to operate under a high flux of incident X-ray. However, having a high speed is often at the cost of power consumption. The counter 320 is configured to register a number of X-ray photons reaching the diode or resistor. The counter 320 may be a software component (e.g., a number stored in a computer memory) or a hardware component (e.g., a 4017 IC and a 7490 IC). The controller 310 may be a hardware component such as a microcontroller and a microprocessor. The controller 310 is configured to start a time delay from a time at which the first voltage comparator 301 determines that the absolute value of the voltage equals or exceeds the absolute value of the first threshold (e.g., the absolute value of the voltage increases from below the absolute value of the first threshold to a value equal to or above the absolute value of the first threshold). The absolute value is used here because the voltage may be negative or positive, depending on whether the voltage of the cathode or the anode of the diode or which electrical contact is used. The controller 310 may be configured to keep deactivated the second voltage comparator 302, the counter 320 and any other circuits the operation of the first voltage comparator 301 does not require, before the time at which the first voltage comparator 301 determines that the absolute value of the voltage equals or exceeds the absolute value of the first threshold. The time delay may expire before or after the voltage becomes stable, i.e., the rate of change of the voltage is substantially zero. The phase “the rate of change of the voltage is substantially zero” means that temporal change of the voltage is less than 0.1%/ns. The phase “the rate of change of the voltage is substantially non-zero” means that temporal change of the voltage is at least 0.1%/ns. The controller 310 may be configured to activate the second voltage comparator during (including the beginning and the expiration) the time delay. In an embodiment, the controller 310 is configured to activate the second voltage comparator at the beginning of the time delay. The term “activate” means causing the component to enter an operational state (e.g., by sending a signal such as a voltage pulse or a logic level, by providing power, etc.). The term “deactivate” means causing the component to enter a non-operational state (e.g., by sending a signal such as a voltage pulse or a logic level, by cut off power, etc.). The operational state may have higher power consumption (e.g., 10 times higher, 100 times higher, 1000 times higher) than the non-operational state. The controller 310 itself may be deactivated until the output of the first voltage comparator 301 activates the controller 310 when the absolute value of the voltage equals or exceeds the absolute value of the first threshold. The controller 310 may be configured to cause the number registered by the counter 320 to increase by one, if, during the time delay, the second voltage comparator 302 determines that the absolute value of the voltage equals or exceeds the absolute value of the second threshold. The controller 310 may be configured to cause the voltmeter 306 to measure the voltage upon expiration of the time delay. The controller 310 may be configured to connect the electrode to an electrical ground, so as to reset the voltage and discharge any charge carriers accumulated on the electrode. In an embodiment, the electrode is connected to an electrical ground after the expiration of the time delay. In an embodiment, the electrode is connected to an electrical ground for a finite reset time period. The controller 310 may connect the electrode to the electrical ground by controlling the switch 305. The switch may be a transistor such as a field-effect transistor (FET). In an embodiment, the system 121 has no analog filter network (e.g., a RC network). In an embodiment, the system 121 has no analog circuitry. The voltmeter 306 may feed the voltage it measures to the controller 310 as an analog or digital signal. The system 121 may include a capacitor module 309 electrically connected to the electrode of the diode 300 or the electrical contact, wherein the capacitor module is configured to collect charge carriers from the electrode. The capacitor module can include a capacitor in the feedback path of an amplifier. The amplifier configured as such is called a capacitive transimpedance amplifier (CTIA). CTIA has high dynamic range by keeping the amplifier from saturating and improves the signal-to-noise ratio by limiting the bandwidth in the signal path. Charge carriers from the electrode accumulate on the capacitor over a period of time (“integration period”) (e.g., as shown in FIG. 4, between t0 to t1, or t1-t2). After the integration period has expired, the capacitor voltage is sampled and then reset by a reset switch. The capacitor module can include a capacitor directly connected to the electrode. FIG. 4 schematically shows a temporal change of the electric current flowing through the electrode (upper curve) caused by charge carriers generated by an X-ray photon incident on the diode or the resistor, and a corresponding temporal change of the voltage of the electrode (lower curve). The voltage may be an integral of the electric current with respect to time. At time to, the X-ray photon hits the diode or the resistor, charge carriers start being generated in the diode or the resistor, electric current starts to flow through the electrode of the diode or the resistor, and the absolute value of the voltage of the electrode or electrical contact starts to increase. At time t1, the first voltage comparator 301 determines that the absolute value of the voltage equals or exceeds the absolute value of the first threshold V1, and the controller 310 starts the time delay TD1 and the controller 310 may deactivate the first voltage comparator 301 at the beginning of TD1. If the controller 310 is deactivated before t1, the controller 310 is activated at t1. During TD1, the controller 310 activates the second voltage comparator 302. The term “during” a time delay as used here means the beginning and the expiration (i.e., the end) and any time in between. For example, the controller 310 may activate the second voltage comparator 302 at the expiration of TD1. If during TD1, the second voltage comparator 302 determines that the absolute value of the voltage equals or exceeds the absolute value of the second threshold at time t2, the controller 310 causes the number registered by the counter 320 to increase by one. At time te, all charge carriers generated by the X-ray photon drift out of the X-ray absorption layer 110. At time ts, the time delay TD1 expires. In the example of FIG. 4, time ts is after time te; namely TD1 expires after all charge carriers generated by the X-ray photon drift out of the X-ray absorption layer 110. The rate of change of the voltage is thus substantially zero at ts. The controller 310 may be configured to deactivate the second voltage comparator 302 at expiration of TD1 or at t2, or any time in between. The controller 310 may be configured to cause the voltmeter 306 to measure the voltage upon expiration of the time delay TD1. In an embodiment, the controller 310 causes the voltmeter 306 to measure the voltage after the rate of change of the voltage becomes substantially zero after the expiration of the time delay TD1. The voltage at this moment is proportional to the amount of charge carriers generated by an X-ray photon, which relates to the energy of the X-ray photon. The controller 310 may be configured to determine the energy of the X-ray photon based on voltage the voltmeter 306 measures. One way to determine the energy is by binning the voltage. The counter 320 may have a sub-counter for each bin. When the controller 310 determines that the energy of the X-ray photon falls in a bin, the controller 310 may cause the number registered in the sub-counter for that bin to increase by one. Therefore, the system 121 may be able to detect an X-ray image and may be able to resolve X-ray photon energies of each X-ray photon. After TD1 expires, the controller 310 connects the electrode to an electric ground for a reset period RST to allow charge carriers accumulated on the electrode to flow to the ground and reset the voltage. After RST, the system 121 is ready to detect another incident X-ray photon. Implicitly, the rate of incident X-ray photons the system 121 can handle in the example of FIG. 4 is limited by 1/(TD1+RST). If the first voltage comparator 301 has been deactivated, the controller 310 can activate it at any time before RST expires. If the controller 310 has been deactivated, it may be activated before RST expires. FIG. 5 schematically shows a temporal change of the electric current flowing through the electrode (upper curve) caused by noise (e.g., dark current, background radiation, scattered X-rays, fluorescent X-rays, shared charges from adjacent pixels), and a corresponding temporal change of the voltage of the electrode (lower curve), in the system 121 operating in the way shown in FIG. 4. At time to, the noise begins. If the noise is not large enough to cause the absolute value of the voltage to exceed the absolute value of V1, the controller 310 does not activate the second voltage comparator 302. If the noise is large enough to cause the absolute value of the voltage to exceed the absolute value of V1 at time t1 as determined by the first voltage comparator 301, the controller 310 starts the time delay TD1 and the controller 310 may deactivate the first voltage comparator 301 at the beginning of TD1. During TD1 (e.g., at expiration of TD1), the controller 310 activates the second voltage comparator 302. The noise is very unlikely large enough to cause the absolute value of the voltage to exceed the absolute value of V2 during TD1. Therefore, the controller 310 does not cause the number registered by the counter 320 to increase. At time te, the noise ends. At time ts, the time delay TD1 expires. The controller 310 may be configured to deactivate the second voltage comparator 302 at expiration of TD1. The controller 310 may be configured not to cause the voltmeter 306 to measure the voltage if the absolute value of the voltage does not exceed the absolute value of V2 during TD1. After TD1 expires, the controller 310 connects the electrode to an electric ground for a reset period RST to allow charge carriers accumulated on the electrode as a result of the noise to flow to the ground and reset the voltage. Therefore, the system 121 may be very effective in noise rejection. FIG. 6 schematically shows a temporal change of the electric current flowing through the electrode (upper curve) caused by charge carriers generated by an X-ray photon incident on the diode or the resistor, and a corresponding temporal change of the voltage of the electrode (lower curve), when the system 121 operates to detect incident X-ray photons at a rate higher than 1/(TD1+RST). The voltage may be an integral of the electric current with respect to time. At time to, the X-ray photon hits the diode or the resistor, charge carriers start being generated in the diode or the resistor, electric current starts to flow through the electrode of the diode or the electrical contact of resistor, and the absolute value of the voltage of the electrode or the electrical contact starts to increase. At time t1, the first voltage comparator 301 determines that the absolute value of the voltage equals or exceeds the absolute value of the first threshold V1, and the controller 310 starts a time delay TD2 shorter than TD1, and the controller 310 may deactivate the first voltage comparator 301 at the beginning of TD2. If the controller 310 is deactivated before t1, the controller 310 is activated at t1. During TD2 (e.g., at expiration of TD2), the controller 310 activates the second voltage comparator 302. If during TD2, the second voltage comparator 302 determines that the absolute value of the voltage equals or exceeds the absolute value of the second threshold at time t2, the controller 310 causes the number registered by the counter 320 to increase by one. At time te, all charge carriers generated by the X-ray photon drift out of the X-ray absorption layer 110. At time th, the time delay TD2 expires. In the example of FIG. 6, time th is before time te; namely TD2 expires before all charge carriers generated by the X-ray photon drift out of the X-ray absorption layer 110. The rate of change of the voltage is thus substantially non-zero at th. The controller 310 may be configured to deactivate the second voltage comparator 302 at expiration of TD2 or at t2, or any time in between. The controller 310 may be configured to extrapolate the voltage at te from the voltage as a function of time during TD2 and use the extrapolated voltage to determine the energy of the X-ray photon. After TD2 expires, the controller 310 connects the electrode to an electric ground for a reset period RST to allow charge carriers accumulated on the electrode to flow to the ground and reset the voltage. In an embodiment, RST expires before te. The rate of change of the voltage after RST may be substantially non-zero because all charge carriers generated by the X-ray photon have not drifted out of the X-ray absorption layer 110 upon expiration of RST before te. The rate of change of the voltage becomes substantially zero after te and the voltage stabilized to a residue voltage VR after te. In an embodiment, RST expires at or after te, and the rate of change of the voltage after RST may be substantially zero because all charge carriers generated by the X-ray photon drift out of the X-ray absorption layer 110 at te. After RST, the system 121 is ready to detect another incident X-ray photon. If the first voltage comparator 301 has been deactivated, the controller 310 can activate it at any time before RST expires. If the controller 310 has been deactivated, it may be activated before RST expires. FIG. 7 schematically shows a temporal change of the electric current flowing through the electrode (upper curve) caused by noise (e.g., dark current, background radiation, scattered X-rays, fluorescent X-rays, shared charges from adjacent pixels), and a corresponding temporal change of the voltage of the electrode (lower curve), in the system 121 operating in the way shown in FIG. 6. At time to, the noise begins. If the noise is not large enough to cause the absolute value of the voltage to exceed the absolute value of V1, the controller 310 does not activate the second voltage comparator 302. If the noise is large enough to cause the absolute value of the voltage to exceed the absolute value of V1 at time t1 as determined by the first voltage comparator 301, the controller 310 starts the time delay TD2 and the controller 310 may deactivate the first voltage comparator 301 at the beginning of TD2. During TD2 (e.g., at expiration of TD2), the controller 310 activates the second voltage comparator 302. The noise is very unlikely large enough to cause the absolute value of the voltage to exceed the absolute value of V2 during TD2. Therefore, the controller 310 does not cause the number registered by the counter 320 to increase. At time te, the noise ends. At time th, the time delay TD2 expires. The controller 310 may be configured to deactivate the second voltage comparator 302 at expiration of TD2. After TD2 expires, the controller 310 connects the electrode to an electric ground for a reset period RST to allow charge carriers accumulated on the electrode as a result of the noise to flow to the ground and reset the voltage. Therefore, the system 121 may be very effective in noise rejection. FIG. 8 schematically shows a temporal change of the electric current flowing through the electrode (upper curve) caused by charge carriers generated by a series of X-ray photons incident on the diode or the resistor, and a corresponding temporal change of the voltage of the electrode (lower curve), in the system 121 operating in the way shown in FIG. 6 with RST expires before te. The voltage curve caused by charge carriers generated by each incident X-ray photon is offset by the residue voltage before that photon. The absolute value of the residue voltage successively increases with each incident photon. When the absolute value of the residue voltage exceeds V1 (see the dotted rectangle in FIG. 8), the controller starts the time delay TD2 and the controller 310 may deactivate the first voltage comparator 301 at the beginning of TD2. If no other X-ray photon incidence on the diode or the resistor during TD2, the controller connects the electrode to the electrical ground during the reset time period RST at the end of TD2, thereby resetting the residue voltage. The residue voltage thus does not cause an increase of the number registered by the counter 320. FIG. 9A shows a flow chart for a method suitable for detecting X-ray using a system such as the system 121 operating as shown in FIG. 4. In step 901, compare, e.g., using the first voltage comparator 301, a voltage of an electrode of a diode or an electrical contact of a resistor exposed to X-ray, to the first threshold. In step 902, determine, e.g., with the controller 310, whether the absolute value of the voltage equals or exceeds the absolute value of the first threshold V1. If the absolute value of the voltage does not equal or exceed the absolute value of the first threshold, the method goes back to step 901. If the absolute value of the voltage equals or exceeds the absolute value of the first threshold, continue to step 903. In step 903, start, e.g., using the controller 310, the time delay TD1. In step 904, activate, e.g., using the controller 310, a circuit (e.g., the second voltage comparator 302 or the counter 320) during the time delay TD1 (e.g., at the expiration of TD1). In step 905, compare, e.g., using the second voltage comparator 302, the voltage to the second threshold. In step 906, determine, e.g., using the controller 310, whether the absolute value of the voltage equals or exceeds the absolute value of the second threshold V2. If the absolute value of the voltage does not equal or exceed the absolute value of the second threshold, the method goes to step 910. If the absolute value of the voltage equals or exceeds the absolute value of the second threshold, continue to step 907. In step 907, cause, e.g., using the controller 310, the number registered in the counter 320 to increase by one. In optional step 908, measure, e.g., using the voltmeter 306, the voltage upon expiration of the time delay TD1. In optional step 909, determine, e.g., using the controller 310, the X-ray photon energy based the voltage measured in step 908. There may be a counter for each of the energy bins. After measuring the X-ray photon energy, the counter for the bin to which the photon energy belongs can be increased by one. The method goes to step 910 after step 909. In step 910, reset the voltage to an electrical ground, e.g., by connecting the electrode of the diode or an electrical contact of a resistor to an electrical ground. Steps 908 and 909 may be omitted, for example, when neighboring pixels share a large portion (e.g., >30%) of charge carriers generated from a single photon. FIG. 9B shows a flow chart for a method suitable for detecting X-ray using the system such as the system 121 operating as shown in FIG. 6. In step 1001, compare, e.g., using the first voltage comparator 301, a voltage of an electrode of a diode or an electrical contact of a resistor exposed to X-ray, to the first threshold. In step 1002, determine, e.g., with the controller 310, whether the absolute value of the voltage equals or exceeds the absolute value of the first threshold V1. If the absolute value of the voltage does not equal or exceed the absolute value of the first threshold, the method goes back to step 1001. If the absolute value of the voltage equals or exceeds the absolute value of the first threshold, continue to step 1003. In step 1003, start, e.g., using the controller 310, the time delay TD2. In step 1004, activate, e.g., using the controller 310, a circuit (e.g., the second voltage comparator 302 or the counter 320) during the time delay TD2 (e.g., at the expiration of TD2). In step 1005, compare, e.g., using the second voltage comparator 302, the voltage to the second threshold. In step 1006, determine, e.g., using the controller 310, whether the absolute value of the voltage equals or exceeds the absolute value of the second threshold V2. If the absolute value of the voltage does not equal or exceed the absolute value of the second threshold, the method goes to step 1010. If the absolute value of the voltage equals or exceeds the absolute value of the second threshold, continue to step 1007. In step 1007, cause, e.g., using the controller 310, the number registered in the counter 320 to increase by one. The method goes to step 1010 after step 1007. In step 1010, reset the voltage to an electrical ground, e.g., by connecting the electrode of the diode or an electrical contact of a resistor to an electrical ground. The semiconductor X-ray detector 100 may be used for phase-contrast X-ray imaging (PCI) (also known as phase-sensitive X-ray imaging). PCI encompasses techniques that form an image of an object at least partially using the phase shift (including the spatial distribution of the phase shift) of an X-ray beam caused by that object. One way to obtain the phase shift is transforming the phase into variations in intensity. PCI can be combined with tomographic techniques to obtain the 3D-distribution of the real part of the refractive index of the object. PCI is more sensitive to density variations in the object than conventional intensity-based X-ray imaging (e.g., radiography). PCI is especially useful for imaging soft tissues. According to an embodiment, FIG. 10 schematically shows a system 1900 suitable for PCI. The system 1900 may include at least two X-ray detectors 1910 and 1920. One or both of the two X-ray detectors 1910 is the semiconductor X-ray detector 100 described herein. The X-ray detectors 1910 and 1920 may be spaced apart by a spacer 1930. The spacer 1930 may have very little absorption of the X-ray. For example, the spacer 1930 may have a very small mass attenuation coefficient (e.g., <10 cm2g−1, <1 cm2g−1, <0.1 cm2g−1, or <0.01 cm2g−1,). The mass attenuation coefficient of the spacer 1930 may be uniform (e.g., variation between every two points in the spacer 1930 less than 5%, less than 1% or less than 0.1%). The spacer 1930 may cause the same amount of changes to the phase of X-ray passing through the spacer 1930. For example, the spacer 1930 may be a gas (e.g., air), a vacuum chamber, may comprise aluminum, beryllium, silicon, or a combination thereof. The system 1900 can be used to obtain the phase shift of incident X-ray 1950 caused by an object 1960 being imaged. The X-ray detectors 1910 and 1920 can capture two images (i.e., intensity distributions) simultaneously. Because of the X-ray detectors 1910 and 1920 are separated by the spacer 1930, the two images are different distances from the object 1960. The phase may be determined from the two images, for example, using algorithms based on the linearization of the Fresnel diffraction integral. According to an embodiment, FIG. 11 schematically shows a system 1800 suitable for PCI. The system 1800 comprises the semiconductor X-ray detector 100 described herein. The semiconductor X-ray detector 100 is configured to move to and capture images of an object 1860 exposed to incident X-ray 1850 at different distances from the object 1860. The images may not necessarily be captured simultaneously. The phase may be determined from the images, for example, using algorithms based on the linearization of the Fresnel diffraction integral. FIG. 12 schematically shows a system comprising the semiconductor X-ray detector 100 described herein. The system may be used for medical imaging such as chest X-ray radiography, abdominal X-ray radiography, etc. The system comprises an X-ray source 1201. X-ray emitted from the X-ray source 1201 penetrates an object 1202 (e.g., a human body part such as chest, limb, abdomen), is attenuated by different degrees by the internal structures of the object 1202 (e.g., bones, muscle, fat and organs, etc.), and is projected to the semiconductor X-ray detector 100. The semiconductor X-ray detector 100 forms an image by detecting the intensity distribution of the X-ray. FIG. 13 schematically shows a system comprising the semiconductor X-ray detector 100 described herein. The system may be used for medical imaging such as dental X-ray radiography. The system comprises an X-ray source 1301. X-ray emitted from the X-ray source 1301 penetrates an object 1302 that is part of a mammal (e.g., human) mouth. The object 1302 may include a maxilla bone, a palate bone, a tooth, the mandible, or the tongue. The X-ray is attenuated by different degrees by the different structures of the object 1302 and is projected to the semiconductor X-ray detector 100. The semiconductor X-ray detector 100 forms an image by detecting the intensity distribution of the X-ray. Teeth absorb X-ray more than dental caries, infections, periodontal ligament. The dosage of X-ray radiation received by a dental patient is typically small (around 0.150 mSv for a full mouth series). FIG. 14 schematically shows a cargo scanning or non-intrusive inspection (NII) system comprising the semiconductor X-ray detector 100 described herein. The system may be used for inspecting and identifying goods in transportation systems such as shipping containers, vehicles, ships, luggage, etc. The system comprises an X-ray source 1401. X-ray emitted from the X-ray source 1401 may backscatter from an object 1402 (e.g., shipping containers, vehicles, ships, etc.) and be projected to the semiconductor X-ray detector 100. Different internal structures of the object 1402 may backscatter X-ray differently. The semiconductor X-ray detector 100 forms an image by detecting the intensity distribution of the backscattered X-ray and/or energies of the backscattered X-ray photons. FIG. 15 schematically shows another cargo scanning or non-intrusive inspection (NII) system comprising the semiconductor X-ray detector 100 described herein. The system may be used for luggage screening at public transportation stations and airports. The system comprises an X-ray source 1501. X-ray emitted from the X-ray source 1501 may penetrate a piece of luggage 1502, be differently attenuated by the contents of the luggage, and projected to the semiconductor X-ray detector 100. The semiconductor X-ray detector 100 forms an image by detecting the intensity distribution of the transmitted X-ray. The system may reveal contents of luggage and identify items forbidden on public transportation, such as firearms, narcotics, edged weapons, flammables. FIG. 16 schematically shows a full-body scanner system comprising the semiconductor X-ray detector 100 described herein. The full-body scanner system may detect objects on a person's body for security screening purposes, without physically removing clothes or making physical contact. The full-body scanner system may be able to detect non-metal objects. The full-body scanner system comprises an X-ray source 1601. X-ray emitted from the X-ray source 1601 may backscatter from a human 1602 being screened and objects thereon, and be projected to the semiconductor X-ray detector 100. The objects and the human body may backscatter X-ray differently. The semiconductor X-ray detector 100 forms an image by detecting the intensity distribution of the backscattered X-ray. The semiconductor X-ray detector 100 and the X-ray source 1601 may be configured to scan the human in a linear or rotational direction. FIG. 17 schematically shows an X-ray computed tomography (X-ray CT) system. The X-ray CT system uses computer-processed X-rays to produce tomographic images (virtual “slices”) of specific areas of a scanned object. The tomographic images may be used for diagnostic and therapeutic purposes in various medical disciplines, or for flaw detection, failure analysis, metrology, assembly analysis and reverse engineering. The X-ray CT system comprises the semiconductor X-ray detector 100 described herein and an X-ray source 1701. The semiconductor X-ray detector 100 and the X-ray source 1701 may be configured to rotate synchronously along one or more circular or spiral paths. FIG. 18 schematically shows an electron microscope. The electron microscope comprises an electron source 1801 (also called an electron gun) that is configured to emit electrons. The electron source 1801 may have various emission mechanisms such as thermionic, photocathode, cold emission, or plasmas source. The emitted electrons pass through an electronic optical system 1803, which may be configured to shape, accelerate, or focus the electrons. The electrons then reach a sample 1802 and an image detector may form an image therefrom. The electron microscope may comprise the semiconductor X-ray detector 100 described herein, for performing energy-dispersive X-ray spectroscopy (EDS). EDS is an analytical technique used for the elemental analysis or chemical characterization of a sample. When the electrons incident on a sample, they cause emission of characteristic X-rays from the sample. The incident electrons may excite an electron in an inner shell of an atom in the sample, ejecting it from the shell while creating an electron hole where the electron was. An electron from an outer, higher-energy shell then fills the hole, and the difference in energy between the higher-energy shell and the lower energy shell may be released in the form of an X-ray. The number and energy of the X-rays emitted from the sample can be measured by the semiconductor X-ray detector 100. The semiconductor X-ray detector 100 described here may have other applications such as in an X-ray telescope, X-ray mammography, industrial X-ray defect detection, X-ray microscopy or microradiography, X-ray casting inspection, X-ray non-destructive testing, X-ray weld inspection, X-ray digital subtraction angiography, etc. It may be suitable to use this semiconductor X-ray detector 100 in place of a photographic plate, a photographic film, a PSP plate, an X-ray image intensifier, a scintillator, or another semiconductor X-ray detector. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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description | Referring now to the drawings, particularly to FIG. 1, there is illustrated a core spray nozzle, generally designated 10, for a nuclear reactor, the vessel wall of which is shown at 12. The core spray nozzle 10 of FIG. 1 illustrates the nozzle after replacement of the thermal sleeve portion previously employed to convey cooling water from the core spray piping external of the vessel through the nozzle safe end 14 and the vessel wall 12 to the core spray piping within the vessel wall. More particularly, the T-box 54 and thermal sleeve 56 as described below with reference to FIG. 2, are replaced by a T-box/thermal sleeve assembly 16 as illustrated in FIG. 1 from within the vessel wall. The replacement assembly is mechanically connected with the nozzle safe end 14 within the nozzle bore 15. As illustrated, the vessel wall 12 includes as part of the core spray nozzle 10 a laterally projecting pipe 18 to which the safe end 14 is welded. The safe end 14 includes a cylindrical pipe or thermal sleeve remnant 20 of the previously extant thermal sleeve in the nozzle 10. Particularly, the T-box/thermal sleeve includes a T-junction or T-connection 22 which, in final assembly, lies adjacent the interior of the vessel wall 12. The T-box includes a pair of lateral passages 24 for connection with piping which extends about the interior of the vessel wall approximately 90xc2x0 in opposite directions from the T-box. The piping, of course, connects the cooling water supplied through the core spray nozzle 10 to interior piping of the core spray system. The T-box 22 necks down to a smaller diameter replacement thermal sleeve 26 terminating at an outer end within the nozzle bore in a counterbore 28. The thermal sleeve 26 includes stepped diameter axially spaced threaded portions 30 and 32 along its interior wall surfaces. The thermal sleeve 26 is sized to receive the end portion of the thermal sleeve remnant 20 in the couterbore 28. A seal 34, preferably a Belleville washer, is disposed between the opposed ends of the replacement thermal sleeve 26 and the thermal sleeve remnant 20 such that the Belleville washer 34 is compressed, rendering portions of the thermal sleeve remnant 20 and the replacement thermal sleeve 26 in compression. To retain the thermal sleeve 26 mechanically coupled to the thermal sleeve remnant 20, a groove 36 (FIGS. 3 and 4) is provided along the interior wall surface of the thermal sleeve remnant 20 spaced back from its end. A collet 38 comprising a cylindrical sleeve having a plurality of fingers 40 adjacent one end and an externally threaded portion 42 adjacent an opposite end is threaded along threads 30 within the replacement thermal sleeve 26. It will be appreciated that with radially directed flanges 44 on the ends of the fingers 40 engaged in groove 36, the T-box/thermal sleeve assembly 16 is maintained mechanically assembled to the thermal sleeve remnant 20. To retain the fingers 40 with the flanges 44 engaged in the groove 36, a retention sleeve 46 externally threaded at one end is received within the collet 38. It will be appreciated by threading the retention sleeve 46 along the interior threads 32 of the replacement thermal sleeve 26, the opposite end of the retention sleeve 46 overlies the fingers 40, preventing the finger flanges 44 from removal within the groove 36. A cover 50 is also secured to the open end of the T-box 22 and a seal washer, preferably a Belleville spring washer 52, is disposed between the cover and the T-box end. The collet 38 is preferably formed of Inconel. Inconel has a linear coefficient of expansion less than that of steel. The thermal sleeve remnant 20 and replacement thermal sleeve 26 are formed of steel. Accordingly, when the system heats up, the collet will not expand as much as the thermal sleeve remnant 20 and the replacement thermal sleeve 26. The Belleville washer 34 is also formed of Inconel. Therefore, at temperature, the Belleville washer 34 is placed under further compression desirably enhancing its sealing capability and providing thermal compliance at temperature by accommodating differences in thermal expansion between component materials. Referring now to FIGS. 2-5, the manner of replacing the extant T-box/thermal sleeve with the T-box/thermal sleeve assembly 16 will now be described. The existing T-box 54 and a portion of the thermal sleeve 56 are first removed as illustrated in FIGS. 2 and 3 from the core spray nozzle by employing conventional underwater electric discharge machining, leaving a remnant end 58 (FIG. 3) of the thermal sleeve. This is accomplished underwater and from within the reactor vessel. The newly cut end portion 58 of the thermal sleeve remnant is then prepared for mating with the replacement T-box/thermal sleeve forging. First, the end of the thermal sleeve remnant 20 is machined flat and perpendicular to the thermal sleeve bore. Additionally, the annular groove 36 is formed by machining along the inside diameter of the thermal sleeve remnant 20 at a specified distance from its end portion 58. As evident, the groove 36 provides the mating connection with the finger flanges 44 of the collet to form the mechanical joint between the thermal sleeve remnant 20 and the replacement T-box/thermal sleeve assembly 16 and which joint is sufficient to withstand axial loading therebetween. Next, the collet 38 is screwthreaded onto the replacement T-box/thermal sleeve assembly. It will be appreciated that the threading action between the collet 38 and the thermal sleeve 26 along threads 30 provides a length adjustability between the collet fingers 40 and the seal interface of the T-box/thermal sleeve end and the thermal sleeve remnant end portion 58. A seal washer, preferably a Belleville spring seal 60, is provided on the end of the replacement thermal sleeve 26 and the entire assembly is inserted from the interior of the vessel wall 12 into the nozzle bore 15 such that the counterbore 28 of the replacement thermal sleeve 26 receives the thermal sleeve remnant end portion 58. The Belleville spring washer 60 lies between the opposing end faces of the thermal sleeve remnant 20 and replacement thermal sleeve 26. Also, as the replacement thermal sleeve 26 is inserted to receive the end portion 58 of the thermal sleeve remnant 20, the flanges 44 of the fingers 40 engage in the groove 36 to mechanically retain the parts in assembly. The Belleville spring washer 60 provides compression loading on portions of each of the thermal sleeve remnant 20 and replacement thermal sleeve 26. To retain the replacement thermal sleeve 26 and thermal sleeve remnant 20 mechanically engaged one with the other, the retention sleeve 46 is inserted through the open end of the T-box and threaded along the threads 32 of the replacement thermal sleeve 26. It will be appreciated that the opposite end of the retention sleeve advances and radially overlies the internal surfaces of the fingers 40 whereby the flanges 44 of the fingers are maintained in the groove 36. The ends of the fingers 40 are tapered to facilitate their axial insertion through the end of the thermal sleeve remnant 20. Thus, the fingers 40 are initially biased radially inwardly for passage along the interior surface of the thermal sleeve remnant 20 and are then resiliently flexed radially outwardly to engage the flanges 44 in the groove 36. The retention sleeve is rotated into position by using a spanner wrench or other suitable tool and is then locked against rotation, for example, by staking to the replacement thermal sleeve 26, to prevent inadvertent rotation and possible removal. The T-box/thermal sleeve replacement assembly 16 is supported within the nozzle bore by three or four equally spaced adjustable wedge blocks 62. Each wedge block 62 comprises a keeper 64, a wedge 66 and a jack screw 68. By suitably adjusting the jack screws driving the wedges, the inserted assembly can be centered vis-a-vis the thermal sleeve remnant 20. Thus, the wedge blocks maintain axial alignment of the assembly with the thermal sleeve and minimize bending stresses of the collet fingers. Also, by using the wedge blocks 62, field measuring or machining is not necessary to ensure proper fit-up in the core spray nozzle 10. Finally, the end cap 50 is threaded onto the open end of the T-box. Another Belleville seal ring 52 is preferably disposed between the cap and the T-box. An anti-rotation feature may be incorporated between the closure end cap 50 and the T-box 22 to prevent loosening of the closure cap 50. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims. |
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description | The present application is a continuation-in-part of U.S. patent application Ser. No. 16/902,387, filed Jun. 16, 2020, which is a continuation of U.S. patent application Ser. No. 15/689,571, filed Aug. 29, 2017, now U.S. Pat. No. 10,692,617, which is a continuation of U.S. patent application Ser. No. 14/239,752, filed Mar. 21, 2014, now U.S. Pat. No. 9,748,009, which is a national stage entry under 35 U.S.C. § 371 of PCT/US2012/051634, filed Aug. 20, 2012, which claims priority to U.S. Provisional Patent Application Ser. No. 61/525,583, filed Aug. 19, 2011. The present application is also a continuation-in-part of U.S. patent application Ser. No. 16/729,654, filed Dec. 30, 2019, which is a divisional of U.S. patent application Ser. No. 15/596,444, filed May 16, 2017, now U.S. Pat. No. 10,535,440, which is a divisional of U.S. patent application Ser. No. 13/925,585, filed Jun. 24, 2013, now U.S. Pat. No. 9,685,248, which claims priority to U.S. Provisional Patent Application Ser. No. 61/663,316, filed Jun. 22, 2012. The present application is also a continuation-in-part of U.S. patent application Ser. No. 15/973,966, filed May 8, 2018, which is a continuation of U.S. patent application Ser. No. 14/424,149, filed Feb. 26, 2015, now U.S. Pat. No. 9,991,010, which is a national stage entry under 35 U.S.C. § 371 of PCT/US2013/057115, filed Aug. 28, 2013, which claims priority to U.S. Provisional Patent Application Ser. No. 61/694,058, filed Aug. 28, 2012. The present application is also a continuation-in-part of U.S. patent application Ser. No. 16/401,891, filed May 2, 2019, which is a continuation of U.S. patent application Ser. No. 15/584,692, filed May 2, 2017, now U.S. Pat. No. 10,297,356, which is a continuation of U.S. patent application Ser. No. 14/912,754, filed Feb. 18, 2016, now U.S. Pat. No. 9,640,289, which is a national stage entry under 35 U.S.C. § 371 of PCT/US2015/027455, filed Apr. 24, 2015, which claims priority to U.S. Provisional Patent Application Ser. No. 61/983,606, filed Apr. 24, 2014. The present invention is also a continuation-in-part of U.S. patent application Ser. No. 16/022,935, filed Jun. 29, 2018, which is a continuation of U.S. patent application Ser. No. 14/811,454, filed Jul. 28, 2015, now U.S. Pat. No. 10,037,826, which claims priority to U.S. Provisional Patent Application Ser. No. 62/029,931, filed Jul. 28, 2014. The present application is also a continuation-in-part of U.S. patent application Ser. No. 16/584,892, filed Sep. 26, 2019, which is a continuation of U.S. patent application Ser. No. 14/877,217, filed Oct. 7, 2015, now U.S. Pat. No. 10,468,145, which claims priority to U.S. Provisional Patent Application Ser. No. 62/061,089, filed Oct. 7, 2014. The present application is also a continuation-in-part of U.S. patent application Ser. No. 16/871,221, filed May 11, 2020, which is a continuation of U.S. patent application Ser. No. 14/935,221, filed Nov. 6, 2015, now U.S. Pat. No. 10,650,933, which claims priority to U.S. Provisional Patent Application Ser. No. 62/076,138, filed Nov. 6, 2014. The present application is also a continuation-in-part of U.S. patent application Ser. No. 16/513,815, filed Jul. 17, 2019, which is a continuation of U.S. patent application Ser. No. 15/634,408, filed Jun. 27, 2017, now U.S. Pat. No. 10,418,137, which claims priority to U.S. Provisional Patent Application Ser. No. 62/355,057, filed Jun. 27, 2016. Damaged nuclear fuel is nuclear fuel that is in some way physically impaired. Such physical impairment can range from minor cracks in the cladding to substantial degradation on various levels. When nuclear fuel is damaged, its uranium pellets are no longer fully contained in the tubular cladding that confines the pellets from the external environment. Moreover, damaged nuclear fuel can be distorted from its original shape. As such, special precautions must be taken when handling damaged nuclear fuel (as compared to handling intact nuclear fuel) to ensure that radioactive particulate matter is contained. Please refer to USNRC's Interim Staff Guidance #2 for a complete definition of fuel that cannot be classified as “intact” and, thus, falls into the category of damaged nuclear fuel for purposes of this application. As used herein, damaged nuclear fuel also includes nuclear fuel debris. Containers and systems for handling damaged nuclear fuel are known. Examples of such containers and systems are disclosed in U.S. Pat. No. 5,550,882, issued Aug. 27, 1996 to Lehnart et al., and U.S. Patent Application Publication No. 2004/0141579, published Jul. 22, 2004 to Methling et al. While the general structure of a container and system for handling damaged nuclear fuel is disclosed in each of the aforementioned references, the containers and systems disclosed therein are less than optimal for a number of reasons, including inferior venting capabilities of the damaged nuclear fuel cavity, difficulty of handling, inability to be meet tight tolerances dictated by existing fuel basket structures, lack of adequate neutron shielding, and/or manufacturing complexity or inferiority. Thus, a need exists for an improved container and system for handling damaged nuclear fuel, and methods of making the same. Nuclear power plants currently store their spent fuel assemblies on site for a period after being removed from the reactor core. Such storage is typically accomplished by placing the spent fuel assemblies in closely packed fuel racks located at the bottom of on-site storage pools. The storage pools provide both radiation shielding and much needed cooling for the spent fuel assemblies. Fuel racks often contain a large number of closely arranged adjacent storage cells wherein each cell is capable of accepting a spent fuel assembly. In order to avoid criticality, which can be caused by the close proximity of adjacent fuel assemblies, a neutron absorbing material is positioned within the cells so that a linear path does not exist between any two adjacent cells (and thus the fuel assemblies) without passing through the neutron absorbing material. Early fuel racks utilized a layer of neutron absorbing material attached to the cell walls of the fuel rack. However, these neutron absorbing materials have begun to deteriorate as they have been submerged in water for over a decade. In order to either extend the period over which the fuel assemblies may be stored in these fuel racks, it is necessary to either replace the neutron absorber in the cell walls or to add an additional neutron absorber to the cell or the fuel assembly. In an attempt to remedy the aforementioned problems with the deteriorating older fuel racks, the industry developed removable neutron absorbing assemblies, such as those disclosed in U.S. Pat. Nos. 5,841,825; 6,741,669; and 6,442,227. Neutron absorbing assemblies such as these have become the primary means by which adjacent fuel assemblies are shielded from one another when supported in a submerged fuel rack. Thus, newer fuel racks are generally devoid of the traditional layer of neutron absorbing material built into the structure of the fuel rack itself that can degrade over time. Instead, fuel assembly loading and unloading procedures utilizing neutron absorbing assemblies have generally become standard in the industry. In older racks, the neutron absorbing assemblies are added over the older, often degrading, layer of neutron absorbing material. While the neutron absorbing assemblies disclosed in the prior art have proved to be preferable to the old fuel racks having the neutron absorbing material integrated into the cell walls, these neutron absorbing assemblies are less than optimal for a number of reasons, including without limitation complexity of construction, the presence of multiple welds, complicated securing mechanisms, and multi-layered walls that take up excessive space within the fuel rack cells. Additionally, with existing designs of neutron absorbing assemblies, the inserts themselves must be removed prior to or concurrently with the fuel assemblies in order to get the fuel assemblies out of the fuel rack. This not only complicates the handling procedure but also leaves certain cells in a potentially unprotected state. A freestanding fuel rack includes an array of vertical storage cavities used to store nuclear fuel in an upright configuration. Each storage cavity generally provides a square prismatic opening to store one spent nuclear or fresh (unburned) fuel. The cross section of the openings is slightly larger than that of the fuel assembly to facilitate the latter's insertion or withdrawal. From the structural standpoint, the fuel rack is a cellular structure supported on a number of pedestals that transfer the dead load of the rack and its stored fuel to the pool's slab. It is preferable to install the racks in a freestanding configuration to minimize cost and dose (if the pool is populated with irradiated fuel). The rack modules in a fuel pool typically have the appearance of a set of rectangular cavities arranged in a rectilinear array. The racks are typically separated by small gaps. Freestanding racks, however, are liable to slide or rotate during seismic event. If the plant's design basis is moderate then the kinematic movement of the racks may not be enough to cause inter-rack collisions or rack-to-wall impacts. However, if the seismic event is strong then the response of the racks may be too severe (e.g., large displacements, significant rack impact forces, etc.) to be acceptable. Reducing the kinematic response of the racks under strong seismic events (e.g., earthquakes) while preserving their freestanding disposition is therefore desirable. The present invention relates, in one aspect, generally to nuclear fuel containment, and more particularly to a capsule and related method for storing or transporting individual nuclear fuel pins or rods including damaged rods. Reactor pools store used fuel assemblies after removal and discharge from the reactor. The fuel assemblies and individual fuel rods therein may become damaged and compromised during the reactor operations, resulting in cladding defects, breaking, warping, or other damage. The resulting damaged fuel assemblies and rods are placed into the reactor pools upon removal and discharge from the reactor core. Eventually, the damaged fuel assemblies, rods, and/or fuel debris must be removed from the pools, thereby allowing decommissioning of the plants. The storage and transport regulations in many countries do not allow storage or transport of damaged fuel assemblies without encapsulation in a secondary capsule that provides confinement. Due to the high dose rates of used fuel assemblies post-discharge, encapsulating fuel assemblies is traditionally done underwater. Furthermore, some countries may require removal of individual damaged fuel rods from the fuel assembly and separate storage in such secondary capsules. Processes already exist for removing single rods from a used fuel assembly and encapsulation. Subsequent drying of damaged fuel after removal from the reactor pool using traditional vacuum drying is exceedingly challenging because water can penetrate through cladding defects and become trapped inside the cladding materials. An improved fuel storage system and method for drying, storing, and transporting damaged fuel rods is desired. In the nuclear power industry, the nuclear energy source is in the form of hollow zircaloy tubes filled with enriched uranium, known as fuel assemblies. Upon being depleted to a certain level, spent fuel assemblies are removed from a reactor. At this time, the fuel assemblies not only emit extremely dangerous levels of neutrons and gamma photons (i.e., neutron and gamma radiation) but also produce considerable amounts of heat that must be dissipated. It is necessary that the neutron and gamma radiation emitted from the spent fuel assemblies be adequately contained at all times upon being removed from the reactor. It is also necessary that the spent fuel assemblies be cooled. Because water is an excellent radiation absorber, spent fuel assemblies are typically submerged under water in a pool promptly after being removed from the reactor. The pool water also serves to cool the spent fuel assemblies by drawing the heat load away from the fuel assemblies. The water may also contain a dissolved neutron shielding substance. The submerged fuel assemblies are typically supported in the fuel pools in a generally upright orientation in rack structures, commonly referred to as fuel racks. It is well known that neutronic interaction between fuel assemblies increases when the distance between the fuel assemblies is reduced. Thus, in order to avoid criticality (or the danger thereof) that can result from the mutual inter-reaction of adjacent fuel assemblies in the racks, it is necessary that the fuel racks support the fuel assemblies in a spaced manner that allows sufficient neutron absorbing material to exist between adjacent fuel assemblies. The neutron absorbing material can be the pool water, a structure containing a neutron absorbing material, or combinations thereof. Fuel racks for high density storage of fuel assemblies are commonly of cellular construction with neutron absorbing plate structures (i.e., shields) placed between the storage cells in the form of solid sheets. For fuel assemblies that have a square horizontal cross-sectional profile, the storage cells are usually long vertical square tubes which are open at the top through which the fuel elements are inserted. In order to maximize the number of fuel assemblies that can be stored in a single rack, the fuel racks for these square tubes are formed by a rectilinear array of the square tubes. Similarly, for fuel assemblies that have a hexagonal horizontal cross-sectional profile, the storage cells are usually long vertical hexagonal tubes which are open at the top through which the fuel elements are inserted. For such storage cells, in order to maximize the number of fuel assemblies that can be stored in a single rack, the fuel racks for these hexagonal tubes are formed by a honeycomb array of the hexagonal tubes. Regardless of whether the storage cells are square tubes or hexagonal tubes, the storage cells of some fuel racks may include double walls that can serve two functions. The first function of a double cell wall may be to encapsulate neutron shield sheets to protect the neutron shield from corrosion or other deterioration resulting from contact with water. The second function of a double cell wall may be to provide flux traps to better prevent undesirable heat build-up within the array of storage cells. When both of these double-wall functions are incorporated into a fuel rack array, it necessarily decreases the storage density capability. Thus, improvements are desired in design a fuel racks that provide both these functions and improve the overall storage density capability. The present invention generally relates, in one embodiment, to storage of nuclear fuel assemblies, and more particularly to an improved spent fuel pool for wet storage of such fuel assemblies. A spent fuel pool (sometimes, two or more) is an integral part of every nuclear power plant. At certain sites, standalone wet storage facilities have also been built to provide additional storage capacity for the excess fuel discharged by the reactors. An autonomous wet storage facility that serves one or more reactor units is sometimes referred to by the acronym AFR meaning “Away-from-Reactor.” While most countries have added to their in-plant used fuel storage capacity by building dry storage facilities, the French nuclear program has been the most notable user of AFR storage. As its name implies, the spent fuel pool (SFP) stores the fuel irradiated in the plant's reactor in a deep pool of water. The pool is typically 40 feet deep with upright Fuel Racks positioned on its bottom slab. Under normal storage conditions, there is at least 25 feet of water cover on top of the fuel to ensure that the dose at the pool deck level is acceptably low for the plant workers. Fuel pools at most (but not all) nuclear plants are at grade level, which is desirable from the standpoint of structural capacity of the reinforced concrete structure that forms the deep pond of water. To ensure that the pool's water does not seep out through the voids and discontinuities in the pool slab or walls, fuel pools in nuclear plants built since the 1970s have always been lined with a thin single-layer stainless steel liner (typically in the range of 3/16 inch to 5/16 inch thick). The liner is made up of sheets of stainless steel (typically ASTM 240-304 or 304L) seam welded along their contiguous edges to form an impervious barrier between the pool's water and the undergirding concrete. In most cases, the welded liner seams are monitored for their integrity by locating a leak chase channel underneath them (see, e.g. FIG. 57). The leak chase channels' detection ability, however, is limited to welded regions only; the base metal area of the liner beyond the seams remains un-surveilled. The liners have generally served reliably at most nuclear plants, but isolated cases of water seepage of pool water have been reported. Because the pool's water bears radioactive contaminants (most of it carried by the crud deposited on the fuel during its “burn” in the reactor), leaching out of the pool water to the plant's substrate, and possibly to the underground water, is evidently inimical to public health and safety. To reduce the probability of pool water reaching the ground water, the local environment and hence some AFR pools have adopted the pool-in-pool design wherein the fuel pool is enclosed by a secondary outer pool filled with clean water. In the dual-pool design, any leakage of water from the contaminated pool will occur into the outer pool, which serves as the barrier against ground water contamination. The dual pool design, however, has several unattractive aspects, viz.: (1) the structural capacity of the storage system is adversely affected by two reinforced concrete containers separated from each other except for springs and dampers that secure their spacing; (2) there is a possibility that the outer pool may leak along with the inner pool, defeating both barriers and allowing for contaminated water to reach the external environment; and (3) the dual-pool design significantly increases the cost of the storage system. Prompted by the deficiencies in the present designs, a novel design of a spent nuclear fuel pool that would guarantee complete confinement of pool's water and monitoring of the entire liner structure including seams and base metal areas is desirable. High density spent fuel racks are used in Light Water Reactor (LWR) installations to store nuclear fuel assemblies underwater in deep ponds of water known as Spent Fuel Pools. The current state-of-the-art in the design of Fuel Racks is described in “Management of Spent Nuclear Fuel,” Chapter 53, by Drs. Tony Williams and Kris Singh in the ASME monograph Companion Guide to the ASME Boiler & Pressure Vessel Code, Third (3rd) Edition, edited by K. R. Rao (2009). As described in the above mentioned chapter, contemporary fuel racks are cellular structures mounted on a common Baseplate supported on four or more pedestals and made up of a rectangular assemblage of “storage cells” with plates (or panels) of neutron absorber affixed to the walls separating each cell. The neutron absorber serves to control the reactivity of the fuel assemblies arrayed in close proximity to each other. The neutron absorber is typically made of a metal matrix composite such as aluminum and boron carbide, the boron serving to capture the thermalized neutrons emitted by the fuel to control reactivity. Typical areal density of the B-10 isotope (the neutron capture agent in boron carbide) in the absorber plates used in BWR and PWR racks are 0.02 and 0.03 gm/sq. cm, respectively. The overwhelming majority of fuel racks in use in the United States have discrete panels of neutron absorber secured to the side walls of the storage cell boxes. To eliminate the separate neutron absorber panels that must be affixed to the cell walls, an alternative design that uses borated stainless steel that renders both neutron capture and structural function, has been used in the industry but failed to gain wide acceptance because of the limited quantity of boron that can be introduced in the stainless steel grain structure and other structural limitations. In view of the shortcomings of the alternative designs using borated stainless steel, different alternative designs are needed to fuel racks in order to eliminate the need to use separate neutron absorber panels. A conventional free-standing, high density nuclear fuel storage rack is a cellular structure typically supported on a set of pedestals from the floor or bottom slab of the water-filled spent fuel pool. The bottom extremity of each fuel storage cell is welded to a common baseplate which serves to provide the support surface for the upwardly extending vertical storage cells and stored nuclear fuel therein. The cellular region comprises an array of narrow prismatic cavities formed by the cells which are each sized to accept a single nuclear fuel assembly comprising a plurality of new or spent nuclear fuel rods. The term “active fuel region” denotes the vertical space above the baseplate within the rack where the enriched uranium is located. High density fuel racks used to store used nuclear fuel employ a neutron absorber material to control reactivity. The commercially available neutron absorbers are typically in a plate or sheet form and are either metal or polymer based. The polymeric neutron absorbers commonly used in the industry were sold under trade names Boraflex and Tetrabor, with the former being the most widely used material in the 1980s. The neutron absorber panels have been typically installed on the four walls of the storage cells encased in an enveloping sheathing made of thin gage stainless steel attached to the cell walls in the active fuel region. Unfortunately, the polymeric neutron absorbers have not performed well in service. Widespread splitting and erosion of Boraflex and similar degradation of Tetrabor have been reported in the industry, forcing the plant owners to resort to reducing the density of storage (such as a checkered board storage arrangement) thereby causing an operational hardship to the plant. A neutron absorber apparatus is desired which can be retrofit in existing fuel racks suffering from neutron absorber material degradation in order to fully restore reactivity reduction capacity of the storage cells. In one embodiment, the invention can be a method of forming an elongated tubular container for receiving damaged nuclear fuel, the method comprising: a) extruding, from a material comprising a metal and a neutron absorber, an elongated tubular wall having a container cavity; b) forming, from a material comprising a metal that is metallurgically compatible with the metal of the elongated tubular wall, a bottom cap comprising a first screen having a plurality of openings; and c) autogenously welding the bottom cap to a bottom end of the elongated tubular wall, the plurality of openings of the first screen forming vent passageways to a bottom of the container cavity. In another embodiment, the invention can be a container for receiving damaged nuclear fuel, the method comprising: an extruded tubular wall forming a container cavity about a container axis, the extruded tubular wall formed of a metal matrix composite having neutron absorbing particulate reinforcement; a bottom cap coupled to a bottom end of the extruded tubular wall; a top cap detachably coupled to a top end of the extruded tubular wall; a first screen comprising a plurality of openings that define lower vent passageways into a bottom of the container cavity; and a second screen comprising a plurality of openings that define upper vent passageways into a top of the container cavity. In yet another embodiment, the invention can be a system for storing and/or transporting nuclear fuel comprising: a vessel comprising defining a vessel cavity and extending along a vessel axis; a fuel basket positioned within the vessel cavity, the fuel basket comprising a grid forming a plurality of elongated cells, each of the cells extending along a cell axis that is substantially parallel to the vessel axis; and at least one elongated tubular container comprising a container cavity containing damaged nuclear fuel positioned within one of the cells, the elongated tubular container comprising: an extruded tubular wall forming a container cavity about a container axis, the extruded tubular wall formed of a metal matrix composite having neutron absorbing particulate reinforcement; a bottom cap coupled to a bottom end of the extruded tubular wall; a top cap detachably coupled to a top end of the extruded tubular wall; a first screen comprising a plurality of openings that define lower vent passageways into a bottom of the container cavity; and a second screen comprising a plurality of openings that define upper vent passageways into a top of the container cavity. In still another embodiment, the invention can be a system for storing and/or transporting nuclear fuel comprising: a vessel defining a vessel cavity and extending along a vessel axis; a fuel basket positioned within the vessel cavity, the fuel basket comprising a plurality of elongated cells; an elongated tubular container positioned within one of the cells, the elongated tubular container comprising: an elongated tubular wall forming a container cavity about a container axis, the tubular wall comprising a top portion having a plurality of locking apertures and a top edge defining a top opening into the container cavity; a bottom cap coupled to a bottom end of the elongated tubular wall; a top cap comprising a plurality of locking elements that are alterable between a retracted state and an extended state, the locking elements biased into the extended state; a first screen comprising a plurality of openings that define lower vent passageways between the vessel cavity and a bottom of the container cavity; a second screen comprising a plurality of openings that define upper vent passageways between the vessel cavity and a top of the container cavity; and the top cap and the elongated tubular wall configured so that upon the top cap being inserted through the top opening, contact between the locking element and the elongated tubular wall forces the locking elements into a retracted state, and wherein upon the locking element becoming aligned with the locking apertures, the locking elements automatically returning the extended state such that the locking member protrude into the locking apertures, thereby detachably coupling the top cap to elongated tubular wall. In a further embodiment, the invention can be a system for storing and/or transporting nuclear fuel comprising: a vessel defining a vessel cavity and extending along a vessel axis; a fuel basket positioned within the vessel cavity, the fuel basket comprising a plurality of elongated cells; an elongated tubular container comprising a container cavity for containing damaged nuclear fuel positioned within one of the cells, the elongated tubular container comprising: a first screen comprising a plurality of openings that define lower vent passageways between the vessel cavity and a bottom of the container cavity, the plurality of openings of the first screen comprising a lowermost opening that is a first distance from a floor of the vessel cavity and an uppermost opening that is a second distance from the floor of the vessel cavity, the second distance being greater than the first distance; and a second screen comprising a plurality of openings that define upper vent passageways between the vessel cavity and a top of the container cavity. In an even further embodiment, the invention can be a system for storing and/or transporting nuclear fuel comprising: a vessel defining a vessel cavity and extending along a vessel axis; a fuel basket positioned within the vessel cavity, the fuel basket comprising a plurality of elongated cells; an elongated tubular container comprising a container cavity for containing damaged nuclear fuel positioned within one of the cells, the elongated tubular container comprising: a first screen comprising a plurality of openings that define lower vent passageways between the vessel cavity and a bottom of the container cavity, the first screen located on an upstanding portion of the elongated tubular container that is substantially non-perpendicular to the vessel axis; and a second screen comprising a plurality of openings that define upper vent passageways between the vessel cavity and a top of the container cavity. In a still further embodiment, the invention can be a damaged fuel container, or system incorporating the same, in which the one or more of the screens of the container are integrally formed into the body of the container. In another aspect of the present invention, a neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. The absorption sheet extends along the corner spine along a greater length than the guide sheet. In yet another aspect of the present invention, a neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. At least one of the walls also includes a locking protuberance coupled to the respective guide sheet and protruding through an opening formed in the respective absorption sheet. In still another aspect of the present invention, a system for supporting spent nuclear fuel in a submerged environment includes a fuel rack, a fuel assembly, and a neutron absorbing apparatus. The fuel rack includes an array of cells, with each cell being separated from adjacent cells by a cell wall. The fuel assembly is positioned within one of the cells, and the neutron absorbing apparatus is also disposed within that cell. The neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. At least one of the cell wall in which the fuel assembly is disposed, adjacent the first wall or the second wall of the neutron absorbing apparatus, and the first wall or the second wall include a locking protuberance positioned to retain the neutron absorbing apparatus in the first cell during removal of the fuel assembly from the first cell. In another aspect of the present invention, a method of retrofitting a spent nuclear fuel cell storage system includes inserting a neutron absorbing apparatus into one cell of an array of cells, wherein each cell is separated from each adjacent cell by a cell wall. The neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. At least one of the walls also includes a first locking protuberance coupled to the respective guide sheet and protruding through an opening formed in the respective absorption sheet. The method further includes creating a second locking protuberance in a first cell wall adjacent the neutron absorbing apparatus, wherein the first locking protuberance and the second locking protuberance are positioned to interlock to retain the neutron absorbing apparatus in the one cell. In yet another aspect of the present invention, any of the foregoing aspects may be employed in combination. Accordingly, an improved neutron absorption apparatus for spent nuclear fuel pools and casks is disclosed. Advantages of the improvements will be apparent from the drawings and the description of the preferred embodiment. In another embodiment, the present invention is directed toward a system and method for minimizing lateral movement of one or more nuclear fuel storage racks in a storage pool during a seismic event. In both the system and the method. Lateral movement of a storage rack may be limited either by limiting lateral movement of the rack toward the side wall of the storage pool, or by limiting lateral movement of a first storage rack with respect to another object. In another aspect of the present invention, a system for storing nuclear fuel includes a nuclear fuel storage rack and a bearing pad. The storage rack includes an array of cells, each cell configured to receive and store nuclear fuel rods, a base plate configured to support the array of cells, and a support structure configured to support the base plate and to allow cooling fluid to circulate under and up through apertures in the base plate. The bearing pad is coupled to the support structure and configured to limit lateral movement of the storage rack independent from lateral movement of the bearing pad. The base plate defines a base plate profile in a horizontal plane of the base plate, and the bearing plate defines a bearing pad profile in the horizontal plane of the base plate, wherein the bearing pad profile extends outside of the base plate profile. In another aspect of the present invention, the system for storing nuclear fuel includes first and second adjacent storage racks and a bearing pad. Each storage rack includes, respectively, an array of cells, each cell configured to receive and store nuclear fuel rods, a base plate configured to support the array of cells, and a support structure configured to support the base plate and to allow cooling fluid to circulate under and up through apertures in the base plate. The bearing pad is coupled to the support structure of each of the storage racks, and it is configured to limit lateral movement of each storage rack independent from lateral movement of the bearing pad. In a further aspect of the present invention, a method of placing a nuclear fuel storage rack into a storage pool includes placing a bearing pad on the bottom of the storage pool, then placing a storage rack into the storage pool. The storage rack includes an array of cells, a base plate configured to support the array of cells, and a support structure configured to support the base plate, wherein each cell of the array of cells being configured to receive and store nuclear fuel rods. In placing the storage rack, the bearing pad is coupled to the support structure, and the bearing pad is configured to limit lateral movement of the storage rack independent from lateral movement of the bearing pad. The base plate defines a base plate profile in a horizontal plane of the base plate, the bearing pad defines a bearing pad profile in the horizontal plane of the base plate, and the bearing pad profile extends outside of the base plate profile. In another aspect of the present invention, a method of placing a first nuclear fuel storage rack and a second nuclear fuel storage rack into a storage pool includes placing a bearing pad on a bottom of a storage pool, placing the first storage rack into the storage pool, then placing the second storage rack into the storage pool. Each storage rack includes, respectively, an array of cells, each cell configured to receive and store nuclear fuel rods, a base plate configured to support the array of cells, and a support structure configured to support the base plate and to allow cooling fluid to circulate under and up through apertures in the base plate. The first storage rack is placed into the storage pool so that the bearing pad is coupled to the respective support structure of the first storage rack. The second storage rack is placed into the storage pool so that the bearing pad is coupled to the respective support structure of the second storage rack. The bearing pad is configured to limit lateral movement of each storage rack independent from lateral movement of the bearing pad. In yet another aspect of the present invention, any of the foregoing aspects may be employed in combination. Accordingly, an improved system and method for minimizing lateral movement of one or more nuclear fuel storage racks in a storage pool during a seismic event are disclosed. Advantages of the improvements will be apparent from the drawings and the description of the preferred embodiment. A nuclear fuel storage system and related method are provided that facilitates drying and storage of individual fuel rods, which may be used for damaged and intact fuel rods and debris. The system includes a capsule that is configured for holding a plurality of fuel rods, and further for drying the internal cavity of the capsule and fuel rods stored therein using known inert forced gas dehydration (FGD) techniques or other methods prior to long term storage. Existing forced gas dehydration systems and methods that may be used with the present invention can be found in commonly owned U.S. Pat. Nos. 7,096,600, 7,210,247, 8,067,659, 8,266,823, and 7,707,741, which are all incorporated herein by reference in their entireties. In one embodiment, a storage capsule for nuclear fuel rods includes: an elongated body defining a vertical centerline axis, the body comprising an open top end, a bottom end, and sidewalls extending between the top and bottom ends; an internal cavity formed within the body; a lid attached to and closing the top end of the body; and an array of axially extending fuel rod storage tubes disposed in the cavity; wherein each storage tube has a transverse cross section configured and dimensioned to hold no more than one fuel rod. In one embodiment, a fuel storage system for storing nuclear fuel rods includes: an elongated capsule defining a vertical centerline axis, the capsule comprising a top end, a bottom end, and sidewalls extending between the top and bottom ends; an internal cavity formed within the capsule; a lid attached to the top end of the capsule, the lid including an exposed top surface and a bottom surface; an upper tubesheet and a lower tubesheet disposed in the cavity; a plurality of vertically oriented fuel rod storage tubes extending between the upper and lower tubesheets; and a central drain tube extending between the upper and lower tubesheets; wherein each storage tube has a transverse cross section configured and dimensioned to hold no more than one fuel rod. A method for storing nuclear fuel rods is provided. The method includes: providing an elongated vertically oriented capsule including an open top end, a bottom end, and an internal cavity, the capsule further including a plurality of vertically oriented fuel rod storage tubes each having a top end spaced below the top end of the capsule, the storage tubes each having a transverse cross section configured and dimensioned to hold no more than a single fuel rod; inserting a first fuel rod into a first storage tube; inserting a second fuel rod into a second storage tube; attaching a lid to the top end of the capsule; and sealing the lid to the capsule to form a gas tight seal. A method for storing and drying nuclear fuel rods includes: providing an elongated vertically oriented capsule including an open top end, a bottom end, and an internal cavity, the capsule further including a plurality of vertically oriented fuel rod storage tubes each having a top end spaced below the top end of the capsule, the storage tubes each having a transverse cross section configured and dimensioned to hold no more than a single fuel rod; inserting a fuel rod into each of the storage tubes; attaching a lid to the top end of the capsule, the lid including a gas supply flow conduit extending between top and bottom surfaces of the lid and a gas return flow conduit extending between the top and bottom surfaces of the lid; sealing the lid to the capsule to form a gas tight seal; pumping an inert drying gas from a source through the gas supply conduit into the cavity of the capsule; flowing the gas through each of the storage tubes; collecting the gas leaving the storage tubes; and flowing the gas through the gas return conduit back to the source. The present invention is directed to an apparatus for supporting spent nuclear fuel. Specifically, the apparatus enables the high density storage of spent nuclear fuel. In one aspect of the invention, a fuel rack apparatus includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to the upper surface of the base plate in a side-by-side arrangement to form a rectilinear array of the storage tubes. Each of the storage tubes extends along a longitudinal axis and includes: a rectangular outer tube having an inner surface defining an inner cavity; a first chevron plate comprising a first wall plate and a second wall plate; and a second chevron plate comprising a first wall plate and a second wall plate. The first and second chevron plates are positioned in the inner cavity in opposing relation to divide the inner cavity into: (1) a first chamber formed between the first wall plate of the first chevron plate and a first corner section of the rectangular outer tube; (2) a second chamber formed between the second wall plate of the first chevron plate and a second corner section of the rectangular outer tube; (3) a third chamber formed between the first wall plate of the second chevron plate and a third corner section of the rectangular outer tube; (4) a fourth chamber formed between the second wall plate of the second chevron plate and a fourth corner section of the rectangular outer tube; and (5) a fuel storage cell having a hexagonal transverse cross-section and configured to receive a fuel assembly containing spent nuclear fuel. In another aspect of the invention, a fuel rack apparatus for storing spent nuclear fuel includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to and extending upward from the upper surface of the base plate, the storage tubes arranged in a side-by-side arrangement to form an array of the storage tubes. Each of the storage tubes extend along a longitudinal axis and include: an outer tube having an inner surface defining an inner cavity; and an inner plate-assemblage positioned within the outer tube that divides the inner cavity into a plurality of interior flux trap chambers and a fuel storage cell. In yet another aspect of the invention, a fuel rack apparatus includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to the upper surface of the base plate in a side-by-side arrangement to form a rectilinear array of the storage tubes. Each of the storage tubes extends along a longitudinal axis and includes: a rectangular outer tube having an inner surface defining an inner cavity; and a plurality of wall plates positioned in the inner cavity that divide the inner cavity into: (1) a first interior flux chamber formed between a first one of the wall plates and a first corner section of the rectangular outer tube; (2) a second interior flux chamber formed between a second one of the wall plates and a second corner section of the rectangular outer tube; (3) a third interior flux chamber formed between a third one of the wall plates and a third corner section of the rectangular outer tube; (4) a fourth interior flux chamber formed between a fourth one of wall plates and a fourth corner section of the rectangular outer tube; and (5) a fuel storage cell having a hexagonal transverse cross-section and configured to receive a fuel assembly containing spent nuclear fuel. In an embodiment, the present invention provides an environmentally sequestered spent fuel pool system having a dual impervious liner system and leak detection/evacuation system configured to collect and identify leakage in the interstitial space formed between the liners. The internal cavity of the pool has not one but two liners layered on top of each other, each providing an independent barrier to the out-migration (emigration) of pool water. Each liner encompasses the entire extent of the water occupied space and further extends above the pool's “high water level.” The top of the pool may be equipped with a thick embedment plate (preferably 2 inches thick minimum in one non-limiting embodiment) that circumscribes the perimeter of the pool cavity at its top extremity along the operating deck of the pool. Each liner may be independently welded to the top embedment plate. The top embedment plate features at least one telltale hole, which provides direct communication with the interstitial space between the two liner layers. In one implementation, a vapor extraction system comprising a vacuum pump downstream of a one-way valve is used to draw down the pressure in the inter-liner space through the telltale hole to a relatively high state of vacuum. The absolute pressure in the inter-liner space (“set pressure”) preferably should be such that the pool's bulk water temperature is above the boiling temperature of water at the set pressure as further described herein. In one embodiment, an environmentally sequestered nuclear spent fuel pool system includes: a base slab; a plurality of vertical sidewalls extending upwards from and adjoining the base slab, the sidewalls forming a perimeter; a cavity collectively defined by the sidewalls and base slab that holds pool water; a pool liner system comprising an outer liner adjacent the sidewalls, an inner liner adjacent the outer liner and wetted by the pool water, and an interstitial space formed between the liners; a top embedment plate circumscribing the perimeter of the pool at a top surface of the sidewalls adjoining the cavity; and the inner and outer sidewalls having top terminal ends sealably attached to the embedment plate. In another embodiment, an environmentally sequestered nuclear spent fuel pool with leakage detection system includes: a base slab; a plurality of vertical sidewalls extending upwards from and adjoining the base slab, the sidewalls forming a perimeter; a cavity collectively defined by the sidewalls and base slab that holds pool water; at least one fuel storage rack disposed in the cavity that holds a nuclear spent fuel assembly containing nuclear fuel rods that heat the pool water; a pool liner system comprising an outer liner adjacent the sidewalls and base slab, an inner liner adjacent the outer liner and wetted by the pool water, and an interstitial space formed between the liners; a top embedment plate circumscribing the perimeter of the pool, the embedment plate embedded in the sidewalls adjoining the cavity; the inner and outer liners attached to the top embedment plate; a flow plenum formed along the sidewalls that is in fluid communication with the interstitial space; and a vacuum pump fluidly coupled to the flow plenum, the vacuum pump operable to evacuate the interstitial space to a negative set pressure below atmospheric pressure. A method for detecting leakage from a nuclear spent fuel pool is provided. The method includes: providing a spent fuel pool comprising a plurality of sidewalls, a base slab, a cavity containing cooling water, and a liner system disposed in the cavity including an outer liner, an inner liner, and an interstitial space between the liner; placing a fuel storage rack in the pool; inserting at least one nuclear fuel assembly into the storage rack, the fuel assembly including a plurality of spent nuclear fuel rods; heating the cooling water in the pool to a first temperature from decay heat generated by the spent nuclear fuel rods; drawing a vacuum in the interstitial space with a vacuum pump to a negative pressure having a corresponding boiling point temperature less than the first temperature; collecting cooling water leaking from the pool through the liner system in the interstitial space; converting the leaking cooling water into vapor via boiling; and extracting the vapor from the interstitial space using the vacuum pump; wherein the presence of vapor in the interstitial space allows detection of a liner breach. The method may further include discharging the vapor extracted by the vacuum pump through a charcoal filter to remove contaminants. The method may further include: monitoring a pressure in the interstitial space; detecting a first pressure in the interstitial space prior to collecting cooling water leaking from the pool through the liner system in the interstitial space; and detecting a second pressure higher than the first pressure after collecting cooling water leaking from the pool through the liner system in the interstitial space; wherein the second pressure is associated with a cooling water leakage condition. In another embodiment, the present invention is directed toward a fuel rack for the storage of spent nuclear fuel. The rack employs a plurality of slotted plates to form an array of cells for storing nuclear fuel assemblies. The slotted plates are constructed from two different types of materials which are metallurgically incompatible, one which provides strength to the array of cells and the other which is a neutron absorber. The design reduces the complexity of the design for fuel racks, while at the same time still providing the necessary safety systems for the long term storage of nuclear fuel. In one aspect, the invention may be a fuel rack for nuclear fuel assemblies, the fuel rack including a base plate and an array of cells for holding the fuel assemblies. The array of cells includes: a plurality of first slotted plates slidably interlocked with one another to form a top portion of the array of cells, the plurality of first slotted plates formed of a first material; a plurality of second slotted plates slidably interlocked with one another to form a middle portion of the array of cells, the plurality of second slotted plates formed of a second material, the first and second materials being metallurgically incompatible; and a plurality of third slotted plates slidably interlocked with one another to form a bottom portion of the array of cells, the plurality of third slotted plates formed of the first material and connected to a top surface of the base plate. In another aspect, the invention may be a nuclear fuel storage apparatus including: a fuel assembly and a fuel rack. The fuel assembly has a top section, a middle section, and a bottom section, with nuclear fuel being stored within the middle section. The fuel rack includes a base plate and an array of cells, with the fuel assembly located in a first cell of the array of cells. The array of cells includes: a plurality of first slotted plates slidably interlocked with one another to form a top portion of the array of cells, the plurality of first slotted plates formed of a first material; a plurality of second slotted plates slidably interlocked with one another to form a middle portion of the array of cells, the plurality of second slotted plates formed of a second material, the first and second materials being metallurgically incompatible, and the middle section of the fuel assembly located entirely within the middle portion of the first cell of the array of cells; and a plurality of third slotted plates slidably interlocked with one another to form a bottom portion of the array of cells, the plurality of third slotted plates formed of the first material and connected to a top surface of the base plate. In still another aspect, the invention may be a fuel rack for nuclear fuel assemblies, the fuel rack including: a base plate; an array of cells for holding fuel assemblies, the array of cells including: a plurality of first slotted plates slidably interlocked with one another to form a top portion of the array of cells, the plurality of first slotted plates welded together and formed of a first material; a plurality of second slotted plates slidably interlocked with one another to form a middle portion of the array of cells, the plurality of second slotted plates formed of a second material, the first and second materials being metallurgically incompatible; and a plurality of third slotted plates slidably interlocked with one another to form a bottom portion of the array of cells, the plurality of third slotted plates formed of the first material and welded to a top surface of the base plate; and a plurality of tie members, each tie member welded to each of the top and bottom portions of the array of cells. Embodiments of the present invention provide a neutron absorber insert system which can be readily added in situ to existing storage cells of the fuel rack having degraded neutron absorbers and reduced reactivity reduction capacity. The system comprises a plurality of neutron absorber apparatuses which may be in the form of absorber inserts configured for direct insertion into and securement to the fuel storage cells. The inserts have a low-profile small and thin cross sectional footprint which does not significantly reduce the storage capacity of each storage cell. A fuel assembly may be inserted into a central longitudinally-extending cavity of the insert and removed therefrom without first removing the insert. The inserts include a locking feature which is automatically deployed and secures the insert in the cell, as further described herein. Advantageously, the absorber insert may utilize an available edge surface on an existing storage tube of the fuel rack which can be engaged by the locking feature of the absorber tube. This eliminates the need for modifying the existing fuel rack in order to accommodate the insert, thereby saving time and expense. In one embodiment, the edge surface may be part of an existing neutron absorber sheathing structure on the fuel storage tube. The inserts may advantageously be deployed in the existing fuel rack storage cells via remote handling equipment such as cranes while the rack remains submerged underwater in the spent fuel pool. In one aspect, a neutron absorber apparatus for a nuclear fuel storage system includes: a fuel rack comprising a vertical longitudinal axis and plurality of longitudinally-extending storage cells, each cell comprising a plurality of cell sidewalls defining a cell cavity configured for storing nuclear fuel therein; a sheath integrally attached to a first cell sidewall of a first cell and defining a sheathing cavity configured for holding a neutron absorber material; an absorber insert comprising plural longitudinally-extending neutron absorber plates each comprising a neutron absorber material, the insert disposed in the first cell; and an elastically deformable locking protrusion disposed on one of the absorber plates, the locking protrusion resiliently movable between an outward extended position and an inward retracted position; the locking protrusion lockingly engaging the sheath to axially restrain the insert and prevent removal of the insert from the first cell. In another aspect, a neutron absorber apparatus for a nuclear fuel storage system includes: a fuel rack comprising a vertical longitudinal axis and plurality of longitudinally-extending storage tubes each defining a cell, each storage tube comprising a plurality of tube sidewalls defining a primary cavity; an absorber insert insertably disposed in the primary cavity of a first storage tube, the absorber insert comprising a plurality of absorber plates arranged to form a longitudinally-extending neutron absorber tube having an exterior and an interior defining a secondary cavity configured for storing a nuclear fuel assembly therein, each absorber plate formed of a neutron absorber material; an upper stiffening band extending perimetrically around an upper end of the absorber tube, the upper stiffening band attached to the exterior of the absorber tube and protruding laterally outwards beyond the absorber plates to engage the tube sidewalls of the first storage tube; a lower stiffening band extending perimetrically around a lower end of the absorber tube and disposed at least partially inside the secondary cavity, the lower stiffening band attached to the interior of the absorber tube; wherein the absorber plates of the insert assembly are spaced laterally apart from the tube sidewalls of the first storage tube by the upper stiffening band forming a clearance gap therebetween. In another aspect, a neutron absorber apparatus for a nuclear fuel storage system includes: a fuel rack comprising a plurality of longitudinally-extending storage cells, each cell comprising a plurality of cell walls defining a cell cavity for storing nuclear fuel; a longitudinally-extending absorber tube insertably disposed in a first cell of the fuel rack and having an exterior and an interior, the absorber tube comprising: an elongated chevron-shaped first absorber plate comprising a first section and a second section angularly bent to the first section along a bend line of the first absorber plate; an elongated chevron-shaped second absorber plate comprising a third section and a fourth section angularly bent to the third section along a bend line of the second absorber plate; an upper stiffening band extending perimetrically around upper ends of the first and second absorber plates and coupling the first and second absorber plates together. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. All drawings are schematic and not necessarily to scale. Parts shown and/or given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. References herein to a figure number (e.g. FIG. 1) shall be construed to be a reference to all subpart figures in the group (e.g. FIGS. 1A, 1B, etc.) unless otherwise indicated. The following description of the illustrated embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. The description of illustrative embodiments according to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments of the invention disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,” “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivatives thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation unless explicitly indicated as such. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Moreover, the features and benefits of the invention are illustrated by reference to the exemplified embodiments. Accordingly, the invention expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features; the scope of the invention being defined by the claims appended hereto. Multiple inventive concepts are described herein and are distinguished from one another using headers in the description that follows. Specifically, FIGS. 1-12 are relevant to a first inventive concept, FIGS. 13-20 are relevant to a second inventive concept, FIGS. 21-31 are relevant to a third inventive concept, FIGS. 32-49 are relevant to a fourth inventive concept, FIGS. 50-56 are relevant to a fifth inventive concept, FIGS. 57-63 are relevant to a sixth inventive concept, FIGS. 64-73 are relevant to a seventh inventive concept, and FIGS. 74-85 are relevant to an eighth inventive concept. The first through eighth inventive concepts should be considered in isolation from one another. It is possible that there may be conflicting language or terms used in the description of the first through eighth inventive concepts. For example, it is possible that in the description of the first inventive concept a particular term may be used to have one meaning or definition and that in the description of the second inventive concept the same term may be used to have a different meaning or definition. In the event of such conflicting language, reference should be made to the disclosure of the relevant inventive concept being discussed. Similarly, the section of the description describing a particular inventive concept being claimed should be used to interpret claim language when necessary. I. Inventive Concept 1 With reference to FIGS. 1-12, a first inventive concept will be described. Referring first to FIGS. 1-4 concurrently, a damaged fuel container (“DFC”) 100 according to an embodiment of the present invention is illustrated. The DFC 100 incorporates an inventive design (and is formed by an inventive method) that allows high density packaging of damaged fuel in pressure vessels, such as metal casks or multi-purpose canisters (described in greater detail below). The DFC 100 can be used to package damaged nuclear fuel at nuclear reactors, such as the Fukushima Daiichi site. The DFC 100 can be used to safely containerize nuclear fuel of compromised cladding integrity and is a unitary waste package for the fuel that may be in various stages of dismemberment ranging from minor cracks in the cladding to its substantial degradation. As described in greater detail below, the DFC 100 is designed to be loaded with damaged nuclear fuel and positioned within a fuel basket which, in turn, is housed in a pressure vessel such as a metal cask or a multi-purpose canister. The DFC 100 is an elongated tubular container that extends along a container axis C-C. As will become more apparent from the description below, the DFC 100 is specifically designed so as to not form a fluid-tight container cavity 101 therein. This allows the container cavity 101 of the DFC 100, and its damaged nuclear fuel payload, to be adequately dried for dry storage using standard dry storage dehydration procedures. Suitable dry storage dehydration operations and equipment that can be used to dry the DFC 100 (and the system 999) are disclosed in, for example: U.S. Patent Application Publication No. 2006/0288607, published Dec. 28, 2006 to Singh; U.S. Patent Application Publication No. 2009/0158614, published Jun. 2, 2009 to Singh et al.; and U.S. Patent Application Publication No. 2010/0212182, published Aug. 22, 2010 to Singh. While a fluid-tight boundary is not formed by the DFC 100, the DFC 100 (when fully assembled as shown in FIGS. 1-4) creates a particulate confinement boundary for its damaged nuclear fuel payload, thereby preventing radioactive particles and debris from escaping the container cavity 101. The DFC 100 generally comprises an elongated tubular wall 10, a bottom cap 20 and a top cap 30. In one embodiment, the elongated tubular wall 10 is formed of a material comprising a metal and a neutron absorber. As used herein the term “metal” includes metals and metal alloys. In certain embodiments, suitable metals may include without limitation aluminum, steel, lead, and titanium while suitable neutron absorbers may include without limitation boron, boron carbide and carborundem. As used herein, the term “aluminum” includes aluminum alloys. In one specific embodiment, the metal is an aluminum and the neutron absorber material is boron or boron carbide. In other embodiments, the elongated tubular wall 10 is formed entirely of a metal matrix composite having neutron absorbing particulate reinforcement. Suitable metal matrix composites having neutron absorbing particulate reinforcement include, without limitation, a boron carbide aluminum matrix composite material, a boron aluminum matrix composite material, a boron carbide steel matrix composite material, a carborundum aluminum matrix composite material, a carborundum titanium matrix composite material and a carborundum steel matrix composite material. Suitable aluminum boron carbide metal matrix composites are sold under the name Metamic® and Boralyn®. The use of an aluminum-based metal matrix composite ensures that the DFC 100 will have good heat rejection capabilities. The boron carbide aluminum matrix composite material of which the elongated tubular wall 10 is constructed, in one embodiment, comprises a sufficient amount of boron carbide so that the elongated tubular wall 10 can effectively absorb neutron radiation emitted from the damage nuclear fuel loaded within the container cavity 101, thereby shielding adjacent nuclear fuel (damaged or intact) in the fuel basket 400 from one another (FIG. 10). In one embodiment, the elongated tubular wall 10 is constructed of an aluminum boron carbide metal matrix composite material that is greater than 25% by volume boron carbide. In other embodiments, the elongated tubular wall 10 is constructed of an aluminum boron carbide metal matrix composite material that is between 20% to 40% by volume boron carbide, and more preferably between 30% to 35%. Of course, the invention is not so limited and other percentages may be used. The exact percentage of neutron absorbing particulate reinforcement required to be in the metal matrix composite material will depend on a number of factors, including the thickness of the elongated tubular wall 10, the spacing/pitch between adjacent cells within the fuel basket 400 (FIG. 10), and the radiation levels of the damaged nuclear fuel. As will be discussed in greater detail below, the elongated tubular wall 10 is formed by an extrusion process in certain embodiments and, thus, the DFC 100 can be considered an extruded tubular container in such embodiments. Extrusion is preferred because it results in an elongated tubular wall 10 that is free of bending or warping that can be caused by welding processes that are used to create tubes. The elongated tubular wall 10 extends along the container axis C-C from a top end 11 to a bottom end 12. The top end 11 terminates in a top edge 13 while the bottom end 12 terminates in a bottom edge 14. The elongated tubular wall 10 also comprises an outer surface 15 and an inner surface 16 that forms a container cavity 101. The top edge 13 defines a top opening 17 that leads into the container cavity 101. The elongated tubular wall 10 comprises a top portion 18 and a bottom portion 19. In the exemplified embodiment, the bottom portion 19 extends from the bottom edge 14 to a transition shoulder 21 while the top portion 18 extends from the transition shoulder 21 to the top edge 13. The top portion 19, in the exemplified embodiment, is an upper section of the elongated tubular wall 10 that flares slightly outward moving from the transition shoulder 21 to the top edge 13. Thought of another way, the top portion 19 of the elongated tubular wall 10 has a transverse cross-section that gradually increases in size moving from the transition shoulder 21 to the top edge 13. The bottom portion 18, in the exemplified embodiment, has a substantially constant transverse cross-section along its length, namely from the bottom edge 14 to the transition shoulder 21. In other embodiments, the top portion 19 can also have a transverse cross-section that is substantially constant along its length from the transition shoulder 21 to the top edge 13. In such an embodiment, the transverse cross-section of the top portion can be larger than the transverse cross-section of the bottom portion 18. In still other embodiments, the elongated tubular wall 10 may have a substantially constant transverse cross-section along its entire length from the bottom edge 14 to the top edge 13. In such an embodiment, the elongated tubular wall 10 will be devoid of a transition shoulder 21 and the top and bottom portions 18, 19 would have no physical distinction. In the exemplified embodiment, the elongated tubular wall 10 has a substantially constant thickness along its entire length. In one embodiment, the elongated tubular wall 10 has a wall thickness between 1 mm to 3 mm, with about 2 mm being preferred. Of course, the invention is not so limited and the elongated tubular wall 10 can have wall thickness that is variable and of different empirical values and ranges. The inner surface 16 of the elongated tubular wall 10 defines the container cavity 101. In the exemplified embodiment, the portion of the container cavity 101 defined by the bottom portion 18 has a transverse cross-section that is substantially constant in size while the portion of the container cavity 101 defined by the top portion 19 has a transverse cross-section that increases in size moving from the transition shoulder 21 to the top edge 13. In the exemplified embodiment, the elongated tubular wall 10 has a transverse cross-section that is substantially rectangular in shape along its entire length from the bottom edge 14 to the top edge 13. Similarly, the container cavity 101 also has a transverse cross-section that is substantially rectangular in shape along its entire length. Of course, the transverse cross-sections can be other shapes in other embodiments, and can even be dissimilar shapes between the top and bottom portions 18, 19. The bottom cap 20 is fixedly coupled to the bottom end 12 of the elongated tubular wall 10 while the top cap 30 is detachably coupled to the top end 11 of the elongated tubular wall 10. More specifically, the bottom cap 20 is coupled to the bottom edge 14 of the elongated tubular wall 10. As will be described in greater detail below, in the exemplified embodiment, the bottom cap 20 is fixedly coupled to the bottom end 12 of the elongated tubular wall 10 by an autogenous welding technique, such as by friction stir welding. In other embodiments, the bottom cap 20 is fixedly coupled to the bottom end 12 of the elongated tubular wall 10 using other connection techniques. The bottom cap 20, in certain embodiments, is formed of a material comprising a metal that is metallurgically compatible with the metal of the elongated tubular wall 10 for welding. In one embodiment, the bottom cap is formed of aluminum. The bottom cap 20, in a preferred embodiment, is formed by a casting process. The bottom cap 20 comprises a plurality of first screens 22. Each of the first screens 22 comprises a plurality of openings 23 that define lower vent passageways into a bottom 102 of the container cavity 101. While in the exemplified embodiment the first screens 22 are incorporated into the bottom cap 20, the first screens 22 can be incorporated into the bottom end 12 of the elongated tubular wall 10 in other embodiments. Furthermore, while the exemplified DFC 100 comprises four first screens in the exemplified embodiment, more or less first screens 22 can be included in other embodiments. In one embodiment, the openings 23 of the first screens 22 are small enough so that radioactive particulate matter cannot pass therethrough but are provided in sufficient density (number of openings/area) to allow sufficient venting of air, gas or other fluids through the container cavity 101. In one embodiment, the openings 23 have a diameter in a range of 0.03 mm to 0.1 mm, and more preferably a diameter of about 0.04 mm. The openings 23 may be provided for each of the first screens 22, in certain embodiments, to have a density of 200 to 300 holes per square inch. The invention, however, is not limited to any specific dimensions or hole density unless specifically claimed. In the exemplified embodiment, the first screens 22 are integrally formed into a body 24 of the bottom cap 20 by creating the openings 23 directly into the body 24 of the bottom cap 20. The openings 23 can be formed into the body 24 of the bottom cap 20 by punching, drilling or laser cutting techniques. In one embodiment, it is preferred to form the openings using a laser cutting technique. Laser cutting allows very fine openings 23 to be formed with precision and efficiency. In alternate embodiments, the openings of the first screens 22 may not be integrally formed into the bottom cap 20 (or the elongated tubular wall 10). Rather, larger through holes can be formed in the bottom cap 20 that are then covered by separate first screens 22, such as wire mesh screens. Referring now to FIGS. 2 and 5 concurrently, the bottom cap 20 generally comprises a floor plate 25 and an oblique wall 26 extending upward from a perimeter of the floor plate 25. In the exemplified embodiment, the oblique wall 26 is integrally formed with the floor plate 25, for example, during the casting formation process. The oblique wall 26 is a rectangular annular wall that forms a tapered end of the DFC 100, which helps with inserting the DFC 100 into a cell 403B of the fuel basket 400 (FIGS. 10 and 11). The oblique wall 26 extends oblique to the container axis C-C and terminates in an upper edge 27. The upper edge 27 of the oblique wall 26 is coupled to the bottom edge 14 of the elongated tubular wall 10 by an autogenous butt weld 29 that seals the interface and integrally couples the components together so as to produce a junction that is smooth with the outer surface 15. The floor plate 25 comprises a top surface 28 that forms a floor of the container cavity 101. As can be seen in FIG. 5, one of the first screens 22 is located on each of the four sections of the oblique wall 26, which collectively form its rectangular transverse cross-sectional shape. The oblique wall 26 is an upstanding portion of the DFC 100. By locating the first screens 22 on an upstanding portion of the DFC 100 (rather than a portion that only has a horizontal component, such as the floor plate 25), the openings 23 of the first screens 23 are less susceptible to becoming clogged from particulate matter from the damaged nuclear fuel. Moreover, the openings 23 do not become choked-off (i.e., blocked) when the DFC 100 is supported upright in a fuel basket 400 and the floor plate 25 is in surface contact with a floor 505 of the vessel 500 (FIG. 12). In certain embodiments, an additional first screen 22 may be added to the floor plate 25 of the bottom cap 20 to ensure adequate leakage of retained water. The openings 23 of each of the first screens 22 comprise a lowermost opening(s) 23A and an uppermost opening(s) 23C. The lowermost opening 23A is located a first axial distance d1 above the floor 28 of the container cavity 101 while the uppermost most opening 23C is located a second distance d2 above the floor 28 of the container cavity 101. The second distance d2 is greater than the first distance d1. As discussed below, the DFC 100, in certain embodiments, is intended to be oriented so that the container axis C-C is substantially vertical when the DFC 100 is positioned within the fuel basket 400 of the vessel 500 for transport and/or storage. Thus, in the exemplified embodiment, both the lowermost and uppermost openings 23A, C are located a vertical distance above the floor 28 of the container cavity 101. As a result, the first screens 22 are unlikely to become clogged by settling particulate debris as each of d1 and d2 are vertical distances. As mentioned above, it is beneficial to have the first screens 22 located on an upstanding portion of the DFC 100, which in the exemplified embodiment is the oblique wall 26 of the bottom cap 20. In other embodiments, the bottom cap 20 is designed so that the wall 26 is not oblique to the container axis C-C but rather substantially parallel thereto. In such and embodiment, the first screens 22 are located on this vertical annular wall of the bottom cap 20. In still another embodiment, the bottom cap 20 may simply be a floor plate without any significant upstanding potion. In such an embodiment, the first screens 22 can be located on the bottom end 12 of the elongated tubular wall 10 itself, which would be considered the upstanding portion that is substantially parallel to container axis C-C. Of course, in such embodiments, the upstanding portion of the elongated tubular wall 10 on which the first screens 22 are located can be oriented oblique to the container axis C-C. Referring now to FIGS. 3-4 and 6 concurrently, the details of the top cap 30, along with its detachable coupling to the elongated tubular body 10 will be discussed in greater detail. The top cap 30 is shaped to provide a strong attachment location for lifting the loaded DFC 100. A handle 31 is fixedly coupled to the top cap 30 and extends upward from a top surface 32 of the top cap 30 so that the DFC 100 can be easily handled by a crane or other handling equipment. As can be seen, when the top cap 30 is detachably coupled to the elongated tubular wall 10 (shown in FIGS. 3-4), the entirety of the top cap 30 is disposed within the top portion 19 of the elongated tubular wall 10. A portion of the handle 31, however, protrudes axially from the top edge 13 of the elongated tubular wall 13. Nonetheless, the entirety of the handle 31 is located fully within a transverse perimeter defined by the top edge 13 of the elongated tubular wall 10 (viewed from a plane that is substantially perpendicular to the container axis C-C). As a result, the handle 31 can be easily grabbed by lifting mechanisms when the DFC 100 is fully inserted into a fuel cell of a fuel rack, without the grid 401 of the fuel basket 400 interfering with the lifting mechanism (FIGS. 10 and 11). The top cap 30 comprises a body 33. In one embodiment, the body 33 is formed of any of the materials described above for the elongated tubular wall 10. In another embodiment, the body 33 is formed of any of the materials described above for the bottom cap 20. The top cap 30 has a bottom surface 34, a top surface 32 and a peripheral sidewall 35. The peripheral sidewall 35 comprises a chamfered portion 36 at a lower edge thereof to facilitate insertion into the top opening 17 of the elongated tubular wall 10. The top cap 30 has a transverse cross-sectional shape that is the same as the transverse cross-sectional shape of the container cavity 101. A plurality of locking elements 37 protrude from the peripheral sidewall 35 of the top cap 30 and, as discussed in greater below, are alterable between a fully extended state (shown in FIGS. 3-4 and 6) and a fully retracted state (shown in FIG. 9) to facilitate repetitive coupling and uncoupling of the top cap 30 to the elongated tubular wall 10. In the exemplified embodiment, the locking elements 37 are spring-loaded pins. In other embodiments, the locking elements 37 can be tabs, protuberances, clamps, tangs, and other known mechanisms for locking components together The top cap 30 also comprises a second screen 38. The second screen 38 comprises a plurality of openings 39 that define upper vent passageways into a top 103 of the container cavity 101. While in the exemplified embodiment the second screen 38 is incorporated into the top cap 30, the second screen 38 can be incorporated into the elongated tubular wall 10 at a position below where the top cap 30 couples to the elongated tubular wall 10 in other embodiments. In one embodiment, the openings 39 of the top cap are small enough so that radioactive particulate matter cannot pass therethrough but are provided in sufficient hole density (number of openings/area) to allow sufficient venting of air and gases (or other fluids) through the container cavity 101. In one embodiment, the openings 39 have a diameter in a range of 0.03 mm to 0.1 mm, and more preferably a diameter of about 0.04 mm. The openings 39 may be provided for the second screen 38, in certain embodiments, to have a density of 200 to 300 holes per square inch. The invention, however, is not limited to any specific dimensions or hole density of the openings 39 unless specifically claimed. In the exemplified embodiment, the second screen 38 is integrally formed into the body 33 of the top cap 30 by creating the openings 39 directly into the body 33 of the bottom cap 20. The openings 39 can be formed into the body 33 of the top cap 30 by punching, drilling or laser cutting techniques. In one embodiment, it is preferred to form the openings 39 using a laser cutting technique. Laser cutting allows very fine openings 39 to be formed with precision and efficiency. In alternate embodiments, the openings 39 of the second screen 38 may not be integrally formed into the top cap 30 (or the elongated tubular wall 10). Rather, larger through holes can be formed in the top cap 30 that are then covered by a separate second screen(s), such as a wire mesh screen(s). Referring now to FIGS. 7-9, additional details of the locking elements 37 of the top cap 30, and the coupling of the top cap 30 to the elongated tubular wall 10, will be described. As mentioned above, the locking elements 37 are alterable between a fully extended state (FIG. 7) and a fully retracted state (FIG. 9). Referring solely now to FIG. 7, each of the locking elements 37 is biased into the fully extended state by a resilient element 40, which in the exemplified embodiment is a coil spring that is fitted over a shaft portion 41 of the locking element 37. In the exemplified embodiment, the springs 40 bias the locking elements 37 into the extended state through contact with a first wall 43 of the top cap 30 on one end and a flange 44 of the shaft portion 41 of the locking element 37 on the other end. Overextension of the locking elements 37 out of the peripheral sidewall 35 is prevented by contact interference between the flanges 44 of the shaft portions 41 and second walls 45 of the top cap. Upon the application of adequate force to the locking elements 37, the spring force of the springs 40 is overcome and each of the locking elements 37 will translate along its locking element axis L-L (FIG. 4) to the fully retracted state. In the exemplified embodiment, the locking element axes L-L are substantially perpendicular to the container axis C-C. In certain embodiments, the internal chambers 45 in which the springs 40 and portions of the locking elements 37 nest are hermetically sealed. This can be accomplished by incorporating a suitable gasket about the shaft portion 41 of the locking element at the peripheral sidewall 35. In the exemplified embodiment, a locking element 37 is provided on each one of the four sections of the peripheral sidewall 35 and are centrally located thereon at the cardinal points. As described in greater detail below, the locking elements 37 are forced from the fully extended state to the fully retracted state due to contact between the extruded tubular wall 10 and the locking elements 37 during insertion of the top cap 30 into the container cavity 101. As can be seen in FIG. 7, the portion of the container cavity 101 defined by the top portion 19 of extruded tubular wall has a transverse cross-section that gradually tapers (i.e. decreases in size) moving away (i.e., downward in the illustration) from a top edge 13 of the elongated tubular wall 10. Thus, the container cavity 101 has a transverse cross-section A1 at the top opening 17 that is greater than the transverse cross-section A2 of the container cavity 101 at an axial position immediately above locking apertures 50 formed into the elongated tubular wall 10. As mentioned above, the locking elements 37 are biased into a fully extended state and, thus, protrude from all four sections of the peripheral sidewall 35. As a result of the protruding locking elements 37, the top cap 37 has an effective transverse cross-section A3 when the locking elements 37 are in the fully extended state. The DFC 100 is designed, in the exemplified embodiment, so that the effective transverse cross-section A3 of the top cap 30 is the same as or smaller than the transverse cross-section A1 of the top opening 17 of the internal cavity 101. The effective transverse cross-section A3 of the top cap 30, however, is greater than the transverse cross-section A2 of the container cavity 101 at the axial position immediately above locking apertures 50. Referring now to FIG. 8, as a result of the relative dimensions described immediately above, when the top cap 30 is initially aligned with and lowered into the top opening 17 of the container cavity 101, the top cap 70 (including the locking elements 70) can pass through the top opening 17 while the locking elements 37 remain in the fully extended state. Thought of another way, the top edge 13 defines the top opening 17 so as to have a transverse cross-section through which the top cap 30 can be inserted while the locking elements 37 are in the fully extended state. As the top cap 30 continues to be inserted (i.e., lowered in the illustration), the locking elements 37 come into contact with the inner surface 16 of the top portion 19 of the elongated tubular wall 10 that defines that portion of the container cavity 101. Due to the fact that the inner surface 16 is sloped such that the transverse cross-section of the container cavity 101 continues to decrease with distance from the top edge 13, the locking elements 37 are further forced into retraction by the inner surface 16 of the elongated tubular wall 10 until a fully retracted state is achieved at the axial position immediately above locking apertures 50 (FIG. 9). Referring to FIG. 9, the locking elements 37 are at the axial position immediately above locking apertures 50 of the elongated tubular wall 10 and are in the fully retracted state. In the fully retracted state, the springs 40 are fully compressed and the locking elements 37 have been translated inward along the locking element axis L-L. As lowering of the top cap 30 is continued, the locking elements 37 become aligned with the locking apertures 50 of the elongated tubular wall 10 and are automatically returned back into the fully extended state in which the locking elements 37 protrude into the locking apertures 50 due to the bias of the springs 40 (shown in FIG. 4). As a result of the locking elements 37 protruding into the locking apertures 50, the top cap 30 is coupled to the elongated tubular wall 10 so that the DFC 100 can be lifted by the handle 31. The locking elements 37 cannot be forced back into the retracted state due to contact with the edges that define the locking apertures 50. In other words, once the top cap 30 is coupled to the elongated tubular wall 10 as described above, the locking elements 37 cannot be retracted by applying a lifting or pulling force (i.e. an axial force along the container axis C-C) to the top cap 30. Thus, a secure connection between the top cap 30 and the elongate tubular wall 10 is provided. In order to remove the top cap 30 from the elongated tubular wall 10, a tool is required to unlock the top cap 30 from the elongated tubular wall 10 by pressing the locking elements 37 radially inward along their locking element axes L-L. In the exemplified embodiment, the locking apertures 50 are through-holes and, thus, the locking elements 37 can be pressed inward by the access provided to the locking elements 37 by the locking apertures 50. The exemplified embodiment is only one structural implementation in which the top cap 30 and the elongated tubular wall 10 are configured so that upon the top cap 30 being inserted through the top opening 17, contact between the locking elements 37 and the elongated tubular wall 10 forces the locking elements 37 into a retracted state. In other embodiments, the effective transverse cross-section A3 of the top cap 30 may be larger than the transverse cross-section A1 of the top opening 17 of the internal cavity 101. In such an embodiment, the lower edges of the locking elements 37 can be appropriately chamfered and/or rounded so that upon coming into contact with the top edge 13 of the elongated tubular wall 10 during lowering, contact between the lower edges of the locking elements 37 and the top edge 13 of the elongated tubular wall 10 forces the locking elements 37 to translate inward along their locking element axes L-L. In other embodiments, the top edge 13 of the elongated tubular wall 10 may be appropriately chamfered to achieve the desired translation of the locking elements 37. Referring now to FIGS. 10-12 concurrently, a system 999 for storing and/or transporting damaged nuclear fuel is illustrated according to an embodiment of the present invention. The system 999 generally comprises a vessel 500, a fuel basket 400 and at least one of the DFCs 100 described above. The vessel 500, when fully assembled, forms a fluid-tight vessel cavity 501 in which the fuel basket 400, the DFC 100 containing damaged nuclear fuel and intact nuclear fuel 50 are housed (in FIG. 10, the loaded DFC 100 and the intact nuclear fuel 50 are schematically illustrated for simplicity). Thus, the vessel 500 can be considered a pressure vessel that forms a fluidic containment boundary about the vessel cavity 501. In the exemplified embodiment, the vessel 500 is a canister, such as a multi-purpose canister. In embodiments, where the vessel is an MPC, the system 100 may also comprises an overpack cask, such as an above-ground or below-ground ventilated vertical overpack. In other embodiments, the vessel 500 may be a metal cask. The vessel 500 comprises a cylindrical shell 502, a lid plate 503 and a floor plate 504. The lid plate 503 and the floor plate 504 are seal welded to the cylindrical shell 502 so to form the hermetically sealed vessel cavity 501. A top surface 505 of the floor plate 504 forms a floor of the vessel cavity 501. The vessel 500 extends along a vessel axis V-V, which is arranged substantially vertical during normal operation and handling procedures. The fuel basket 400 is positioned within the vessel cavity 502 and comprises a gridwork 401 forming a plurality of elongated cells 403A-B. In the exemplified embodiment, the gridwork 401 is formed by a plurality of intersecting plates 402 that form the cells 403A-B. In one embodiment, the plates 402 that form the gridwork 401 are formed of stainless steel. Because the elongated tubular wall 10 of the DFC 100 is made of a boron carbide aluminum matrix composite material, or a boron aluminum matrix composite material, and the gridwork 401 is made of stainless steel, there is no risk of binding from the cohesion effect of materials of identical genre. Each of the elongated cells 403A-B extend along a cell axis B-B that is substantially parallel to the vessel axis V-V. The plurality of cells 403A-B comprises a first group of cells 403A that are configured to receive intact nuclear fuel 50 and a second group of cells 403B configured to receive DFCs 100 containing damage nuclear fuel. Each of the cells 403A of the first group comprise neutron absorbing liner panels 404 while the each of the cells 403B of the second group are free of the neutron absorbing liner panels 404. In one embodiment, the neutron absorbing liner panels 404 can be constructed of the same material that is described above for the elongated tubular wall 10. Because the elongated tubular wall 10 of the DFC 100 incorporate neutron absorber as described above, the cells 403B of the fuel basket 400 that are to receive the DFCs 100 do not require such neutron absorber plates 404, leading to an increased cell cavity size which is large enough to enable free insertion or extraction of the DFC 100 from the fuel basket 400. In certain embodiments, the cell opening of the cells 403B is 6.24 inches, which means that there is a ¼ inch lateral gap between the DFC 100 and the grid that forms the storage cell 403B. Moreover, because the DFC 100 is extruded and the cells 403A-B of the fuel basket 400 are of honeycomb construction made of thick plate stock (¼ inch wall), there is a high level of confidence that the DFCs 100 can be inserted into the storage cells 403B without interference. In the exemplified embodiment, all of the cells 403A-B have the same pitch therebetween. Referring now to FIGS. 11 and 12, each of the DFCs 100 is loaded into one of the cells 403B by aligning the DFC 100 with the cell 403B and lowering the DFC 100 therein until the floor plate 25 of the DFC 100 comes into surface contact with and rests on the top surface 505 of the floor plate 504 of the vessel 500. When positioned within the cell 403B, the container axis C-C of the DFC 100 is substantially parallel to the cell axis B-B and, in certain embodiments, substantially coaxial therewith. As mentioned above, the cell axis B-B is substantially parallel to the vessel axis V-V. Thus, when the DFC 100 is loaded within the cell 403B, the oblique wall 26 of the bottom cap 20 is oblique to both the cell axis B-B and the vessel axis V-V. As mentioned above, the top surface 505 of the floor plate 504 forms a floor of the vessel cavity 501. Thus, when the DFC 100 is loaded within the cell 403B, the lowermost opening(s) 23A of the first vent(s) 22 is a distance d3 above the floor 505 of the vessel 500 while the uppermost opening(s) 23C of the first vent(s) 22 is a distance d4 above the floor 505 of the vessel 500. In summary, the DFC 100 of the present invention fits in the storage cell 403B with adequate clearance. The DFC 100 also provides adequate neutron absorption to meet regulatory requirements. The DFC 100 also confines the particulates but allow water and gases to escape freely. The DFC 100 also features a robust means for handling and includes a smooth external surface to mitigate the risk of hang up during insertion in or removal from the storage cell 403 B. The DFC also provides minimal resistance to the transmission of heat from the contained damaged nuclear fuel. The loaded DFC 100 can be handled by a grapple from the Fuel Handling Bridge. All lifting appurtenances are designed to meet ANSI 14.6 requirements with respect to margin of safety in load handling. Specifically, the maximum primary stress in any part of the DFC 100 will be less than its Yield Strength at 6 times the dead weight of the loaded DFC,W. and less than the Ultimate Strength at 10 times W. The table below provides design data for one embodiment of the DFC 100. DFC: Design DataOuter Dimension152 mm(5.99″)Corner Radius6 mm(0.24″ nominal)Wall Thickness2.0 mm(0.079″)DFC Cell I.D.148 mm(5.83″)Total Height4680 mm(184.25″)Boron Carbide Concentration32%(nominal)Empty Weight, Kg25(55 lbs)Permissible Planar Average Enrichment4.8% A method of manufacturing the DFC 100 according to an embodiment of the present invention will now be described. First, the elongated tubular wall 10 is formed via an extrusion process using a metal matrix composite having neutron absorbing particulate reinforcement. A boron carbide aluminum matrix composite material is preferred. At this stage, the extruded elongated tubular wall 10 (and the container cavity 101) has a substantially constant transverse cross-section, with the elongated tubular wall 10 also having a substantially uniform wall thickness. The elongated tubular wall 10 is then taken and a portion thereof is expanded so that the container cavity 101 has an increased transverse cross-section, thereby forming the top portion 19 and the bottom portion 18 elongated tubular wall 10. Expansion of the container cavity 101 (which can also be considered expansion of the elongated tubular wall 10) can be accomplished using a swaging process using an appropriate mandrel, die and/or press. Said swaging process can be a hot work in certain embodiments. In an alternate embodiment, the difference sizes in transverse cross-section of the container cavity 101 can be accomplished by performing a drawing process to reduce the bottom portion 18 of the elongate tubular wall 10. The locking apertures 50 are then formed into the top portion of the elongated tubular wall 10 via a punching, drilling, or laser cutting technique. The bottom cap 20 is then formed. Specifically, the bottom cap 20 is formed by casting aluminum to form the cap body 24. The plurality of openings 23 are then integrally formed therein using a laser cutting process to form the first screens 22 on the oblique wall 26. The bottom cap 20 is then autogenously welded to the bottom end 12 of the elongated tubular wall 10. More specifically, the bottom cap 20 is butt welded to the bottom end 12 of the elongated tubular wall 10 to produce a weld junction that is smooth with the outer surface 15 of the elongated tubular wall 10. A friction stir weld technique may be used. The top cap 30 is then formed and coupled to the elongated tubular wall 10 as described above. II. Inventive Concept 2 With reference to FIGS. 13-20, a second inventive concept will be described. FIG. 13 shows a fuel rack 2101 having an array of cells 2103 into which spent nuclear fuel assemblies may be inserted. The fuel rack 2101 may be part of a submerged storage system for spent nuclear fuel, or it may be part of a transportation system for spent nuclear fuel, such as dry or wet spent fuel casks. As shown, the cell walls include a feature for interlocking with a locking protuberance included as part of a neutron absorbing assembly inserted into one or more of the cells. This feature may be a complementary locking protuberance, or a complementary receptacle to receive the locking protuberance of the neutron absorbing assembly. The feature may be created by bending, punching, welding, riveting, or otherwise permanently deforming the cell walls of the rack or the fuel cask, or by securing attachments to the cell walls, for holding the absorption assembly in place once it is inserted into the fuel cell. In some embodiments, if the fuel rack 2101 has too small of a cell opening to accommodate thickness of the fuel insert, the insert may be directly inserted into the guide tubes of the fuel assembly. FIGS. 13-16 show a neutron absorbing assembly 2111 which may be used in conjunction with both PWR or BWR storage requirements. The neutron absorbing assembly 2111 is configured to be slidably inserted at strategic locations within the cell array of a submerged fuel rack. However, the absorbing assembly can be used in any environment (and in conjunction with any other equipment) where neutron absorption is desirable. Furthermore, based on the disclosed process for bending a metal matrix composite having neutron absorbing particulate reinforcement (or the resulting angled plate structure), an absorbing assembly may be configured for use in any environment and/or used to create a wide variety of structures, including without limitation fuel baskets, fuel racks, sleeves, fuels tubes, housing structures, etc. The neutron absorbing assembly 2111 includes a corner spine 2113, to which are fastened two walls 2115 to form a chevron-shaped structure (when viewed from the top or bottom). For a cell with a square cross-sectional configuration, the corner spine 2113 creates a relative angle between the two walls 2115 of about 90 degrees. Other relative angles may also be used, primarily depending upon the cross-sectional configuration of the cell into which the neutron absorbing assembly 2111 is to be inserted (e.g., triangular, pentagonal, hexagonal, etc.). Each wall has an absorption sheet 2117, constructed from a neutron absorbing material, and a guide sheet 2119. Since the walls may be mirror images of each other, the following addresses the configuration of only one of the walls, with the understanding that the second wall may be similarly configured. However, in one embodiment, one of the walls includes a locking feature, and one does not. In other embodiments, both walls include a locking feature. In certain embodiments, additional corner spines and walls may be added to provide neutron absorption on more than two sides of a cell. The absorption sheet 2117 is affixed to and extends much the length of the corner spine 2113, and it may extend the entire length or only part of the length, depending upon the requirements for neutron absorption within the cell, e.g., the linear space within the cell occupied by the spent fuel rods. The absorption sheet 2117 may be affixed to the corner spine 2113 using any suitable fastener, such as rivets. The bottom edge 2118 of the absorption sheet 2117 has a skewed shape to facilitate ease of insertion of the neutron absorbing assembly 2111 into a cell of a fuel rack. Specifically, the bottom edge 2118 of the absorption sheet 2117 taper upward and away from the corner spine 2113. The guide sheet 2119 is affixed to only a top portion of the absorption sheet 2117 by suitable fasteners, such as rivets, and the guide sheet 2119 extends along less of a length of the corner spine 2113 than the absorption sheet 2117. The edge of the guide sheet 2119 abuts up against the edge of the corner spine 2113 along a common edge 2121 to help reduce the overall thickness of the assembly. As shown in FIG. 13, the absorption sheet extends along most of the length of the corner spine 2113, and the guide sheet 2113 extends along a short top portion of the corner spine 2113. The difference in lengths reflects the difference in functions between the absorption sheet 2117 and the guide sheet 2119. Where the absorption sheet 2117 is included for neutron absorption, the guide sheet 2119 is included, at least to aid in guiding a spent nuclear fuel assembly into the cell after the absorption assembly 2111 is in place within the cell, to protect the top edge of the absorption assembly from damage, to provide a support surface for a locking protuberance, and to provide a structure by which the absorption assembly 2111 may be supported during installation into the cell. The guide sheet also includes an extension portion 2123 which extends over and above the top edge 2125 of the absorption sheet 2117. This extension portion 2123 provides a surface to aid in guiding a spent fuel assembly into a cell in which the absorption assembly is 2111 placed. The extension portion 2123 also protects the top edge 2125 of the absorption sheet 2117 from damage during the process of loading a spent fuel assembly into the cell. The top portion of each absorption sheet 2117 includes a cut-out 2125, and a tab 2127 (which is a locking protuberance in the embodiment shown) extends from the guide sheet 2119, through the cut-out 2125, and beyond the outer surface of the absorption sheet 2117. The tab 2127 includes a lower part 2129 affixed to the guide sheet, using any suitable fastener, such as rivets, and an upper part 2131 which is bent away from the guide sheet 2119 to extend through the cut-out 2125. A locking protuberance may be formed in any other manner to provide the same locking functionality as described in connection with the tab herein. In addition, a locking protuberance may be included on both the absorption assembly 2111 and the cell wall (See FIG. 18), or in other embodiments it may be included on only one of the absorption assembly 2111 and the cell wall. As seen in FIG. 17, one suspension aperture 2135 is included at the top of the corner spine 2113, and one suspension aperture 2137 is included in the extension portion 2123 of each guide sheet 2119. These suspension apertures 2135, 2137 are included to facilitate robotically placing the absorption assembly 2111 in a cell within a submerged storage system. The shape and positioning of the suspension apertures is a matter of design choice. A single cell 2151 for receiving a spent nuclear fuel assembly and an absorption assembly is shown in FIG. 18. Two walls of the cell 2151 each include a feature 2153 near the top of the cell wall 2155, and the feature 2153 is configured to engage the absorption assembly to retain the absorption assembly when the spent nuclear fuel assembly is removed from the cell. This feature 2153 may be an indentation, a cut-out, or a protuberance, depending upon what type of corresponding locking feature is included on the absorption assembly. The type of feature and its configuration are a matter of design choice. A detailed cross-sectional view of the locking features of the absorption assembly 2111 and the cell 2151 are shown in FIG. 19. As described above, the locking feature may be a tab, and such a tab 2127 is shown with its top portion 2131 in locking engagement with a second tab 2161, this second tab 2161 being formed in the cell wall 2155. When manufacturing the absorption assembly for a fuel rack that has not yet been placed in service, the order of making the locking protuberances, the type of locking protuberance used, and even whether one or both of the cell wall and the absorption assembly include a locking protuberance, are anticipated to be variables that may be addressed by design decisions for a particular configuration. However, when retrofitting a fuel rack or cask that is already in use, and a tab is used in the cell wall as a locking protuberance, preferably the absorption assembly is first manufactured and placed into the cell before the tab in the cell wall is created. This permits maximization of space use within a pool or cask by minimizing the space requirements of the absorption assembly, because the tab effectively reduces the overall nominal width of the cell. When retrofitting an existing and in-use fuel rack or cask, the tab 2161 in the cell wall may be formed just above the position of the tab in the absorption assembly as a half-shear using a C-shaped tool which spans the extension portion 2123 of the guide sheet 2119. With such a tool, a double-acting hydraulic cylinder may be used to push a wedge-shaped piece of the tool into the cell wall, thereby creating the half-sheared tab 2161 extending toward the inner space of the cell. The cell 2151 has an overall length L, and the corner spine is configured to have approximately the same length, as shown in FIG. 20. As shown, the corner spine 2113 clears the top 2157 of the cell wall 2159 by a sufficient amount to make the suspension aperture 2135 of the corner spine 2113 accessible, even when the spent nuclear fuel assembly 2159 is placed within the cell 2151. The length of the corner spine 2113 is such that the bottom edge 2162 rests against the bottom 2163 of the cell 2151. The absorption sheet 2117 need not extend all the way to the bottom 2163 of the cell 2151, as the length of the absorption sheet 2117 may extend as far down into the cell as needed so that it shields adjacent fuel assemblies from one another. This is because adjacent spent nuclear fuel rods may not extend the entire length of the cell either, and the length of the absorption sheet 2117 need only be as long as the spent nuclear fuel rods within the spent nuclear fuel assembly 2159, although they may be longer if desired. Since there is a need to maximize space use within a fuel pond or cask, it is desirable that the absorption assembly 2111 take up as little room as possible in the cell of the fuel rack. To this end, the absorption sheets 2117 are preferably constructed of an aluminum boron carbide metal matrix composite material having a percentage of boron carbide greater than 25%. While the addition of boron carbide particles to the aluminum matrix alloy increases the ultimate tensile strength, increases yield strength, and dramatically improves the modulus of elasticity (stiffness) of the material, it also results in a decrease in the ductility and fracture toughness of the material compared to monolithic aluminum alloys. The boron carbide aluminum matrix composite material of which the absorption sheets are constructed includes a sufficient amount of boron carbide so that the absorption sheets can effectively absorb neutron radiation emitted from a spent fuel assembly, and thereby shield adjacent spent fuel assemblies in a fuel rack from one another. The absorption sheets may be constructed of an aluminum boron carbide metal matrix composite material that is about 20% to about 40% by volume boron carbide. Of course, other percentages may also be used. The exact percentage of neutron absorbing particulate reinforcement which is in the metal matrix composite material, in order to make an effective neutron absorber for an intended application, will depend on a number of factors, including the thickness (i.e., gauge) of the absorption sheets 2107, the spacing between adjacent cells within the fuel rack, and the radiation levels of the spent fuel assemblies. Other metal matrix composites having neutron absorbing particulate reinforcement may also be used. Examples of such materials include, without limitation, stainless steel boron carbide metal matrix composite. Of course, other metals, neutron absorbing particulate and combinations thereof may be used including without limitation titanium (metal) and carborundum (neutron absorbing particulate). Suitable aluminum boron carbide metal matrix composites are sold under the trade names Metamic® and Boralyn®. The center spine, the guide sheets, and the locking protuberance may be formed from steel or other materials, or they may alternatively be formed from non-metallic materials. When the locking protuberance is configured as a tab affixed to the guide sheet of the absorption assembly, the tab is preferably formed from a sheet of 2301 stainless spring steel, tempered to about ¾ hard. In a preferred embodiment, the tab is about 0.035 inches thick, about 0.7 inches wide, and about 1.7 inches long, with the upper portion of the tab being about 1.09 inches long and bent to extend beyond the outer side of the absorption layer by between 0.125 inches to 0.254 inches, depending upon how thick the absorption layer is and whether the absorption assembly is being placed over an existing absorption layer within the cell. In the latter instance, the tab should be configured so that the upper portion extends beyond the existing absorption layer. The extent to which the tab extends beyond the absorption layer is a matter of design choice, as it depends upon several factors such as the type of locking feature included on the cell wall, how much the tab needs to deflect upon insertion, and how much removal force the tab should be able to withstand. For example, with a tab extending 0.125 inches beyond the absorption layer, it may be desirable to have the tab be able to deflect by approximately 0.124 inches upon insertion. Such a configuration is anticipated to withstand at least a 200 lb removal force once the tab is interlocked with a second tab formed in the cell wall. It should be noted that the tab will remain in a substantially deflected state once the absorption assembly is inserted into cell wall III. Inventive Concept 3 With reference to FIGS. 21-31, a third inventive concept will be described. An array of fuel storage racks 3101 is shown in FIG. 21. Each storage rack 3101 is itself an array of fuel cells 3103, and each is generally square in cross section, with each fuel cell 3103 also being square in cross section. Such storage racks, and their construction, are generally known in the art. For example, U.S. Pat. No. 4,382,060 to Holtz et al. describes a storage rack and details how each fuel cell is configured to receive and store nuclear fuel. Typically, the storage racks are used for storing nuclear fuel underwater in storage pools. Each storage rack 3101 includes a base plate 3105, which may be formed integrally as the bottom of the fuel cells 3103, or it may be coupled with an appropriate fastening system. Each base plate 3105 is disposed atop a bearing pad 3107, with a support structure (not shown in FIG. 21; See, e.g., FIG. 24) providing structural support between, and coupling together, the base plate 3105 and the bearing pad 3107. The bearing pad 3107 may, in certain instances, be considered a coupler pad in that it couples multiple fuel racks together as discussed in greater detail below. The support structure, as is further discussed below, is also constructed to allow cooling fluid (e.g., water, among other liquids) to circulate under the base plate and up through apertures in the base plate. As shown in the embodiment depicted in FIG. 21, the bearing pad 3107 may be a single sheet of material that contiguously extends under all the storage racks 3101 forming the array. When used in this configuration, the bearing pad acts to couple the various racks of the array to each other, so that each storage rack 3101 is limited in the amount of independent lateral movement with respect to both the bearing pad 3107 and each of the other storage racks 3107. By restricting the lateral movement of the individual storage racks in this manner, the bearing pad causes all the storage racks coupled thereto to move largely in unison in any direction, and significant movement of the entire coupled array occurs only when the bearing pad slides on the bottom surface of the pool. Thus, the bearing pad aids in reducing the kinematic response of individual racks under strong seismic conditions by coupling together the individual racks so that the kinematic responses of all the racks together are effectively coupled together, and the kinematic response of the some racks within the array may serve as at least a partial offset to the kinematic response of other racks within the array. In addition, while the bearing pad serves to could each storage rack in the array of storage racks together, it also enables each storage rack to effectively remain free-standing. Having free-standing storage racks in a pool is important in that each storage rack may be placed and removed individually and separately from each of the other storage racks. A top view of an array of storage racks 3111 is shown in FIG. 22. These storage racks 3111 are coupled to a bearing pad 3113 as discussed above. In this embodiment, the bearing pad 3113 extends outward from the periphery of the array of storage racks 3111. This outward extension of the bearing pad 3113 is configured to maintain a predetermined distance between the storage racks and the side of a storage pool (not shown). By maintaining the predetermined distance between the storage racks and the side of a storage pool, the array of storage racks 3111 may be prevented from moving close enough to the side of the storage pool so that an impact between one or more of the storage racks 3111 and the side wall of the storage pool is likely during a seismic incident. This predetermined distance, which is the distance the bearing pad 3113 extends beyond the outer lateral dimensions of the storage racks, may be as little as about ½ inch. Preferably, the largest outer lateral dimension of each storage rack is defined by the base plate for each storage rack. Those of skill in the art will recognize that the size of this predetermined distance may be influenced by many other factors associated with the configuration of storage racks and the configuration of the storage pool. By coupling multiple storage racks with one or more bearing pads, the movement of the freestanding racks can be significantly reduced, if not minimized, on the pool's surface under a severe earthquake. For purposes of this disclosure, a severe earthquake or seismic event is empirically defined as one in which the seismic accelerations are large enough to move a short square block of steel (i.e., a squat and rigid body) on the pool slab by at least 2 inches. By coupling storage racks together using the bearing pads, the relatively uncoordinated motion of the freestanding storage racks produced by a seismic event is exploited to dissipate dynamic energy of the various individual storage racks. During a seismic event, the fuel modules attempt to move in various different directions and thereby exert the lateral forces on the storage racks, which in turn exert lateral forces on the bearing pad(s). This leads to a reduced net resultant force, when the lateral forces of all coupled storage racks are combined. The bearing pad therefore preferably has a bottom surface which provides sufficient friction, under load, with the bottom of the storage pool. During seismic events that are less than a severe seismic event, the lateral forces generated by coupled storage tanks will generally not exceed the friction force between the loaded bearing pad and the bottom of the storage pool, wherein the load on the bearing pad has contribution from the combined vertical load of all participating pedestals. In such circumstances, the bearing pad should not slide on the bottom of the storage pool, and thus the kinematic movement of the racks will be substantially suppressed. A seismic analysis of the coupled storage rack array shown in FIG. 22 has been performed, and the under three dimensional seismic motion, the sliding response of the coupled storage rack array may be reduced by an order of magnitude as compared to the sliding response of freestanding storage racks that are not coupled by a bearing pad. FIGS. 23 and 24 illustrate an embodiment of the support structure that may be used to couple between the base plates of the storage racks and the bearing pad. For simplicity and purposes of illustration, a smaller version of a storage rack 3121 is shown in FIG. 23, having only two fuel cells 3123 per side. In addition, as an alternative embodiment, only one storage rack 3121 is placed on the bearing pad 3125. In this alternative embodiment, the bearing pad 3125 helps to maintain spacing between the storage rack 3121 and the walls of the storage pool, and between other storage racks placed on their own bearing pads that may be placed within the same storage pool. However, by placing each storage rack within a storage pool on its own individual bearing pad, much of the advantage of coupling the storage racks to help offset the kinematic response of individual storage racks may be lost. The base plate 3127 of the storage rack 3121 has multiple support pedestals 3129 affixed thereto, and these pedestals serve as the support structure between the base plate 3127 and the bearing pad 3125. The spacing between the support pedestals 3129 is provided for liquid to circulate between the base plate 3127 and the bearing pad 3125. The base plate 3127 also includes apertures 3131, which allow the cooling liquid to pass through the base plate 3127 and rise up into the fuel cells 3123. The support pedestals 3129 in this embodiment are each disposed within a recess cavity 3133 formed in the bearing pad 3125. The support pedestals 3129 and the respective recess cavities 3133 may have any desired shape which enables the support pedestals to couple with the recess cavities. Two design features for a support pedestal and/or a recess cavity are preferably included in the configuration of one or both of the paired support pedestals and the recess cavities. The first feature is the inclusion of a guide surface on one or both of the support pedestal 3129 and the recess cavity 3133. The guide surface aids in guiding one into the other when the storage rack 3121 is lowered onto the bearing pad 3125 within the storage pool. As can be seen in FIG. 24A, the support pedestal 3129 includes a rounded end 3137 to serve as a guide surface, and the recess cavity 3133 includes a beveled edge 3139 to server as a guide surface. Both the rounded end 3137 and the beveled edge 3139 aid in guiding the support pedestal 3129 into the recess cavity 3133 when the storage rack 3121 is lowered into position on the bearing pad 3125 within a storage pool, especially when every support pedestal 3129 and every recess cavity 3133 include such guide surfaces. The second feature that is included in the pairs of support pedestals and recess cavities is the lateral tolerance, t, between the maximum effective outer dimension of the support pedestal, OD, and the minimum effective inner dimension of the recess cavity, ID. FIG. 24B shows the profile 3141 of the support pedestal 3129 and the profile 3143 of the recess cavity 3133 along the line T. Since each profile 3141, 3143 is round, the maximum effective outer dimension of the support pedestal, OD, is the diameter of the support pedestal, and the minimum effective inner dimension of the recess cavity, ID, is the diameter of the recess cavity, along the line T. When this lateral tolerance, t, for each support pedestal/recess cavity pair is the same, it defines the maximum lateral distance the storage rack 3121 can move laterally independent of the bearing pad 3125. Preferably, this lateral tolerance, t, is no more than the predetermined distance that the bearing pad 3125 extends beyond the outer lateral dimensions of the storage rack, the latter being discussed above. In the case of two storage racks coupled together by a bearing pad, this lateral tolerance is preferably less than or equal to half the predetermined distance separating the base plates of adjacent storage racks. Those of skill in the art will recognize that either or both of the support pedestals and the recess cavities may have profiles that are of any desired geometrical shape that enables coupling between the base plate and the bearing pad, and allows for limited lateral movement of the storage rack with respect to the bearing pad within an established lateral tolerance. By including the lateral tolerance, t, at the point of coupling between the bearing pad and the storage rack, movement of the storage rack, independent of movement of the bearing pad, is limited by the amount of the lateral tolerance, t. Any lateral movement of the storage rack that is greater than the lateral tolerance, t, will necessarily require either movement of the bearing pad or decoupling of the storage rack from the bearing pad. Due to the weight of a fully loaded storage rack, decoupling is unlikely. A bearing pad 3151 having multiple recess cavities 3153 is illustrated in FIG. 25. This bearing pad is configured to be placed in the bottom of a storage pool and have a plurality of storage racks lowered into the pool so that each support pedestal of the storage racks couples into one of the recess cavities 3153 of the bearing pad 3151. The bearing pad 3151 may therefore have as many recess cavities as all the storage racks combined have support pedestals. The bearing pad also has a substantially flat bottom, which enables it to slide on the bottom of the pool under the loads that may be caused by a seismic event. The bottom of the bearing pad may also be coated to help control the amount of sliding that may occur. As an alternative, if the storage racks have support pedestals of different lengths extending from the base plate, then the longer support pedestals may be coupled into recess cavities, and the shorter support pedestals may extend to the top surface of the bearing pad for supporting the storage rack, but such shorter support pedestals would not couple to the bearing pad, in that they would not serve to restrict lateral movement of the storage rack during a seismic event. An alternative embodiment for the support structure between the base plate 3161 of a storage rack and a bearing pad 3163 is shown in FIG. 26. In this embodiment, the bearing pad 3163 includes upward-extending support columns 3165, and the base plate 3161 includes downward-extending receptacles 3167 to couple with each support column. The support columns include top beveled edges 3169 to act as a guide surface, and the receptacles include a lower beveled edge 3171 to similarly act as a guide surface. As should be evident from the different embodiments described, the support structure and the base plate be couple together by forming the support structure as a first engagement feature affixed to the base plate (e.g. support pedestals, receptacles) and coupling the first engagement feature to a second engagement feature formed as part of or affixed to the bearing pad (e.g., recess cavities, support columns). Thus, it should be apparent that the first and second engagement features may take on any desirable configuration, from those described above, to combinations of those described above, and to other structural configurations, with the following concepts generally taken into account: 1) providing appropriate structural support and lift to the storage rack to thereby allow circulation of cooling liquid under and up through the base plate, and 2) limiting lateral movement of the storage rack independent from the bearing pad. The first aforementioned concept allows appropriate circulation of cooling liquid, while the second concept is used to reduce the likelihood of an impact with the wall of a storage pool when the bearing pad is used with a single storage rack, and also to reduce lateral movement of an array of storage racks during a seismic event when the bearing pad couples two or more storage racks together. An array of two storage racks 3181 disposed in a storage pool 3191 is shown in FIG. 27. The two storage racks 3181 are coupled together by a single bearing pad 3183, with the base plates 3185 of the storage racks 3181 having support pedestals 3187 that extend down into recess cavities (not shown in this figure) formed in the bearing pad 3183. As an alternative, the bearing pad may be integrally formed in the bottom surface 3193 of the storage pool 3191. Each storage rack 3181 also includes a collar 3189 affixed to a top of and extending around each rack 3181, each collar 3189 forming a spacer at the top of each storage rack 3181. Each collar 3189 extends outward from the sides of the storage rack 3181 to which it is affixed, respectively, toward the collar 3189 on the other storage rack 3181, so that there is a second predetermined distance between the two collars 3189. The base plates 3185 of each storage rack 3181 extends outward from the respective storage rack 3181 further than the collar 3189, such that the predetermined distance between the two base plates 3185 is greater than the predetermined distance between the two collars 3189. Configured in this way, and considering the lateral tolerance of the support pedestals 3187 within the recess cavities, during a seismic event, the support pedestals and the recess cavities form a primary impact zone, the base plates 3185 of the adjacent storage racks 3181 form a secondary impact zone, and the collars 3189 of the adjacent storage racks form a tertiary impact zone. The spacer for each storage rack may have other configurations, and need not extend around the entire top of the storage rack. For example, the spacers may be formed as individual outcroppings affixed to the storage racks, and set so that the spacers on one storage rack are opposite the spacers on an adjacent storage rack. The purpose is to set spacers between adjacent racks so that the spacers impact each other during a seismic event instead of the fuel cells of the adjacent racks impacting. FIG. 28 shows profiles of a storage rack and the bearing pad to which it is coupled in the horizontal plane of the base plate of the base plate of the storage rack, to show the difference in sizes, although each profile of each part shown in this figure is not to scale. In the configuration shown, the bearing pad extends entirely under the storage rack. The portion of the storage rack which includes the array of cells is the storage rack profile 3201. The collar profile 3203 is shown, along with the profile of attachment points 3205 to the storage rack profile 3201. The collar profile 3203 is larger than, and extends outside of, the storage rack profile 3201. The base plate profile 3207 is shown, and it is larger than, and extends outside of, both the storage rack profile 3201 and the collar profile 3203. The bearing pad profile 3209 is larger than, and extends outside of, the base plate profile 3207. FIG. 29 shows profiles of an array of two storage racks and the associated bearing pad to which both are coupled, with the profiles being shown in the horizontal plane of the base plates of the storage racks. In this configuration, the bearing pad extends entirely under both storage racks. The portion of the storage racks which include the respective arrays of cells are the storage rack profiles 3211. The collar profiles 3213 for each storage rack are larger than the storage rack profile 3211 for each respective storage rack. Similarly, the base plate profiles 3215 for each storage rack are larger than the respective collar profiles 3213. The bearing pad profile 3217 is larger than the combined two base plate profiles 3215, extending outside of both. An alternative embodiment of a bearing pad 3221 is shown in FIGS. 30A-C. This bearing pad 3221 includes four recess cavities 3223. This bearing pad 3221 may be placed under adjacent sides of two adjacent storage racks, with two support pedestals from each storage rack being placed in the four recess cavities 3223. Alternatively, as illustrated in FIG. 30B, it may be placed under the corners of four adjacent storage racks (the outlines of the corners 3225 are shown), with one support pedestal from each of the four storage racks being placed in the four recess cavities 3223. In either of these embodiments, the support pedestals placed in the recess cavities are adjusted to be shorter than those that extend to the bottom of the storage pool and not placed in recess cavities. FIG. 31 shows profiles of an array of two storage racks and the associated bearing pads, of the type shown in FIGS. 30A-C, to which both the storage racks are coupled, with the profiles being shown in the horizontal plane of the base plates of the storage racks. The portion of the storage racks which include the respective arrays of cells are the storage rack profiles 3231. The collar profiles 3233 for each storage rack are larger than the storage rack profile 3231 for each respective storage rack. Similarly, the base plate profiles 3235 for each storage rack are larger than the respective collar profiles 3233. In this configuration, each base plate is coupled at the corners to one of four separate bearing pads, and the bearing pad profiles 3237 are shown in position with respect to the base plate profile 3235. In this configuration, even though the bearing pads are dimensionally smaller than the base plates, the smaller bearing pad profiles 3237 still extend outside of the base plate profiles 3235, and each bearing pad is also coupled to both storage racks. As should be understood from the various embodiments of the bearing pad disclosed above, the bearing pad may couple to the entire support structure of a storage rack, or it may couple to only a portion of the support structure. For example, a bearing pad may be configured to couple to just the corners of the support structure, or one may be configured to couple along an entire side of the support structure, but not the support structure nearer the middle of the storage rack. IV. Inventive Concept 4 With reference to FIGS. 32-49, a fourth inventive concept will be described. Nuclear fuel assemblies (also referred to as “bundles” in the art) each comprise a plurality of fuel pins or rods mechanically coupled together in an array which is insertable as a unit into a reactor core. The fuel assemblies traditionally have a rectilinear cross-sectional configuration such as square array and contain multiple fuel rods. A reactor core contains multiple such fuel assemblies. The fuel rods are generally cylindrical elongated metal tubular structures formed of materials such as zirconium alloy. The tubes hold a plurality of vertically-stacked cylindrical fuel pellets formed of sintered uranium dioxide. The fuel rod tubes have an external metal cladding formed of corrosion resistant material to prevent degradation of the tube and contamination of the reactor coolant water. The opposite ends of the fuel rod are sealed. FIGS. 32-40B show a damaged nuclear fuel storage system 4100 according to the present disclosure. The system includes a vertically elongated fuel rod enclosure capsule 4110 configured to hold multiple damaged fuel rods and a closure lid 4200 mounted thereto. The lid 4200 is configured for coupling and permanent sealing to the capsule 4200, as further described herein. Capsule 4110 has an elongated and substantially hollow body formed by a plurality of adjoining sidewalls 4118 defining an internal cavity 4112 that extends from a top end 4114 to a bottom end 4116 along a vertical centerline axis Cv. The bottom end 4116 of the capsule is closed by a wall. The top end 4114 of the capsule is open to allow insertion of the damaged rods therein. The sidewalls 4118 are sold in structure so that the cavity 4112 is only accessible through the open top end 4114 before the lid is secured on the capsule. In one embodiment, capsule 4110 may have a rectilinear transverse cross-sectional shape such as square which conforms to the shape of a typical fuel assembly. This allows storage of the capsule 4110 in the same type of radiation-shielded canister or cask used to store multiple spent fuel assemblies, for example without limitation a multi-purpose canister (MPC) or HI-STAR cask such as those available from Holtec International of Marlton, N.J. Such canisters or casks have an internal basket with an array of rectilinear-shaped openings for holding square-shaped fuel assemblies. It will be appreciated however that other shaped capsules 4110 may be used in other embodiments and applications. The body of the capsule 4110 may be formed of any suitable preferably corrosion resistant material for longevity and maintenance of structural integrity. In one non-limiting exemplary embodiment, the capsule 4110 may be made of stainless steel and have a nominal wall thickness of 6 mm. In certain embodiments, the capsule 4110 may further include a laterally enlarged mounting flange 4111 disposed at and adjacent to the top end 4114, as shown in FIGS. 32-34 and 7-9A. Mounting flange 4111 extends laterally outwards from the sidewalls 4118 on all sides and vertically downwards from top end 4114 along the sidewalls for a short distance. The mounting flange 4111 is configured and dimensioned to engage a mounting opening 4302 formed in a storage canister 4300, thereby supporting the entire weight of a loaded capsule 4110 in a vertically cantilevered manner as shown in FIGS. 43-45 and further describe herein. In other embodiments, different methods may be used to support the capsule 4110 in the storage canister and mounting flange 4111 may be omitted. Referring now particularly to FIGS. 34, 38, 39 and 40A, the capsule 4110 further includes an internal basket assembly configured to store and support a plurality of damaged fuel rods. The assembly includes an upper tubesheet 4120 and lower tubesheet 4122 spaced vertically apart therefrom. The upper and lower tubesheets are horizontally oriented. The lower tubesheet 4122 is separated from the interior bottom surface 4116a of bottom end 4116 of the capsule 4110 by a vertical gap to form a bottom flow plenum 4124. The upper tubesheet 4120 is spaced vertically downwards from the top end 4112 of the capsule 4110 by a distance D1 sufficient to form a top flow plenum 4126 when the closure lid 4200 is mounted on the capsule as shown in FIG. 46. Top plenum 4126 is therefore formed between the bottom 4204 of the lid 4200 and top surface 4128 of the upper tubesheet 4120. Both the bottom and top plenums 4124, 4126 are part of flow paths used in conjunction with the gas fuel rod drying/dehydration process after the capsule is closed and sealed, as further described herein. A plurality of fuel rod storage tubes 4130 are each supported by the upper and lower tubesheets 4120, 4122 for holding the damaged (i.e. broken and/or leaking) fuel rods. In certain embodiments, intermediate supporting tubesheets or other support elements (not shown) may be used to provide supplementary support and lateral stability to the storage tubes 4130 for seismic events. In one embodiment, the storage tubes 4130 each have a diameter and internal cavity 4131 with a transverse cross section configured and dimensioned to hold no more than a single fuel rod. Accordingly, the tubes 4130 extend vertically along and parallel to the vertical centerline axis Cv of the capsule 4110 from the upper tubesheet 4120 to the lower tubesheet 4122. Each of the tubes 4130 is accessible through the upper tubesheet 4120 (see, e.g. FIG. 40A). In one embodiment, the tubes 4130 each have an associated machined lead-in guide in the upper tubesheet 4120 to support the insertion of the fuel rods. An annular tapered or chamfered entrance 4136 is therefore formed in the upper tubesheet 4120 adjacent and proximate to the top open end 4132 of each tube 4130. The obliquely angled surface (with respect to the vertical centerline axis Cv) of the chamfered entranceways 4136 help center and guide loading of the damaged fuel rods into each of the storage tubes 4130. The top end 4132 of the tubes may therefore be spaced slightly below the top surface 4128 of the upper tubesheet 4120 as shown. The bottom ends 4134 of the fuel rod storage tubes 4130 may rest on the bottom interior surface 4116a of the capsule 4110. Each storage tube 4130 includes one or more flow openings 4133 of any suitable shape located proximate to the bottom ends 4134 of the tubes below the bottom tubesheet 4122. The openings 4133 allow gas to enter the tubes from the bottom plenum 4124 during the forced gas dehydration process and rise upward through the tubes to dry the damaged fuel rods. The fuel rod storage tubes 4130 may be mounted in the upper and lower tubesheets 4120, 4122 by any suitable method. In certain embodiments, the tubes 4130 may be rigidly coupled to upper and/or lower tubesheets 4120, 4122 such as by welding, soldering, explosive tube expansion techniques, etc. In other embodiments, the tubes 4130 may be movably coupled to the upper and/or lower tubesheets to allow for thermal expansion when heated by waste heat generated from the decaying fuel rods and heated forced gas dehydration. Accordingly, a number of possible rigid and non-rigid tube mounting scenarios as possible and the invention is not limited by any particular one. The fuel rod storage tubes 4130 may be arranged in any suitable pattern so long as the fuel rods may be readily inserted into each tube within the fuel pool. In the non-limiting exemplary embodiment shown, the tubes 4130 are circumferentially spaced apart and arranged in a circular array around a central drain tube 4150 further described below. Other arrangements and patterns may be used. Referring now to FIGS. 38, 39, 40A, 40B, and 15, the central drain tube 4150 of the capsule 4110 may be mounted at approximately the geometric center of the upper tubesheet 4120 as shown. The center drain tube 4150 in one arrangement is supported by and extends vertically parallel to and coaxially with centerline axis Cv of the capsule from the upper tubesheet 4120 to the bottom tubesheet 4122. The drain tube 4150 may be rigidly coupled to the tubesheets 4120, 4122 using the same techniques described herein for the fuel rod storage tubes. Drain tube 4150 is a hollow structure forming a pathway for introducing insert drying gas into the tube assembly to dry the interior of capsule 4110 following closure and sealing, as further described herein. The drain tube 4150 includes an open top end 4151 and an open bottom end 4152. The top end functions as a gas inlet and the bottom end functions as a gas outlet, with respect to the dehydration gas flow path further described herein. The bottom end 4152 is open into and may extend slightly below the bottom surface of the lower tubesheet 4122 to place the drain tube in fluid communication with the bottom plenum 4124 of the capsule 4110, as shown for example in FIGS. 40A-B. This forms a fluid pathway for introducing drying gas into the bottom of the capsule 4110. The outlet end 4152 of the drain tube 4150 is spaced vertically apart from the interior bottom surface 4116a of the capsule 4110. Drain tube 4150 may include a sealing feature configured to form a substantially gas-tight seal between the closure lid 4200 and drain tube for forced gas dehydration process. In one embodiment, the sealing feature may be a spring-biased sealing assembly 4140 configured to engage and form a seal with the bottom of the closure lid 4200 for gas drying. The sealing assembly 4140 includes a short inlet tube 4141, an enlarged resilient sealing member 4142 disposed on top of the inlet tube, and spring 4143. Inlet tube 4141 has a length less than the length of the drain tube 4150. Spring 4143 may be a helical compression spring in one embodiment having a top end engaging the underside 4142b of the sealing member 4142 which extends laterally (i.e. transverse to vertical centerline axis Cv) and diametrically beyond the inlet tube 4141, and a bottom end engaging the top surface 4128 of the upper tubesheet 4120. The inlet tube 4141 is rigidly coupled to the sealing member 4142 and has a diameter slightly smaller than the drain tube 4150. This allows the lower portion of the inlet tube 4141 to be inserted into the upper portion of the drain tube 4150 through the top inlet end 4151 for upward/downward movement in relation to the drain tube. Spring 4143 operates to bias the sealing member 4142 and inlet tube 4141 assembly into an upward projected inactive position away from the upper tubesheet 4120 ready to engage the closure lid 4200, as further described herein. Accordingly, the sealing assembly 4140 is axially movable along the vertical centerline axis from the upward projected inactive position to a downward active sealing position. In one embodiment, the sealing member 4142 may have a circular shape in top plan view and a convexly curved or domed sealing surface 4142a in side transverse cross-sectional view (see, e.g. FIGS. 40A and 40B). The curved sealing surface 4142a ensures positive sealing engagement with a gas supply outlet extension tube 4210 in the capsule closure lid 4200 (see FIG. 37) to compensate for irregularities in the extension tube end surface edges and less than exact centering of the extension tube with respect to the sealing member 4142, thereby preventing substantial leakage of drying gas when coupled together. The sealing member 4142 includes a vertically oriented through-hole 4144 to form a fluid pathway through the sealing member to the drain tube 4150. In one embodiment, the sealing member 4142 may be made of a resiliently deformable elastomeric material suitable for the environment of a radioactive damaged fuel rod storage capsule. The elastomeric seal provides sufficient sealing and a leak-resistant interface between the central drain tube 4150 and closure lid 4200 to allow the inert drying gas (e.g. helium, nitrogen, etc.) to be pumped down the central drain tube to the bottom of the capsule 4110 during the forced gas dehydration process. It will be appreciated that other types of seals and arrangements may be used. Accordingly, in some embodiments metal or composite metal-elastomeric sealing members may be used. The sealing member may also have other configurations or shapes instead of convexly domed, such as a disk shaped with a flat top surface or other shape. In other embodiments, a non-spring activated sealing assembly may be used. Accordingly, the invention is not limited by the material of construction or design of the seal and sealing assembly so long as a relatively gas-tight seal may be formed between the closure lid gas outlet extension tube 4210 and the drain tube 4150 for forced gas dehydration of the capsule 4110. The fuel rod basket assembly, including the foregoing tubesheets, rod storage tubes, central drain tube, and sealing assembly may be made of any suitable preferably corrosion resistant material such as stainless steel. Other appropriate materials may be used. The closure lid 4200 will now be further described. Referring to FIGS. 32-37 and 46, lid 4200 in one embodiment may have a generally rectilinear cube-shaped body to complement the shape of cavity 4112 in capsule 4110 in which at least a portion of the lid is received. Accordingly, in one embodiment the lid 4200 and capsule 4110 may have a square shape in top plan view. Lid 4200 further has a substantially solid internal structure except for the gas flow conduits formed therein, as further described below. The lid 4200 is formed of a preferably corrosion resistant metal, such as stainless steel. Other materials may be used. Lid 4200 includes a top surface 4202, bottom surface 4204, and lateral sides 4206 extending between the top and bottom surfaces. The lateral sides 4206 of the lid have a width sized to permit insertion of a majority of the height of the lid into the cavity 4112 of the capsule. The bottom of the lid 4200 includes a peripheral skirt 4212 extending around the perimeter of the bottom surface 4204 that engages and rests on the top surface 4128 of the upper tubesheet 4120 of the capsule 4110 when the lid is mounted in the capsule. In one embodiment, the skirt 4212 is continuous in structure and extends around the entire perimeter without interruption. The skirt 4212 projects downward for a distance from the bottom surface 4204 of the lid which is recessed above the bottom edge 4212a of the skirt. The forms a downwardly open space 4211 having a depth commensurate with the height of the skirt 4212. When the bottom edge 4212a of skirt 4212 rests on top surface 4128 of the upper tubesheet 4120, the top plenum 4126 is formed between the bottom surface 4204 of lid 4200 and the upper tubesheet inside and within the skirt 4212. The bottom edge 4212a of the skirt 4212 thereby forms a seal between the upper tubesheet 4120 and lid 4200 for forced gas dehydration of the capsule 4110. An enlarged seating flange 4208 extends around the entire perimeter of the lid 4200 adjacent to top surface 4202 and projects laterally beyond the sides 4206. The top surface 4202 may be recessed below the top edge 4208a of the seating flange 4208 as shown. A stepped shoulder 4213 is formed between seating flange 4208 and sides 4206 which engages and seats on a mating shoulder 4113 formed inside the mounting flange 4111 of capsule 4110 in cavity 4112 (see particularly FIG. 46A). Both mating shoulders 4213 and 4113 extend around the entire perimeter regions of the lid 4200 and capsule 4110 respectively and limit the insertion depth of the lid into the capsule. In one embodiment, the top edges 4111a and 4208a of the mounting flange 4111 and seating flange 4208 respectively are flush with each other and lie in approximately the same horizontal plane when the closure lid 4200 is fully mounted in the capsule 4110 (see, e.g. FIGS. 41A, 41B, and 46A). This facilitates formation of an open V-groove weld 4205 to hermetically seal the lid to the capsule. The mounting and seating flanges 4111, 4208 each include opposing beveled faces 4115, 4208 respectively to form the V-groove. Because of the recessed top surface 4202 of the lid 4200 and mounting flange 4111, access is available to both sides of finished weld which advantageously permits full volumetric inspection of the weld such as by ultrasonic non-destructive testing or other methods. The source and detector of the ultrasonic test (UT) equipment may therefore be placed on opposite sides of the weld for full examination. A multi-pass welding process may be used which prevents any potential through-cracking of a single weld line in the case of an undetected defect. This parallels welding processes used in the United States for Multi-Purpose Canisters (MPCs), but is modified to allow volumetric weld examination (a key consideration for acceptance of weld integrity by some international regulators). Each pass is followed by a Liquid Penetrant Test (LPT) to identify defects in the weld layer as the weld is formed. The finished weld is then volumetrically tested using UT. Unlike a bolted joint sealed with gaskets, a welded joint with volumetric inspection typically does not require leak-monitoring or checks prior to future transport. FIGS. 41A and 41B show the lid 4200 and capsule 4110 before and after welding, respectively. This does not limit the capsule to having a bolted lid, similar to dual-purpose metal casks used for storage and transport of spent nuclear fuel. In such embodiment, the capsule would have one more seals, for example elastomeric or metallic, that would be compressed during tightening of the lid bolts on the capsule, forming a hermetic seal. According to another aspect of the invention, the closure lid 4200 is configured to permit forced gas dehydration of the capsule 4110 and plurality of damaged fuel rods contained therein after the lid is seal welded to the capsule. Accordingly, the lid 4200 includes a combination of gas ports and internal fluid conduits to form a closed flow loop through capsule 4110. Referring now to FIGS. 32-37 and 46, lid 4200 includes a gas supply port 4220 and gas return port 4222 formed in the top surface 4202 of the lid, and a gas supply outlet 4224 and gas return inlet 4226 formed in the bottom surface 4204 of the lid. In one configuration, the gas supply outlet 4224 and return inlet 4226 may be located at diagonally opposite corner regions of the top surface 4202 of the lid 4200 proximate to the lateral sides 4206. The gas supply port 4220 is fluidly coupled to the gas supply outlet 4224 via an internal flow conduit 4228. The gas return port 4222 is fluidly coupled to the gas return inlet 4226 via another separate internal flow conduit 4230 which is fluidly isolated from flow conduit 4228. In one embodiment, the flow conduits 4228, 4230 each follow a torturous multi-directional path through the lid to prevent neutron streaming. In one configuration, flow conduit 4228 includes a vertical section 4222a connected to gas supply outlet 4224, first horizontal section 4228b connected thereto, second horizontal section 4228c connected thereto, and second vertical section 4228d connected thereto and gas supply port 4220. The flow conduit sections 4228a-d may be arranged in a rectilinear pattern. Flow conduit 4228 includes a vertical section 4230a connected to gas return port 4222, horizontal section 4230b connected thereto, and second vertical section 4230c connected thereto and gas return inlet 4226. The flow conduit sections 4230a-c may also be arranged in a rectilinear pattern. Because the lid 4200 has a solid internal structure, the flow conduits may be formed by drilling or boring holes through the lateral sides 4206 and top and bottom surfaces 4202, 4204 of the lid to points of intersection between the conduits as best shown in FIGS. 36 and 46. After formation of the flow conduits, the penetrations 4232 in the lateral sides 4206 of the lid may be closed using threaded and/or seal welded metal caps applied before mounting and welding the lid 4200 to the capsule 4110. The penetrations 4232 in the bottom surface 4204 of the lid may remain open. The gas supply and return port penetrations 4232 in the top surface 4202 of the lid may be threaded and closed using threaded caps 4234 to permit removal and installation of remote valve operating assemblies 4240 (RVOAs) for forced gas dehydration of the capsule, as shown in FIGS. 45 and 46. It should be noted that the gas supply outlet 4224 in lid 4200 is fluidly coupled to the gas supply outlet extension tube 4210. The extension tube 4210 compensates for the height of the lid bottom skirt 4212 to allow physical coupling of the tube to the sealing assembly 4140 when the skirt rests on the top surface 4128 of the upper tubesheet 4120. In one embodiment, the extension tube 4210 and gas supply outlet 4224 are centered on the bottom surface 4204 of the lid 4200. In certain other embodiments, the extension tube may be omitted and the gas supply outlet 4224 penetration may be directly coupled to the sealing assembly 4140. A method for storing and drying fuel rods using capsule 4110 will now be briefly described. The method may be used for storing intact or damaged fuel rods, either of which may be stored in capsule 4110. The process begins with the top of the capsule 4110 being open so that the storage tubes 4130 are accessible for loading. The loading operation involves inserting the fuel rods into the storage tubes 4130. After the capsule is fully loaded, the lid 4200 is attached to the top end 4114 and sealed to the capsule. In one preferred embodiment, the lid is sealed welded to the capsule as described elsewhere herein to form a gas tight seal After lid 4200 is seal welded to the capsule 4110, the interior of the capsule and fuel rods therein may be dried using heated forced gas dehydration (FGD) system such as those available from Holtec International of Marlton, N.J. Commonly owned U.S. Pat. Nos. 7,096,600, 7,210,247, 8,067,659, 8,266,823, and 7,707,741, which are all incorporated herein by reference in their entireties, describe such systems and processes as noted above. The remote operated valve assemblies 4240 are first installed in the gas supply and gas return ports 4220, 4222. The valves are then connected to the gas supply and return lines from the FGD system. The next steps, described in further detail herein, include pumping the inert drying gas from the FGD system or source through the gas supply conduit into the cavity 4112 of the capsule 4110 and into the bottom plenum 4124, flowing the gas through each of the storage tubes 4130 to dry the fuel rods, collecting the gas leaving the storage tubes in the top plenum 4126, and flowing the gas through the gas return conduit back to the FGD source. The process continues for a period of time until analysis of the drying gas shows an acceptable level of moisture removal from the capsule 4110. Referring now to FIGS. 36, 40A, 45, and 46, threaded caps 4234 may first be removed from the gas supply and return ports 4220 and 4222 in the lid 4200 which is welded to the capsule 4110. A remote valve operating assembly 4240 is then threadably coupled to each port 4220, 4222. The gas supply and return lines from the FGD skid which holds the dehydration system equipment are then fluidly coupled to the valve assemblies. The dehydration and drying process is now ready to commence by pumping the inert and heat drying gas from the FGD system through the capsule 4110 to dry the fuel rods in the storage tubes 4130, as further described herein. Gas supplied from the FGD system first flows through the first valve assembly 4240 into the lid 4200 through the gas supply port 4220. The supply gas then flows through flow conduit 4228 to the gas supply outlet 4224 and then into gas supply outlet extension tube 4210. The supply gas enters the sealing assembly 4140 and flows downwards through the central drain tube 4150 into the bottom plenum 4124 of the capsule 4110. The gas in the bottom plenum enters the bottom of the fuel rod storage tubes 4120 through openings 4133 formed in and proximate to the bottom ends 4134 of the tubes. The gas flows and rises upwards through each of the storage tubes 4120 to dry the damaged fuel rods stored therein. The gas then enters the top plenum 4126 above the upper tubesheet 4120 beneath the lid 4200. From here, the gas leaves the top plenum and enters the gas return inlet 4226 in the lid. The gas flows through flow conduit 4230 to the gas return port 4222 and into the remote valve operating assembly 4240 connected thereto. The return gas then flows through the return line back to the FGD system skid to complete the closed flow loop. Advantageously, the present invention allows drying of multiple damaged fuel rods in the capsule 4110 simultaneously instead of on an individual, piece-meal basis. This saves time, money, and operator dosage of radiation. According to another aspect of the invention, the lid 4200 includes a threaded lifting port 4340 configured for temporary coupling to a lifting assembly 4342 that may be used for moving and transporting the capsule 4110 around the fuel pool and loading into transport casks or multi-purpose canisters. The lifting assembly 4342 in one embodiment may include a lifting rod 4344 including a bottom threaded end 4346 for rotatable coupling to the threaded lifting port 4340 and an opposite top operating end 4348 configured for rigging to equipment such as a crane that may be used to lift and maneuver the capsule 4110. According to yet another aspect of the invention, a lid-based capsule storage system is provided which is configured for holding and supporting a plurality of capsules 4110. The capsule storage system includes a cask loading lid 4400 which may be configured to retrofit and replace lids used in existing transport or transfer casks used for loading, storing, and transporting undamaged fuel bundles. Using the temporary lid, the existing casks may used to provide radiation shielding during the capsule 4110 drying and closure operations described herein. Referring to FIGS. 42-46, the loading lid 4400 can be designed for any dual-purpose metal casks, such as those supplied by Holtec, TNI, or GNS or transfer casks, such as the HI-STRAC used by Holtec International in Marlton, N.J. Loading lid 4400 may have multiple mounting cutouts or openings 4302 extending completely through the lid each of which are designed to allow insertion of a single capsule 4110. The mounting openings 4302 are sized smaller than the mounting flange 4111 of the capsule 4110 so that the flange remains above the top surface 4402 of the lid 4400. A shoulder 4404 is formed beneath each mounting flange 4111 between the flange and sidewalls 4118 of the capsule which engages the top surface 4402 of the lid 4400. This allows the capsules to hang from the lid 4400 in a vertically cantilevered manner. The top of the capsule 4110 therefore sites about 10-15 mm above the lid surface 4402 in one representative non-limiting embodiment to enable workers to easily access the top of the capsules to perform the closure operations. The location of the mounting openings 4302 can be optimized to allow easy worker access to the capsules during the drying and closure operations. According to another aspect of the invention shown in FIG. 47, a leak testing lid 4500 is provided which can be coupled and sealed to the mounting flange 4111 of the capsule 4110. The lid 4500 attached to the mounting flange 4111 of capsule 4110 and includes a piping connection assembly 4502 which allows hook-up to leak testing equipment for performance of an integrated leak test of the entire sealed capsule 4110. Although the fuel rod encapsulation capsule is described herein for use with damaged fuel rods, it will be appreciated that the capsule has further applicability for use with intact fuel rods or debris storage as well. Accordingly, the invention is expressly not limited for use with damaged fuel rods alone. V. Inventive Concept 5 With reference to FIGS. 50-56, a fifth inventive concept will be described. Turning in detail to the drawings, FIG. 50 schematically shows a fuel rack 5101, according to one embodiment of the invention, placed in a cooling pool 5103 for the storage of spent nuclear fuel. As is known in the art, the cooling pool 5103 may include treated water to aid in neutron absorption and heat dispersion, with examples including demineralized water and borated water. The fuel rack 5101, as shown in FIG. 51, includes a rectilinear array of hexagonal fuel storage cells 5105. The fuel rack 5101 is a cellular, upright, prismatic module. The illustrated embodiment of the fuel rack 5101 is specifically designed to accommodate hexagonal fuel assemblies, such as VVER 1000 fuel assemblies. To this extent, each fuel storage cell 5105 of the fuel rack 5101 also has a hexagonal cross-sectional profile so as to geometrically accommodate no more than a single hexagonal fuel assembly. In certain embodiments, the hexagonal cross-sectional profile of the storage cell 5101 may have a shape that is other than a regular hexagon. It is to be understood that the concepts of the present invention can be modified to accommodate any shaped fuel assembly, including rectangular, octagonal, round, among others. The fuel rack 5101 includes a base plate 5111, support pedestals 5131, and a plurality of storage tubes 5151 placed together in a side-by-side arrangement to form a rectilinear array as shown in FIG. 52A. The support pedestals 5131 are affixed to a bottom surface 5113 of the base plate 5111, and the array of storage tubes 5151 are affixed to the top surface 5115 of the base plate 5111 in a substantially vertical orientation. Each storage tube 5151 extends along its own longitudinal axis LA, and in addition to being substantially vertical, each longitudinal axis LA is also substantially perpendicular to the top surface 5115 of the base plate 5111. The connection between each of the storage tubes 5151 and the base plate 5111 is achieved by welding the bottom edge of each of the storage tubes 5151 to the top surface 5115 of the base plate 5111. Similarly, the connection between each of the support pedestals 5131 and the base plate 5111 is achieved by welding each of the support pedestals 5131 to the bottom surface 5113 of the base plate 5111. By welding the storage tubes 5151 to the base plate 5111, the flexural strength of the base plate 11 may be increased, thereby making it possible to support the combined weight of the fuel rack and fuel assemblies with the support pedestals 5131 located only near the edges of the base plate 5111. Of course, other connection techniques can be utilized for either or both of the storage tubes 5151 and the support pedestals 5131 with minor modification, including mechanical connections such as bolting, clamping, threading, and the like. As shown in FIGS. 52A-D, the storage tubes 5151 are connected to the base plate 5111 to form a plurality of rows 5153 and a plurality of columns 5155. The storage tubes 5151 within each row 5153 are placed in a spaced apart manner, with the spacing between adjacent storage tubes 5151 in a row 5153 being maintained by spacers 5157. Spacers 5157 are placed between all adjacent storage tubes 5151 within a row 5153, with several spacers 5157 being used to separate two adjacent storage tubes 5151. The spacers 5157 are welded in place to each of the adjacent storage tubes 5151. Several spacers 5157 are placed between each of the aligned longitudinal edges of adjacent storage tubes 5151, with spacers 5157 being placed at the top and bottom of aligned longitudinal edges, and the other spacers being spaced along the aligned longitudinal edges. The number of spacers 5157 included between adjacent storage tubes 5151 may vary depending on factors such as the desired fluid flow between adjacent storage tubes 5151 and/or between adjacent columns 5155, space considerations, and weight of the entire fuel rack, among other considerations. By having the spacers 5157 distributed in this manner, the space between adjacent columns 5155 forms flux traps 5159, not only between adjacent ones of the storage tubes 5151 within each row 5153, but also between entire columns 5155. These flux traps 5159 are exterior to each of the storage tubes 5151, and because the flux trap 5159 of one row 5153 is not partitioned from the flux trap 5159 of an adjacent row 5153, adjacent ones of the flux traps 5159 effectively separate one column 5155 from another. The width of the spacers 5157, and thus the width of the flux traps 5159, may be selected to tailor the ability to control criticality of the nuclear fuel stored within the fuel rack 5101. The storage tubes 5151 within each column 5155 are placed adjacent each other so that the outer walls of adjacent storage tubes 5151 within the respective column 5155 are in surface contact with one another. Each of the aligned longitudinal edges of adjacent storage tubes 5151 within a column 5155 may be contiguously welded together to provide additional stability to the overall structure of the fuel rack 5101. With the rectilinear array of the fuel rack 5101 formed with the plurality of rows 5153 and columns 5155 as described above, the longitudinal axes LA of each of the storage tubes 5151 in each of the rows 5153 and in each of the columns 5155 align to form reference planes RP. Also, the longitudinal axes LA of adjacent storage tubes 5151 in one of the rows 5153 may be separated from one another by a distance D1, and the longitudinal axes LA of adjacent storage tubes 5151 in one of the columns 5155 may be separated from one another by a distance D2, which may different, and even greater, than the distance D1. The distance D1 separating adjacent storage tubes 5151 within a row 5153 may be controlled within a design by appropriate selection of either the width of the storage tubes 5151 or the width of the spacers 5157. The distance D2 separating adjacent storage tubes 5151 within a column 5155 may be controlled within a design by appropriate selection of the length of the storage tubes 5151. An exemplary storage tube 5151 is shown in FIG. 53A. The storage tube 5151 includes an outer tube 5161 having a rectangular cross-section, as can be seen in FIG. 53B. The top end of the storage tube 5151 remains open so that a fuel assembly can be inserted into the hexagonal fuel storage cell 5105 formed therein. The storage tube 5151 includes a first pair of opposing wall plates 5163, 5165 and a second pair of opposing wall plates 5167, 5169. The outer walls of the first pair of wall plates 5163, 5165 are placed into surface contact with respective outer walls of wall plates 5163, 5165 of adjacent storage tubes 5151 to form the columns 5155 of the rectilinear array, as discussed above. The storage tube 5151 defines a longitudinal axis LA, which is the center point of the rectangular cross-section, and the wall plates 5163, 5165, 5167, 5169 each have an overall height H1. The top of each of the second pair of opposing wall plates 5167, 5169 includes a guide plate 5171. The guide plate 5171 for each wall plate 5167, 5169 extends at an angle up from the respective wall plate 5167, 5169 and away from the longitudinal axis LA of the storage tube 5151. The guide plates 5171 provide a surface to aid in guiding a fuel assembly into the fuel storage cell 5105 formed within the storage tube 5151. The guide plates 5171 also help reduce the amount of wear and/or damage caused to the top edge of the wall plates 5167, 5169 during the process of loading a fuel assembly into the fuel storage cell 5105. The guide plates 5171 may be integrally formed with the wall plates 5167, 5169, or they be mounted as part of a separate structure to the external walls of the wall plates 5167, 5169. The outer walls of the second pair of opposing wall plates 5167, 5169 each have a neutron-absorbing plate 5173 coupled thereto, and the neutron-absorbing plate 5173 is secured in place against the outer walls of the second pair of opposing wall plates 5167, 5169 by an outer sheath 5175. The outer sheath 5175 encloses the neutron-absorbing plate 5173 in a pocket 5177, which is also shown in FIG. 53C, to protect the pool water from possible deterioration of the neutron-absorbing plate 5173. The neutron-absorbing plate 5173 and the outer sheath 5175 extend a height H2, which is less than the height H1. The height H2 may be the equivalent of the height of a fuel assembly positioned for storage within the fuel storage cell 5105. Of course, the height H2 of the neutron-absorbing plate 5173 and the outer sheath 5175 may, in certain embodiments, be as great as the height H1 of the outer tube 5161. An inner plate-assemblage 5191 is positioned within the outer tube 5161 to help form the fuel storage cell 5105. The inner plate-assemblage 5191 includes two chevron plates 5193a, 5193b, which may be of identical design. An exemplary chevron plate 5193, representative of both chevron plates 5193a, 5193b, is shown in FIG. 52B. The chevron plate 5193 includes two wall plates 5195 adjoined at an apex edge 5197, and each wall plate 5195 may have a height H3, which is slightly less than the height H1 of the wall plates 5163, 5165, 5167, 5169 of the storage tube 5151. The top of each wall plate 5195 includes a guide plate 5199. The guide plate 5199 for each wall plate 5195 extends at an angle up from the respective wall plate 5195, such that when the chevron plate 5193 is in place within the outer tube 5161 of the storage tube 5151, the guide plates 5199 also extend away from the longitudinal axis LA of the storage tube 5151. The guide plates 5199 provide a surface to aid in guiding a fuel assembly into the fuel storage cell 5105 formed within the storage tube 5151. The guide plates 5199 also help reduce the amount of wear and/or damage caused to the top edge of the wall plates 5195 during the process of loading a fuel assembly into the fuel storage cell 5105. The guide plates 5199 may be integrally formed with the wall plates 5195, or they be mounted as part of a separate structure to the external walls of the wall plates 5195. The outer walls of the wall plates 5195 each have a neutron-absorbing plate 5201 coupled thereto, and the neutron-absorbing plate 5201 is secured in place against the outer walls of the wall plates 5195 by an outer sheath 5203. Each outer sheath 5203 encloses the respective neutron-absorbing plate 5201 in a pocket 5205, which is also shown in FIG. 53C, to protect the pool water from possible deterioration of the neutron-absorbing plate 5201. The neutron-absorbing plate 5201 and the outer sheaths 5203 extend a height H2, which is less than the height H3 of the wall plates 5195. The height H2 may be the equivalent of the height of a fuel assembly positioned for storage within the fuel storage cell 5105. Of course, the height H2 of the neutron-absorbing plate 5201 and the outer sheaths 5203 may, in certain embodiments, be as great as the height H3 of the wall plates 5195. The dimension and position of the neutron-absorbing plate 5173 on the wall plates 5167, 5169 of the outer tube 5161, and the neutron-absorbing plate 5201 on the wall plates 5195 of the chevron plates 5193, may be determined by the position and dimension of a fuel assembly positioned for storage within the fuel storage cell 5105, and more particularly by the position and dimension of fuel rods contained within any such fuel storage assembly. The neutron-absorbing plates 5173, 5201 are generally placed on the respective wall plates 5167, 5169, 5195 and dimensioned so that the height H2 is at least as great as the height of stored fuel rods within the fuel storage cell 5105. Such dimensioning of the neutron-absorbing plates 5173, 5201 helps ensure that neutron emissions, directed toward any of the wall plates 5167, 5169, 5195 from the fuel assembly within the fuel storage cell 5105, are incident on the neutron-absorbing plates 5173, 5201. The outer sheaths 5175, 5203 on the wall plates 5167, 5169, 5195 are dimensioned to provide a sufficiently large enclosure to secure the neutron-absorbing plates 5173, 5201 to the respective wall plates 5167, 5169, 5195. The neutron-absorbing plate 5173, 5201 may be formed of a material containing a neutron absorber isotope embedded in the microstructure, such as elemental boron or boron carbide. Metamic, produced by Metamic, LLC, which is made of an aluminum alloy matrix with embedded boron carbide, is an example of an acceptable material. In certain embodiments, the outer sheaths 5175, 5203 may be formed of materials such as stainless steel, borated stainless steel, or any other type of steel appropriate for use in the long term storage environment for spent nuclear fuel. In certain embodiments, particularly those in which the neutron-absorbing plates 5173, 5201 are not formed of a material which is brittle or becomes brittle over time, thereby presenting a risk of deterioration and contamination of the pool water, the neutron-absorbing plates 5173, 5201 may be secured directly to the respective wall plates 5167, 5169, 5195. In such embodiments, the outer sheaths 5175, 5203 may be omitted, or alternatively, the outer sheaths 5175, 5203 may be configured to couple the neutron-absorbing plates 5173, 5201 to the respective wall plates 5167, 5169, 5195 without enclosing the neutron-absorbing plates 5173, 5201 in an envelope. FIG. 53C shows a cross-section of an exemplary storage tube 5151. The outer tube 5161 has a width W in the row direction and a length L in the column direction, and the length L in the column direction is greater than the width w in the row direction. The inner surface 5211 of the outer tube 5161 of the storage tube 5151 defines an inner cavity 5213, and a hexagonal fuel storage cell 5105 is formed within the inner cavity 5213 of the storage tube 5151. The profile of a hexagonal fuel assembly 5109 is shown for reference within the fuel storage cell 5105. In certain embodiments, the gap between the fuel assembly 5109 and the walls forming the fuel storage cell 5105 is less than about 4 mm around all sides of the fuel assembly 5109. The inner plate-assemblage 5191 is positioned within the outer tube 5161 to divide the inner cavity 5213 into a plurality of interior flux trap chambers 5215a-d and the fuel storage cell 5105. In the rectilinear array of the storage tubes 5151, these flux trap chambers 5215a-d serve as interior flux trap chambers between the fuel storage cells 5105 of adjacent storage tubes 5151 in the fuel rack 5101. Thus, storage tubes 5151 that are adjacent within a row have their respective fuel storage cells 5105 separated by four flux trap chambers, two from each of the adjacent storage tubes 5151. The inner plate-assemblage 5191 includes two chevron plates 5193a, 5193b. Each chevron plate 5193a, 5193b includes two wall plates 5195a-d, and each wall plate 5195a-d is oblique to and extends between adjacent sides of the outer tube 5161 to form the plurality of interior flux trap chambers 5215a-d within the inner cavity 5213. Specifically, the wall plate 5195a of the chevron plate 5193a extends between the wall plate 5167 of the outer tube 5161 and the wall plate 5163 of the outer tube 5161 to form the interior flux trap chamber 5215a. With the wall plate 5195a positioned in this manner, the interior flux trap chamber 5215a is formed between the wall plate 5195a of the chevron plate 5193a and a corner section formed at the intersection of wall plates 5163, 5167 of the outer tube 5161. The wall plate 5195b of the chevron plate 5193a extends between the wall plate 5169 of the outer tube 5161 and the wall plate 5163 of the outer tube 5161 to form the interior flux trap chamber 5215b. With the wall plate 5195b positioned in this manner, the interior flux trap chamber 5215b is formed between the wall plate 5195b of the chevron plate 5193a and a corner section formed at the intersection of wall plates 5163, 5169 of the outer tube 5161. The wall plate 5195a and the wall plate 5195b are joined at an apex edge 5197a of the chevron plate 5193a. The edges of the wall plates 5195a, 5195b that are positioned against the wall plates 5167, 5169, respectively, are contiguously welded to the inner surface 5211 of the rectangular outer tube 5161. Similarly, the wall plate 5195c of the chevron plate 5193b extends between the wall plate 5169 of the outer tube 5161 and the wall plate 5165 of the outer tube 5161 to form the interior flux trap chamber 5215c. With the wall plate 5195c positioned in this manner, the interior flux trap chamber 5215c is formed between the wall plate 5195c of the chevron plate 5193b and a corner section formed at the intersection of wall plates 5165, 5169 of the outer tube 5161. The wall plate 5195d of the chevron plate 5193b extends between the wall plate 5167 of the outer tube 5161 and the wall plate 5165 of the outer tube 5161 to form the interior flux trap chamber 5215d. With the wall plate 5195d positioned in this manner, the interior flux trap chamber 5215d is formed between the wall plate 5195d of the chevron plate 5193b and a corner section formed at the intersection of wall plates 5165, 5167 of the outer tube 5161. The wall plate 5195c and the wall plate 5195d are joined at an apex edge 5197b of the chevron plate 5193a. The edges of the wall plates 5195c, 5195d that are positioned against the wall plates 5167, 5169, respectively, are contiguously welded to the inner surface 5211 of the rectangular outer tube 5161. With this configuration of the chevron plates 5193a, 5193b within the outer tube 5161, the hexagonal fuel storage cell 5105 is defined by: the inner surface 5217a of the first wall plate 5195a of the first chevron plate 5193a; the inner surface 5217b of the second wall plate 5195b of the first chevron plate 5193a; the inner surface 5217c of the first wall plate 5195c of the second chevron plate 5193b; the inner surface 5217d of the second wall plate 5195d of the second chevron plate 5193b; a portion of the inner surface 5211 of the wall plate 5167 of the outer tube 5161; and a portion of the inner surface 5211 of the wall plate 5169 of the outer tube 5161. Each of the flux trap chambers 5215a-d formed by this configuration of the chevron plates 5193a, 5193b have triangular transverse cross-sections. The size and hexagonal cross-sectional shape of the fuel storage cell 5105 is designed and constructed so that the fuel storage cell 5105 can accommodate no more than one fuel assembly 5109. Due to the different cross-sectional shape of the flux trap chambers 5215a-d, as compared to the cross-sectional shape of the typical fuel storage assembly, the flux trap chambers 5215a-d are not able to accommodate a fuel assembly that has a square or hexagonal transverse cross-section. The apex edges 5197a, 5197b of each of the chevron plates 5193a, 5193b are located in a reference plane RP that is defined by including the longitudinal axis LA of the storage tube 5151 and being perpendicular to the wall plates 5163, 5165 of the outer tube 5161. The apex edges 5197a, 5197b may form an angle of 5120°, so that the resulting hexagonal cross-sectional shape of the fuel storage cell 5105 forms a regular hexagon. In alternative embodiments, the apex edges 5197a, 5197b may form an angle α of slightly less than 120°, within the range of about 120°-115°, so that the resulting hexagonal cross-sectional shape of the fuel storage cell 5105 varies slightly away from the form of a regular hexagon. When the hexagonal fuel assembly is placed within the fuel storage cell 5105, the fuel assembly may rattle undesirably during a seismic or other rattling event. By having the apex edges 5197a, 5197b forming an angle of slightly less than 120°, the acute edges of the fuel assembly that face the apex edges 5197a, 5197b are prevented from impacting the apex edges 5197a, 5197b during a seismic or other rattling event. A cross-section of the storage tube 5151 is shown in FIG. 53D with a schematic representation of a fuel assembly 5109 disposed within the fuel storage cell 5105. Similar to hexagonal fuel assemblies commonly in use, the fuel assembly 5109 includes a top handle 5233, a body portion 5235, in which a plurality of nuclear fuel rods (not shown) are housed, and a tapered bottom portion 5237. The handle 5233 and the tapered bottom portion 5237 facilitate inserting the fuel assembly 5109 into the fuel storage cell 5105 of the storage tube 5151. When the fuel assembly 5109 is being inserted into the storage tube 5151, the tapered bottom portion 5237 may engage the guide plates 5171, 5199 to aid in centering the fuel assembly 5109 within the fuel storage cell 5105. As shown, with the fuel assembly 5109 fully inserted into the fuel storage cell 5105, the height H1 of the outer tube 5161 is greater than the overall height H4 of the fuel assembly 5109. The height H3 of the chevron plates 5193a, 5193b is also less than the height H1 of the outer tube 5161. The lower edges of the chevron plates 5193a, 5193b do not extend to the lower edge of the outer tube 5161, so that a gap is formed at the lower end of the storage tube 5151 for cooling fluid to flow into the flux trap chambers 5215a-d. In certain embodiments, the chevron plates 5193a, 5193b may include apertures at their bottom edges for cooling fluid to flow into the flux trap chambers 5215a-d, and in such embodiments, the height H3 of the chevron plates 5193a, 5193b may be the same as the height H1 of the outer tube 5161. The height H2 of the neutron-absorbing plates 5201 coupled to the chevron plates 5193a, 5193b (and the neutron-absorbing plates 5173 coupled to the outer tube 5161 as shown in FIG. 53C) is substantially the same as the height of the body portion 5235 of the fuel assembly 5109. In certain embodiments, the height H2 of the of the neutron-absorbing plates 5201 (and 5173) may be less than the height of the body portion 5235 of the fuel assembly 5109. The height H2 of the neutron-absorbing plates 5201 (and 5173) may be designed to provide appropriate shielding of adjacent fuel assemblies from one another. This is because adjacent spent nuclear fuel rods may not extend the entire length of the body portion 5235 of the fuel assembly 5109, and the height of the neutron-absorbing plates 5201 (and 5173) need only be high as the nuclear fuel rods when the fuel assembly 5109 is positioned within the storage tube 5151. The base plate 5111, which is shown in FIG. 54, includes a plurality of flow holes 5117 extending through the base plate 5111 from the bottom surface 5113 to the top surface 5115. The base plate 5111 also includes four oblong holes 5119 (second row in from the corners) for lifting and installing the fuel rack 5101 within the fuel pool 5103. Typically, a special lifting beam with four long reach rods is used to interact with the oblong holes 5119 to grapple the fuel rack 5101 for transfer into or out of, or movement within, the pool 5103. The flow holes 5117 (and oblong holes 5119) create passageways from below the base plate 5111 into the bottom ends of the fuel storage cells 5105 formed by the storage assemblies 5151. As shown, a single flow hole 5117 is provided for each storage assembly 5151. In certain embodiments, multiple flow holes 5117 may be provided for each storage assembly 5151 to provide cooling fluid to the fuel storage cell 5105 and each of the flux trap chambers 5215a-d. The flow holes 5117 serve as fluid inlets to facilitate natural thermosiphon flow of pool water through the fuel storage cells 5105 when fuel assemblies having a heat load are positioned therein. More specifically, when heated fuel assemblies are positioned in the fuel storage cells 5105 in a submerged environment, the water within the fuel storage cells 5105, and within the flux trap chambers 5215a-d, surrounding the fuel assemblies becomes heated, thereby rising due to increased buoyancy. As this heated water rises and exits the storage assemblies 5151 via their open top ends, cool water is drawn into the bottom of the fuel storage cells 5105 and the flux trap chambers 5215a-d via the flow holes 5117. This heat induced water flow along the fuel assemblies then continues naturally. A support pedestal 5131 for the fuel rack 5101 is shown in FIG. 55. The support pedestals 5131 affixed to the bottom surface 5113 of the base plate 5111 ensure that a space exists between the floor of the pool 5103 and the bottom surface 5113 of the base plate 5111, thereby creating an inlet plenum for water to flow through the flow holes 5117. The support pedestal 5131 includes a base portion 5133 and a riser portion 5135 formed about an interior flow space 5139. The riser portion 5135 includes flow apertures 5141 through which water from the pool 5103 may pass from a space external to the support pedestal 5131 into the interior flow space 5139. Water passing into the interior flow space 5139 may then pass up through a flow hole 5117 in the base plate 5111 to enable the cooling process described above. Although the riser portion 5135 is depicted as being annular, in certain embodiments the riser portion 5135 may have any geometrical configuration which supports the base plate 5111 above the floor of the pool 5103 and permits water from the pool 5103 to flow into any flow holes 5117 in the base plate 5111 near which the support pedestal 5131 may be affixed. The fuel rack 5101 described above with reference to FIGS. 50-55 is intended to be placed free standing in a pool 5103, without being coupled to sides or the bottom of the pool. However, in certain embodiments, a coupler may be used to aid in securing the position of the fuel rack 5101 within the pool 5103 during a seismic or other rattling event. Other than the neutron absorbing material described above, the fuel rack may be formed entirely from austenitic stainless steel. Although other materials may be used, some materials, such as borated stainless steel, are not preferred for a free standing fuel rack 5101 within a pool 5103, as the greater weight of materials such as borated steel aggravate the seismic response of the fuel rack 5101, thus forcing the fuel rack 5101 to be anchored. An alternative embodiment of a fuel rack 5301 is shown in FIG. 56. This fuel rack 5301 includes a plurality of storage tubes 5303 affixed to the top surface of a base plate 5309, and support pedestals 5311 affixed to the bottom surface of the base plate 5309. The storage tubes 5303 each include a fuel storage cell 5305, and they are placed together in a side-by-side arrangement to form a plurality of rows 5305 and a plurality of columns 5307 as part of a rectilinear array, in the manner described above. A plurality of auxiliary flow apertures 5313 are included in the storage tubes 5303 at or near their bottom edges. In certain embodiments, at least one auxiliary flow aperture 5313 is included in each face of the storage tubes 5303, even those faces of storage tubes 5303 that are placed in surface contact with the face of an adjacent storage tube 5303. The auxiliary flow apertures 5313 act as additional inlet openings (when combined with flow holes in the base plate 5309) for incoming pool water to facilitate the thermosiphon flow during the cooling process. While an auxiliary flow aperture 5313 is shown in each face of each and every storage tube 5303 in the fuel rack 5301, in certain embodiments the auxiliary flow aperture 5313 may be omitted from a select subset of faces for select storage tubes 5303. VI. Inventive Concept 6 With reference to FIGS. 57-63, a sixth inventive concept will be described. Referring to FIGS. 58-62, an environmentally sequestered spent fuel pool system includes a spent fuel pool 6040 comprising a plurality of vertical sidewalls 6041 rising upwards from an adjoining substantially horizontal base wall or slab 6042 (recognizing that some slope may intentionally be provided in the upper surface of the bottom wall for drainage toward a low point if the pool is to be emptied and rinsed/decontaminated at some time and due to installation tolerances). The base slab 6042 and sidewalls 6041 may be formed of reinforced concrete in one non-limiting embodiment. The fuel pool base slab 6042 may be formed in and rest on the soil sub-grade 6026 the top surface of which defines grade G. In this embodiment illustrated in the present application, the sidewalls are elevated above grade. In other possible embodiments contemplated, the base slab 6042 and sidewalls 6041 may alternatively be buried in sub-grade 6026 which surrounds the outer surfaces of the sidewalls. Either arrangement may be used and does not limit of the invention. In one embodiment, the spent fuel pool 6040 may have a rectilinear shape in top plan view. Four sidewalls 6041 may be provided in which the pool has an elongated rectangular shape (in top plan view) with two longer opposing sidewalls and two shorter opposing sidewalls (e.g. end walls). Other configurations of the fuel pool 6040 are possible such as square shapes, other polygonal shapes, and non-polygonal shapes. The sidewalls 6041 and base slab 6042 of the spent fuel pool 6040 define a cavity 6043 configured to hold cooling pool water W and a plurality of submerged nuclear spent fuel assembly storage racks 6027 holding fuel bundles or assemblies 6028 each containing multiple individual nuclear spent fuel rods. The storage racks 6027 are disposed on the base slab 6042 in typical fashion. With continuing reference to FIGS. 58-62, the spent fuel pool 6040 extends from an operating deck 6022 surrounding the spent fuel pool 6040 downwards to a sufficient depth D1 to submerge the fuel assemblies 6028 (see, e.g. FIG. 62) beneath the surface level S of the pool water W for proper radiation shielding purposes. In one implementation, the fuel pool may have a depth such that at least 10 feet of water is present above the top of the fuel assembly. A nuclear fuel assembly storage rack 6027 is shown in FIGS. 58 and 59, and further described in commonly assigned U.S. patent application Ser. No. 14/367,705 filed Jun. 20, 2014, which is incorporated herein by reference in its entirety. The storage rack 6027 contains a plurality of vertically elongated individual cells as shown each configured for holding a fuel assembly 6028 comprising a plurality of individual nuclear fuel rods. An elongated fuel assembly 6028 is shown in FIG. 62 holding multiple fuel rods 6028a and further described in commonly assigned U.S. patent application Ser. No. 14/413,807 filed Jul. 9, 2013, which is incorporated herein by reference in its entirety. Typical fuel assemblies 6028 for a pressurized water reactor (PWR) may each hold over 150 fuel rods in 10×10 to 17×17 fuel rod grid arrays per assembly. The assemblies may typically be on the order of approximately 14 feet high weighing about 1400-1500 pounds each. The substantially horizontal operating deck 6022 that circumscribes the sidewalls 6041 and pool 6040 on all sides in one embodiment may be formed of steel and/or reinforced concrete. The surface level of pool water W (i.e. liquid coolant) in the pool 6040 may be spaced below the operating deck 6022 by a sufficient amount to prevent spillage onto the deck during fuel assembly loading or unloading operations and to account to seismic event. In one non-limiting embodiment, for example, the surface of the operating deck 6022 may be at least 5 feet above the maximum 100 year flood level for the site in one embodiment. The spent fuel pool 6040 extending below the operating deck level may be approximately 40 feet or more deep (e.g. 42 feet in one embodiment). The fuel pool is long enough to accommodate as many spent fuel assemblies as required. In one embodiment, the fuel pool 6040 may be about 60 feet wide. There is sufficient operating deck space around the pool to provide space for the work crew and for staging necessary tools and equipment for the facility's maintenance. There may be no penetrations in the spent fuel pool 6040 within the bottom 30 feet of depth to prevent accidental draining of water and uncovering of the spent fuel. According to one aspect of the invention, a spent fuel pool liner system comprising a double liner is provided to minimize the risk of pool water leakage to the environment. The liner system is further designed to accommodate cooling water leakage collection and detection/monitoring to indicate a leakage condition caused by a breach in the integrity of the liner system. The liner system comprises a first outer liner 6060 separated from a second inner liner 6061 by an interstitial space 6062 formed between the liners. An outside surface of liner 6060 is disposed against or at least proximate to the inner surface 6063 of the fuel pool sidewalls 6041 and opposing inside surface is disposed proximate to the interstitial space 6062 and outside surface of liner 6061. The inside surface of liner 6061 is contacted and wetted by the fuel pool water W. It bears noting that placement of liner 6060 against liner 6061 without spacers therebetween provides a natural interstitial space of sufficient width to allow the space and any pool leakage there-into to be evacuated by a vacuum system, as further described herein. The natural surface roughness of the materials used to construct the liners and slight variations in flatness provides the needed space or gap between the liners. In other embodiments contemplated, however, metallic or non-metallic spacers may be provided which are distributed in the interstitial space 6062 between the liners if desired. The liners 6060, 6061 may be made of any suitable metal which is preferably resistant to corrosion, including without limitation stainless steel, aluminum, or other. In some embodiments, each liner may be comprised of multiple substantially flat metal plates which are seal welded together along their peripheral edges to form a continuous liner system encapsulating the sidewalls 6041 and base slab 6042 of the spent fuel pool 6040. The inner and outer liners 6061, 6060 may have the same or different thicknesses (measured horizontally or vertically between major opposing surfaces of the liners depending on the position of the liners). In one embodiment, the thicknesses may be the same. In some instances, however, it may be preferable that the inner liner 6061 be thicker than the outer liner 6060 for potential impact resistant when initially loading empty fuel storage racks 6027 into the spent fuel pool 6040. The outer and inner liners 6060, 6061 (with interstitial space therebetween) extend along the vertical sidewalls 6041 of the spent fuel pool 6040 and completely across the horizontal base slab 6042 in one embodiment to completely cover the wetted surface area of the pool. This forms horizontal sections 6060b, 6061b and vertical sections 6060a, 6061a of the liners 6060, 6061 to provide an impervious barrier to out-leakage of pool water W from spent fuel pool 6040. The horizontal sections of liners 6060b, 6061b on the base slab 6042 may be joined to the vertical sections 6060a, 6061a along the sidewalls 6041 of the pool 6040 by welding. The detail in FIG. 60 shows one or many possible constructions of the bottom liner joint 6064 comprising the use of seal welds 6065 (e.g. illustrated fillet welds or other) to seal sections 6060a to 6060b along their respective terminal edges and sections 6061a to 6061b along their respective terminal edges as shown. Preferably, the joint 6064 is configured and arranged to fluidly connect the horizontal interstitial space 6064 between horizontal liner sections 6060b, 6061b to the vertical interstitial space 6064 between vertical liner sections 6060a, 6061a for reasons explained elsewhere herein. The top liner joint 6065 in one non-limiting embodiment between the top terminal edges 6060c, 6061c of the vertical liner sections 6060a, 6061a is shown in the detail of FIG. 61. The top of the spent fuel pool 6040 is equipped with a substantially thick metal embedment plate 6070 which circumscribes the entire perimeter of the fuel pool. The embedment plate 6070 may be continuous in one embodiment and extends horizontally along the entire inner surface 6063 of the sidewalls 6041 at the top portion of the sidewalls. The embedment plate 6070 has an exposed portion of the inner vertical side facing the pool which extends above the top terminal ends 6060c, 6061c of the inner and outer liners 6060, 6061. The opposing outer vertical side of the plate 6070 is embedded entirely into the sidewalls 6041. A top surface 6071 of the embedment plate 6070 that faces upwards may be substantially flush with the top surface 6044 of the sidewalls 6041 to form a smooth transition therebetween. In other possible implementations, the top surface 6071 may extend above the top surface 6044 of the sidewalls. The embedment plate 6070 extends horizontal outward from the fuel pool 6040 for a distance into and less than the lateral width of the sidewalls 6041 as shown. The embedment plate 6070 has a horizontal thickness greater than the horizontal thickness of the inner liner 6061, outer liner 6060, and in some embodiments both the inner and outer liners combined. The top embedment plate 6070 is embedded into the top surface 6044 of the concrete sidewalls 6041 has a sufficient vertical depth or height to allow the top terminal edges 6060c, 6061c of liners 6060, 6061 (i.e. sections 6060a and 6061a respectively) to be permanently joined to the plate. The top terminal edges of liners 6060, 6061 terminate at distances D2 and D1 respectively below a top surface 6071 of the embedment plate 6070 (which in one embodiment may be flush with the top surface of the pool sidewalls 6041 as shown). Distance D1 is less than D2 such that the outer liner 6060 is vertical shorter in height than the inner liner 6061. In one embodiment, the embedment plate 6070 has a bottom end which terminates below the top terminal edges 6060c, 6061c of the liners 6060, 6061 to facilitate for welding the liners to the plate. In various embodiments, the embedment plate 6070 may be formed of a suitable corrosion resistant metal such as stainless steel, aluminum, or another metal which preferably is compatible for welding to the metal used to construct the outer and inner pool liners 6060, 6061 without requiring dissimilar metal welding. As best shown in FIG. 61, the top terminal edges 6060c, 6061c of inner and outer liners 6060, 6061 may have a vertically staggered arranged and be separately seal welded to the top embedment plate 6070 independently of each other. A seal weld 6066 couples the top terminal edge 6061c of liner 6061 to the exposed portion of the inner vertical side of the embedment plate 6070. A second seal weld 6067 couples the top terminal edge 6060c of liner 6060 also to the exposed portion of the inner vertical side of the embedment plate 6070 at a location below and spaced vertical apart from seal weld 6066. This defines a completely and hermetically sealed enclosed flow plenum 6068 that horizontal circumscribes the entire perimeter of the spent fuel pool 6040 in one embodiment. The flow plenum 6068 is in fluid communication with the interstitial space 6062 as shown. One vertical side of the flow plenum is bounded by a portion of inner liner 6061 and the opposing vertical side of the plenum is bounded by the inner vertical side of the top embedment plate 6070. The top flow plenum 6068 may be continuous or discontinuous in some embodiments. Where discontinuous, it is preferable that a flow passageway 6105 in the top embedment plate 6070 be provided for each section of the separate passageways. Seal welds 6066 and 6067 may be any type of suitable weld needed to seal the liners 6060, 6061 to the top embedment plate 6070. Backer plates, bars, or other similar welding accessories may be used to make the welds as needed depending on the configuration and dimensions of the welds used. The invention is not limited by the type of weld. In one embodiment, the outer and inner liners 6060, 6061 are sealably attached to the spent fuel pool 6040 only at top embedment plate 6070. The remaining portions of the liners below the embedment plate may be in abutting contact with the sidewalls 6041 and base slab 6042 without means for fixing the liners to these portions. It bears noting that at least the inner liner 6061 has a height which preferably is higher than the anticipated highest water level (surface S) of the pool water W in one embodiment. If the water level happens to exceed that for some reason, the top embedment plate 6070 will be wetted directly by the pool water and contain the fluid to prevent overflowing the pool onto the operating deck 6022. According to another aspect of the invention, a vapor extraction or vacuum system 6100 is provided that is used to draw down the air pressure in the interstitial space between the outer and inner liners 6060, 6061 to a relatively high state of vacuum for leakage control and/or detection. FIG. 63 is a schematic diagram of one embodiment of a vacuum system 6100. Referring to FIGS. 61 and 63, vacuum system 6100 generally includes a vacuum pump 6101 and a charcoal filter 6102. Vacuum pump 6101 may be any suitable commercially-available electric-driven vacuum pump capable of creating a vacuum or negative pressure within the interstitial space 6062 between the pool liners 6060 and 6061. The vacuum pump 6101 is fluidly connected to the interstitial space 6068 via a suitable flow conduit 6103 which is fluidly coupled to a telltale or flow passageway 6105 extending from the top surface 6071 of the top embedment plate 6070 to the top flow plenum 6068 formed between the pool liners 6060 and 6061. Flow conduit 6103 may be formed of any suitable metallic or non-metallic tubing or piping capable of withstanding a vacuum. A suitably-configured fluid coupling 6104 may be provided and sealed to the outlet end of the flow passageway 6105 for connecting the flow conduit 6103. The inlet end of the flow passageway penetrates the inner vertical side of top embedment plate 6070 within the flow plenum 6068. The flow passageway 6105 and external flow conduit 6103 provides a contiguous flow conduit that fluidly couples the flow plenum 6068 to the vacuum pump 6101. A one-way check valve is disposed between the flow plenum 6105 and the suction inlet of the vacuum pump 6101 to permit air and/or vapor to flow in a single direction from the liner system to the pump. The absolute pressure maintained by the vacuum system 6100 in the interstitial space 6062 between the liners 6060, 6061 (i.e. “set pressure”) preferably should be such that the bulk water temperature of the spent fuel pool 6040 which is heated by waste decay heat generated from the fuel rods/assemblies is above the boiling temperature of water at the set pressure. The table below provides the boiling temperature of water at the level of vacuum in inches of mercury (Hg) which represent some examples of set pressures that may be used. Pressure in inch, HgABoiling Temp, deg F.1792101311541255133 Any significant rise in pressure would indicate potential leakage of water in the interstitial space 6062 between the liners 6060, 6061. Because of sub-atmospheric conditions maintained by the vacuum pump 6101 in the interstitial space, any water that may leak from the pool into this space through the inner liner 6061 would evaporate, causing the pressure to rise which may be monitored and detected by a pressure sensor 6104. The vacuum pump 6101 preferably should be set to run and drive down the pressure in the interstitial space 6062 to the “set pressure.” In operation as one non-limiting example, if the vacuum pump 6101 is operated to create a negative pressure (vacuum) in the interstitial space 6062 of 2 inches of Hg, the corresponding boiling point of water at that negative pressure is 101 degrees Fahrenheit (degrees F.) from the above Table. If the bulk water temperature of pool water W in the spent fuel pool 6040 were at any temperature above 101 degrees F. and leakage occurred through the inner pool liner 6061 into the interstitial space 6062, the liquid leakage would immediately evaporate therein creating steam or vapor. The vacuum pump 6101 withdraws the vapor through the flow plenum 6068, flow passageway 6105 in the top embedment plate 6070, and flow conduit 6103 (see, e.g. directional flow arrows of the water vapor in FIGS. 61 and 63). Pressure sensor 6104 disposed on the suction side of the pump 6101 would detect a corresponding rise in pressure indicative of a potential leak in the liner system. In some embodiments, the pressure sensor 6104 may be operably linked to a control panel of a properly configured computer processor based plant monitoring system 6107 which monitors and detects the pressure measured in the interstitial space 6062 between the liners on a continuous or intermittent basis to alert operators of a potential pool leakage condition. Such plant monitoring systems are well known in the art without further elaboration. The extracted vapor in the exhaust or discharge from the vacuum pump 6101 is routed through a suitable filtration device 6102 such as a charcoal filter or other type of filter media before discharge to the atmosphere, thereby preventing release of any particulate contaminants to the environment. Advantageously, it bears noting that if leakage is detected from the spent fuel pool 6040 via the vacuum system 6100, the second outer liner 6060 encapsulating the fuel pool provides a secondary barrier and line of defense to prevent direct leaking of pool water W into the environment. It bears noting that there is no limit to the number of vapor extraction systems including a telltale passageway, vacuum pump, and filter combination with leakage monitoring/detection capabilities that may be provided. In some instances, four independent systems may provide adequate redundancy. In addition, it is also recognized that a third or even fourth layer of liner may be added to increase the number of barriers against leakage of pool water to the environment. A third layer in some instances may be used as a palliative measure if the leak tightness of the first inter-liner space could not, for whatever reason, be demonstrated by a high fidelity examination in the field such as helium spectroscopy. VII. Inventive Concept 7 With reference to FIGS. 64-73, a seventh inventive concept will be described. Referring to FIG. 64, a fuel rack 7101 including an array of cells 7103 is shown. The array of cells 7103 is formed by slotted plates 7105 arranged in interlocking arrangement. In the embodiment shown, each storage cell 7107 in the array of cells 7103 has a square profile in plan view, with all the cells having the same dimensions. However, in certain embodiments, each storage cell 7107 in the array of cells 7103 may have an alternative profile shape, including a rectangular profile shape and a hexagonal profile shape, among others. In certain embodiments, the storage cells 7107 in the array of cells 7103 may vary in size. The fuel rack 7101 also includes tie members 7109 affixed to the array of cells 7103 to extend along the external surface of the array of cells 7103. The tie members extend substantially the entire height of the array of cells 7103 to provide vertical stiffness to the interlocking slotted plates 7105. In certain embodiments, the tie members 7109 may be located within the storage cells 7107 and affixed to the array of cells 7103. In still other embodiments, smaller coupling elements may be used which couple adjacent ones of the slotted plates 7105 together instead of the tie members 7109. The fuel rack 7101 also includes a base plate 7111, and the array of cells 7103 is connected to a top surface 7115 of the base plate 7111. Support pedestals 7113 are coupled to the bottom surface 7117 of the base plate 7111. The support pedestals 7113 provide space underneath the base plate 7111 for the circulation of fluid up and through the array of cells 7103. An exploded version of the fuel rack 7101 is shown in FIG. 65. The array of cells 7103 is shown separated into a top portion 7121, a middle portion 7123, and a bottom portion 7125. The entire array of cells 7103 may be formed out of four different types of slotted plates. A plurality of first slotted plates 7131 are slidably interlocked with one another to form the top portion 7121 of the array of cells 7103; a plurality of second slotted plates 7133 are slidably interlocked with one another to form the middle portion 7123 of the array of cells 7103; and a plurality of third slotted plates 7135 are slidably interlocked with one another to form the top portion 7125 of the array of cells 7103. Each of the plurality of first, second, and third slotted plates 7131, 7133, 7135 include one or more of the types of slotted plates shown in FIGS. 67A-D. As shown, in the top portion 7121, the plurality of first slotted plates 7131 includes a plurality of top slotted plates 7141 (FIG. 67A) and a plurality of middle slotted plates 7143 (FIG. 67B); in the middle portion 7123, the plurality of second slotted plates 7133 includes a plurality of the middle slotted plates 7143 (FIG. 67B); and in the bottom portion 7125, the plurality of third slotted plates 7135 includes a plurality of bottom half slotted plates 7145 (FIG. 67C) and a plurality of bottom full slotted plates 7147 (FIG. 67D). The plurality of first slotted plates 7131 and the plurality of third slotted plates 7135 are constructed from a first material, and the plurality of second slotted plates 7133 are constructed from a second material which is metallurgically incompatible with the first material. As used herein, the term “metallurgically incompatible” means that the two materials are not compatible to the extent that they cannot be joined by a weld. The inability to join two materials by a weld arises from the state of the art of welding, in which no weld material and/or no technique are known to exist that could be used to weld the two materials together. In certain embodiments, the first material may be stainless steel and the second material may be a metal matrix composite material. The metal matrix composite material may be, in certain embodiments, a aluminum/boron carbide metal matrix composite material, an non-limiting example of which is a boron impregnated aluminum. One such suitable material for the metal matrix composite material is sold under the tradename Metamic®. The tie members 7109, the base plate 7111, and the pedestals 7113, in certain embodiments, are also formed from the first material. The plurality of first slotted plates 7131 of the top portion 7121 are welded together along adjacent edges. Welding the plurality of first slotted plates 7131 provides overall structure to the top portion 7121 of the array of cells 7103. The plurality of third slotted plates 7135 of the bottom portion 7125 are coupled to the base plate 7111. In certain embodiments, the plurality of third slotted plates 7135 may be welded to the base plate 7111. By welding the plurality of third slotted plates 7135 to the base plate 7111, the base plate 7111 is provided with additional flexural strength, which may be needed when the storage rack 7101 is loaded with fuel assemblies. In certain embodiments, the plurality of third slotted plates 7135 may also be welded together along adjacent edges. Conventional welding materials and processes may be used for these welds when the first material is stainless steel. The plurality of second slotted plates 7133 may be welded together at intersecting slots, insofar as a welding process is known for the second material. When the second material is one such as Metamic®, welding may be performed as taught in WO2014106044, published Jul. 3, 2014 and entitled “Joining process for neutron absorbing materials.” The tie members 7109 extend along an external surface 7119 of the array of cells 7103 and are affixed to the top portion 7121 and the bottom portion 7125 of the array of cells 7103. Particularly, the tie members 7109 are affixed to one or more of the plurality of first slotted plates 7131 and to one or more of the plurality of first slotted plates 7135 that are outward-facing. The tie members 7109 may be affixed to the top portion 7121 and the bottom portion 7125 by welding. The tie members 7109 therefore need not be directly affixed to any of the plurality of second slotted plates 7133 in the middle portion 7123 of the array of cells 7103 to stabilize the entire array of cells 7103. In certain embodiments, fasteners such as screws and/or brackets may couple the tie members 7109 to the top portion 7121 and/or the bottom portion 7125 of the array of cells 7103. The tie members 7109 serve to provide vertical stiffness to the array of cells 7103. As indicated above, because the second plurality of slotted plates 7133 is made from a second material that is metallurgically incompatible with the first material of the first and third plurality of slotted plates 7131, 7135, the middle portion 7123 cannot be welded to the top or bottom portions 7121, 7125 of the array of cells 7103. Thus, by using the tie members 7109 to tie the top and bottom portions 7121, 7125 of the array of cells 7103 together, the second plurality of slotted plates 7133 in the middle portion 7123 of the array of cells 7103 may be securely held in place, and additional stiffness is thereby provided to the entire array of cells 7103 and to the fuel rack 7101 itself. As shown, the tie members 7109 are affixed to corners of the array of cells 7103, and only four tie members 7109 are shown in the depicted embodiment. In certain embodiments, the tie members 7109 may be affixed at different locations on the array of cells 7103. And in certain embodiments, more or fewer tie members 7109 may be used. A middle segment 7161 of the middle portion 7123 of the array of cells 7103 is shown in FIG. 66. Each middle segment 7161 of the array of cells 7103 comprises a gridwork of the middle slotted plates 7143 arranged in a rectilinear configuration so as to form a vertical portion of the storage cells 7107. In creating the middle segment 7161, a first middle slotted plate 7143 is arranged vertically. A second middle slotted plate 7143 is then arranged above and at a generally 90 degree angle to the first middle slotted plate 7143 so that the corresponding slots 7163 of the two middle slotted plates 7143 are aligned. The second middle slotted plate 7143 is then lowered onto the first middle slotted plate 7143, thereby causing the slots 7163 to interlock as illustrated. This is repeated with all middle slotted plate 7143 until the desired rectilinear configuration is created, thereby creating the middle segment 7161. The entire fuel rack body is formed out of three types of slotted plates, a top slotted plate 7141, a middle slotted plate 7143, a bottom half slotted plate 7145, and a bottom full slotted plate 7147, which are respectively shown in FIGS. 67A-D. The top slotted plate 7141 is formed as half of the middle slotted plate 7143. Similarly, the bottom half slotted plate 7145 is formed as half of the middle slotted plate 7143 with the cut outs 7165 added along the remaining slotted edge. The bottom full slotted plate 7147 is formed the same as the middle slotted plate 7143, but with the cut outs 7165 added along one slotted edge. The cut outs 7165 serve as auxiliary flow holes for facilitating thermosiphon flow into the storage cells 7107 as discussed above. The top slotted plate 7141 and the bottom half slotted plate 7145 are only used at the top and bottom, respectively, of the array of cells 7103 to cap the middle segments 7161 (FIG. 66) so that the array of cells 7103 has level top and bottom edges. Each of the slotted plates 7141-7147 includes a plurality of slots 7163, end tabs 7167, and indentations 7169 adjacent the end tabs 7167, all of which are strategically arranged to facilitate sliding assembly to create the array of cells 7103. The slots 7163 are provided in one or both of the top and bottom edges of the plates 7141-7147. The slots 7163 included on the top edges of the plates 7141-7147 are aligned with the slots 7163 included on the bottom edges of that same plate 7141-7147. The slots 7163 extend through the plates 7141-7147 for about one-fourth of the height of the plates 7141-7147. The end tabs 7167 extend from lateral edges of the plates 7141-7147 and are about one-half of the height of the plates 7141-7147. The end tabs 7167 slidably mate with the indentations 7169 in the lateral edges of adjacent plates 7141-7147 that naturally result from the existence of the tabs 7167. By way of example, in creating a middle segment 7161 of the array of cells 7103, the slots 7163 and end tabs 7167 of the middle segment 7161 interlock with adjacent middle segments 7161 so as to prohibit relative horizontal and rotational movement between the adjacent middle segments 7161. The middle segments 7161 intersect and interlock with one another to form a stacked assembly that is the array of cells 7103. The array of cells 7103 may include any number of the middle segments 7161, with the height of the middle segments 7161 in the middle portion 7123 of the array of cells 7103 being constructed so that the fuel storage section of a fuel assembly may be entirely located within the middle portion 7123 of the array of cells 7103. The entire array of cells 7103 may thus be formed of slotted plates 7141-7147 having base configuration, which is the configuration of the middle slotted plate 7143, with the top slotted plate 7141, the bottom half slotted plate 7145, and the bottom full slotted plate 7147 being formed by additional minor modifications of the base configuration. The profile of a fuel assembly 7181, used for the storage of nuclear fuel 7183, is shown in FIG. 68 positioned within a storage cell 7107 of the array of cells 7103. The fuel assembly 7181 includes a top section 7185, a middle section 7187, and a bottom section 7189. The nuclear fuel 7183 is only stored within the middle section 7187 of the fuel assembly 7181. The top and bottom sections 7185, 7189 do not have any nuclear fuel storage capabilities, and thus no nuclear fuel is stored within the top or bottom sections 7185, 7189. As shown, the middle section 7187 of the fuel assembly 7181 is stored entirely within the middle portion 7123 of the storage cell 7107. Thus, the middle section 7187 and the nuclear fuel 7183 are entirely surrounded on 4 sides with the neutron absorbing material from which the slotted plates 7143 of the middle portion 7123 are constructed. The base plate 7111, which is shown in FIG. 69, includes a plurality of flow holes 7201 extending through the base plate 7111 from the bottom surface 7117 to the top surface 7115. The base plate 7111 also includes four oblong holes 7203 (second row in from the corners) for lifting and installing the fuel rack 7101 within the storage pool. Typically, a special lifting beam with four long reach rods is used to interact with the oblong holes 7203 to grapple the fuel rack 7101 for transfer into or out of, or movement within, the storage pool. The flow holes 7201 (and oblong holes 7203) create passageways from below the base plate 7111 into the bottom ends of the storage cells 7107. As shown, a single flow hole 7201 is provided for each storage cell 7107. In certain embodiments, multiple flow holes 7201 may be provided for each storage cell 7107 to provide cooling fluid to the storage cell 7107. The flow holes 7201 serve as fluid inlets to facilitate natural thermosiphon flow of pool water through the storage cells 7107 when fuel assemblies having a heat load are positioned therein. More specifically, when heated fuel assemblies are positioned in the storage cells 7107 in a submerged environment, the water within the storage cells 7107 surrounding the fuel assemblies becomes heated, thereby rising due to increased buoyancy. As this heated water rises and exits the storage cells 7107 via their open top ends, cool water is drawn into the bottom of the storage cells 7107 via the flow holes 7201. This heat induced water flow along the fuel assemblies then continues naturally. A support pedestal 7113 for the fuel rack 7101 is shown in FIG. 70. The support pedestals 7113 affixed to the bottom surface 7115 of the base plate 7111 ensure that a space exists between a floor of a storage pool and the bottom surface 7115 of the base plate 7111, thereby creating an inlet plenum for water to flow through the flow holes 7201. The support pedestal 7113 includes a base portion 7211 and a riser portion 7213 formed about an interior flow space 7215. The riser portion 7213 includes flow apertures 7217 through which water from the storage pool may pass from a space external to the support pedestal 7113 into the interior flow space 7215. Water passing into the interior flow space 7215 may then pass up through a flow hole 7201 in the base plate 7111 to enable the cooling process described above. Although the riser portion 7213 is depicted as being annular, in certain embodiments the riser portion 7213 may have any geometrical configuration which supports the base plate 7111 above the floor of the storage pool and permits water from the storage pool to flow into any flow holes 7201 in the base plate 7111 near which the support pedestal 7113 may be affixed. Another embodiment of a fuel rack 7301 including an array of cells 7303 is shown in FIG. 71. The array of cells 7303 is formed by slotted plates 7305 arranged in interlocking arrangement. In the embodiment shown, each storage cell 7307 in the array of cells 7303 has a square profile in plan view, with all the cells having the same dimensions. However, in certain embodiments, each storage cell 7307 in the array of cells 7303 may have an alternative profile shape, including a rectangular profile shape and a hexagonal profile shape, among others. In certain embodiments, the storage cells 7307 in the array of cells 7303 may vary in size. The slotted plates 7305 are also arranged so that flux traps 7309 are formed around the entire profile of each interior storage cell 7307a. The external walls of each exterior storage cell 7307b does not include flux traps. The fuel rack 7301 also includes tie members 7311 affixed to the array of cells 7303 to extend along the external surface of the array of cells 7303. The tie members extend substantially the entire height of the array of cells 7303 to provide vertical stiffness to the interlocking slotted plates 7305. In certain embodiments, the tie members 7311 may be located within the storage cells 7307 and affixed to the array of cells 7303. In still other embodiments, smaller coupling elements may be used which couple adjacent ones of the slotted plates 7305 together instead of the tie members 7311. The fuel rack 7301 also includes a base plate 7313, and the array of cells 7303 is connected to a top surface 7317 of the base plate 7313. Support pedestals 7315 are coupled to the bottom surface 7319 of the base plate 7313. The support pedestals 7315 provide space underneath the base plate 7313 for the circulation of fluid up and through the array of cells 7303. The array of cells 7303 is shown separated into a top portion 7331, a middle portion 7333, and a bottom portion 7335. The entire array of cells 7303 may be formed out of four different types of slotted plates. A plurality of first slotted plates 7341 are slidably interlocked with one another to form the top portion 7331 of the array of cells 7303; a plurality of second slotted plates 7343 are slidably interlocked with one another to form the middle portion 7333 of the array of cells 7303; and a plurality of third slotted plates 7345 are slidably interlocked with one another to form the top portion 7335 of the array of cells 7303. Each of the plurality of first, second, and third slotted plates 7341, 7343, 7345 include one or more of the types of slotted plates shown in FIGS. 73A-D. In the top portion 7331, the plurality of first slotted plates 7341 includes a plurality of top slotted plates 7351 (FIG. 73A) and a plurality of middle slotted plates 7353 (FIG. 73B); in the middle portion 7333, the plurality of second slotted plates 7343 includes a plurality of the middle slotted plates 7353 (FIG. 73B); and in the bottom portion 7335, the plurality of third slotted plates 7345 includes a plurality of bottom half slotted plates 7355 (FIG. 73C) and a plurality of bottom full slotted plates 7357 (FIG. 73D). The plurality of first slotted plates 7341 and the plurality of third slotted plates 7345 are constructed from a first material, and the plurality of second slotted plates 7343 are constructed from a second material which is metallurgically incompatible with the first material. In certain embodiments, the first material may be stainless steel and the second material may be a metal matrix composite material. The metal matrix composite material may be, in certain embodiments, a aluminum/boron carbide metal matrix composite material, an non-limiting example of which is a boron impregnated aluminum, such as the metal matrix composite material sold under the tradename Metamic®. The tie members 7311, the base plate 7313, and the pedestals 7315, in certain embodiments, are also formed from the first material. The plurality of first slotted plates 7341 of the top portion 7331 are welded together along adjacent edges. Welding the plurality of first slotted plates 7341 provides overall structure to the top portion 7331 of the array of cells 7303. The plurality of third slotted plates 7345 of the bottom portion 7335 are coupled to the base plate 7313. In certain embodiments, the plurality of third slotted plates 7345 may be welded to the base plate 7313. By welding the plurality of third slotted plates 7345 to the base plate 7313, the base plate 7313 is provided with additional flexural strength, which may be needed when the storage rack 7301 is loaded with fuel assemblies. In certain embodiments, the plurality of third slotted plates 7345 may also be welded together along adjacent edges. Conventional welding materials and processes may be used for these welds when the first material is stainless steel. The plurality of second slotted plates 7343 may be welded together at intersecting slots, insofar as a welding process is known for the second material. The tie members 7311 extend along an external surface 7321 of the array of cells 7303 and are affixed to the top portion 7331 and the bottom portion 7335 of the array of cells 7303. Particularly, the tie members 7311 are affixed to one or more of the plurality of first slotted plates 7341 and to one or more of the plurality of first slotted plates 7345 that are outward-facing. The tie members 7311 may be affixed to the top portion 7331 and the bottom portion 7335 by welding. The tie members 7311 therefore need not be directly affixed to any of the plurality of second slotted plates 7343 in the middle portion 7333 of the array of cells 7303 to stabilize the entire array of cells 7303. In certain embodiments, fasteners such as screws and/or brackets may couple the tie members 7311 to the top portion 7331 and/or the bottom portion 7335 of the array of cells 7303. As shown, the tie members 7311 are affixed to corners of the array of cells 7303, and only four tie members 7311 are shown in the depicted embodiment. In certain embodiments, the tie members 7311 may be affixed at different locations on the array of cells 7303. And in certain embodiments, more or fewer tie members 7311 may be used. A middle segment 7361 of the middle portion 7333 of the array of cells 7303 is shown in FIG. 72. Each middle segment 7361 of the array of cells 7303 comprises a gridwork of the middle slotted plates 7353 arranged in a rectilinear configuration so as to form a vertical portion of the storage cells 7307 and the flux traps 7309. In creating the middle segment 7361, a first middle slotted plate 7353 is arranged vertically. A second middle slotted plate 7353 is then arranged above and at a generally 90 degree angle to the first middle slotted plate 7353 so that the corresponding slots 7363 of the two middle slotted plates 7353 are aligned. The second middle slotted plate 7353 is then lowered onto the first middle slotted plate 7353, thereby causing the slots 7363 to interlock. This is repeated with all middle slotted plate 7353 until the desired rectilinear configuration is created, thereby creating the middle segment 7361 having the storage cells 7307 and the flux traps 7309. The entire fuel rack body is formed out of three types of slotted plates, a top slotted plate 7351, a middle slotted plate 7353, a bottom half slotted plate 7355, and a bottom full slotted plate 7357, which are respectively shown in FIGS. 73A-D. The top slotted plate 7351 is formed as half of the middle slotted plate 7353. Similarly, the bottom half slotted plate 7355 is formed as half of the middle slotted plate 7353 with the cut outs 7365 added along the remaining slotted edge. The bottom full slotted plate 7357 is formed the same as the middle slotted plate 7353, but with the cut outs 7365 added along one slotted edge. The cut outs 7365 serve as auxiliary flow holes for facilitating thermosiphon flow into the storage cells 7307 as discussed above. The top slotted plate 7351 and the bottom half slotted plate 7355 are only used at the top and bottom, respectively, of the array of cells 7303 to cap the middle segments 7361 (FIG. 72) so that the array of cells 7303 has level top and bottom edges. Each of the slotted plates 7351-7357 includes a plurality of slots 7363, end tabs 7367, and indentations 7369 adjacent the end tabs 7367, all of which are strategically arranged to facilitate sliding assembly to create the array of cells 7303. The slots 7363 are provided in one or both of the top and bottom edges of the plates 7351-7357. The slots 7363 included on the top edges of the plates 7351-7357 are aligned with the slots 7363 included on the bottom edges of that same plate 7351-7357. The slots 7363 extend through the plates 7351-7357 for about one-fourth of the height of the plates 7351-7357. The end tabs 7367 extend from lateral edges of the plates 7351-7357 and are about one-half of the height of the plates 7351-7357. The end tabs 7367 slidably mate with the indentations 7369 in the lateral edges of adjacent plates 7351-7357 that naturally result from the existence of the tabs 7367. By way of example, in creating a middle segment 7361 of the array of cells 7303, the slots 7363 and end tabs 7367 of the middle segment 7361 interlock with adjacent middle segments 7361 so as to prohibit relative horizontal and rotational movement between the adjacent middle segments 7361. The middle segments 7361 intersect and interlock with one another to form a stacked assembly that is the array of cells 7303. The array of cells 7303 may include any number of the middle segments 7361, with the height of the middle segments 7361 in the middle portion 7333 of the array of cells 7303 being constructed so that the fuel storage section of a fuel assembly may be entirely located within the middle portion 7333 of the array of cells 7303. The entire array of cells 7303 may thus be formed of slotted plates 7351-7357 having base configuration, which is the configuration of the middle slotted plate 7353, with the top slotted plate 7351, the bottom half slotted plate 7355, and the bottom full slotted plate 7357 being formed by additional minor modifications of the base configuration. Furthermore, as a result of the interlocking nature of the slotted plates 7351-7357, spacers are not needed to maintain the flux traps 7309. Thus, in certain embodiments, the array of cells 7303 may be free of spacers in the flux traps 7309. VIII. Inventive Concept 8 With reference to FIGS. 74-85, an eighth inventive concept will be described. Referring to FIGS. 74-78, a nuclear facility which may be a nuclear generating plant includes a fuel pool 8040 according to the present disclosure configured for storing a plurality of nuclear fuel racks 8100. The fuel pool 8040 may comprise a plurality of vertical sidewalls 8041 rising upwards from an adjoining substantially horizontal bottom base wall or slab 8042 (recognizing that some slope may intentionally be provided in the upper surface of the base slab for drainage toward a low point if the pool is to be emptied and rinsed/decontaminated at some time and due to installation tolerances). The base slab 8042 and sidewalls 8041 may be formed of reinforced concrete in one non-limiting embodiment. The fuel pool base slab 8042 may be formed in and rest on the soil sub-grade 8026, the top surface of which defines grade G. In this embodiment illustrated in the present application, the sidewalls are elevated above grade. The base slab 8042 may be located at grade G as illustrated, below grade, or elevated above grade. In other possible embodiments contemplated, the base slab 8042 and sidewalls 8041 may alternatively be buried in sub-grade 8026 which surrounds the outer surfaces of the sidewalls. Any of the foregoing arrangements or others may be used depending on the layout of the nuclear facility and does not limit of the invention. In one embodiment, the fuel pool 8040 may have a rectilinear shape in top plan view. Four sidewalls 8041 may be provided in which the pool has an elongated rectangular shape (in top plan view) with two longer opposing sidewalls and two shorter opposing sidewalls (e.g. end walls). Other configurations of the fuel pool 8040 are possible such as square shapes, other polygonal shapes, and non-polygonal shapes. The sidewalls 8041 and base slab 8042 of the fuel pool 8040 define an upwardly open well or cavity 8043 configured to hold cooling pool water W and the plurality of submerged nuclear fuel racks 8100 each holding multiple nuclear fuel bundles or assemblies 8028 (a typical one shown in phantom view seated in a fuel rack cell in FIG. 78). Each fuel assembly 8028 contains multiple individual new or spent uranium fuel rods. Fuel assemblies are further described in commonly assigned U.S. patent application Ser. No. 14/413,807 filed Jul. 9, 2013, which is incorporated herein by reference in its entirety. Typical fuel assemblies 8028 for a pressurized water reactor (PWR) may each hold over 150 fuel rods in 10×10 to 17×17 fuel rod grid arrays per assembly. The assemblies may typically be on the order of approximately 14 feet high weighing about 1400-1500 pounds each. The fuel racks 8100 storing the fuel assemblies are emplaced on the base slab 8042 in a high-density arrangement in the horizontally-abutting manner as further described herein. The fuel pool 8040 extends from an operating deck 8022 surrounding the fuel pool 8040 downwards to a sufficient vertical depth D1 to submerge the fuel assemblies 8028 in the fuel rack (see, e.g. FIG. 79) beneath the surface level S of the pool water W for proper radiation shielding purposes. The substantially horizontal operating deck 8022 that circumscribes the sidewalls 8041 and pool 8040 on all sides in one embodiment may be formed of steel and/or reinforced concrete. In one implementation, the fuel pool may have a depth such that at least 8010 feet of water is present above the top of the fuel assembly. Other suitable depths for the pool and water may be used of course. The surface level of pool water W (i.e. liquid coolant) in the pool 8040 may be spaced below the operating deck 8022 by a sufficient amount to prevent spillage onto the deck during fuel assembly loading or unloading operations and to account to seismic event. In one non-limiting embodiment, for example, the surface of the operating deck 8022 may be at least 5 feet above the maximum 100 year flood level for the site in one embodiment. The fuel pool 8040 extending below the operating deck level may be approximately 8040 feet or more deep (e.g. 42 feet in one embodiment). The fuel pool is long and wide enough to accommodate as many fuel racks 8100 and fuel assemblies 8028 stored therein as required. There is sufficient operating deck space around the pool to provide space for the work crew and for staging necessary tools and equipment for the facility's maintenance. There may be no penetrations in the fuel pool 8040 within the bottom 30 feet of depth to prevent accidental draining of water and uncovering of the fuel. In some embodiments, a nuclear fuel pool liner system may be provided to minimize the risk of pool water leakage to the environment. The liner system may include cooling water leakage collection and detection/monitoring to indicate a leakage condition caused by a breach in the integrity of the liner system. Liner systems are further described in commonly owned U.S. patent application Ser. No. 14/877,217 filed Oct. 7, 2015, which is incorporated herein by reference in its entirety. The liner system in one embodiment may comprise one or more liners 8060 attached to the inner surfaces 8063 of the fuel pool sidewalls 8041 and the base slab 8042. The inside surface 8061 of liner is contacted and wetted by the fuel pool water W. The liner 8060 may be made of any suitable metal of suitable thickness T2 which is preferably resistant to corrosion, including for example without limitation stainless steel, aluminum, or other. Typical liner thicknesses T2 may range from about and including 3/16 inch to 5/16 inch thick. Typical stainless steel liner plates include ASTM 240-304 or 304L. In some embodiments, the liner 8060 may be comprised of multiple substantially flat metal plates or sections which are hermetically seal welded together via seal welds along their contiguous peripheral edges to form a continuous liner system completely encapsulating the sidewalls 8041 and base slab 8042 of the fuel pool 8040 and impervious to the egress of pool water W. The liner 8060 extends around and along the vertical sidewalls 8041 of the fuel pool 8040 and completely across the horizontal base slab 8042 to completely cover the wetted surface area of the pool. This forms horizontal sections 8060b and vertical sections 8060a of the liner to provide an impervious barrier to out-leakage of pool water W from fuel pool 8040. The horizontal sections of liners 8060b on the base slab 8042 may be joined to the vertical sections 8060a along perimeter corner seams therebetween by hermetic seal welding. The liner 8060 may be fixedly secured to the base slab 8042 and sidewalls 8041 of the fuel pool 8040 by any suitable method such as fasteners. With continuing reference to FIGS. 74-78, the fuel rack 8100 is a cellular upright module or unit. Fuel rack 8100 may be a high density, tightly packed non-flux type rack as illustrated which is designed to be used with fuel assemblies that do not require the presence of a neutron flux trap between adjacent cells 8110. Thus, the inclusion of neutron flux traps (e.g. gaps) in fuel racks when not needed is undesirable because valuable fuel pool floor area is unnecessarily wasted. Of course, both non-flux and flux fuel rack types may be stored side by side in the same pool using the seismic-resistant fuel storage system according to the present disclosure. The invention is therefore not limited to use of any particular type of rack. Fuel rack 8100 defines a vertical longitudinal axis LA and comprises a grid array of closely packed open cells 8110 formed by a plurality of adjacent elongated storage tubes 8120 arranged in parallel axial relationship to each other. The rack comprises peripherally arranged outboard tubes 8120A which define a perimeter of the fuel rack and inboard tubes 8120B located between the outboard tubes. Tubes 8120 are coupled at their bottom ends 8114 to a planar top surface of a baseplate 8102 and extend upwards in a substantially vertical orientation therefrom. In this embodiment, the vertical or central axis of each tube 8120 is not only substantially vertical, but also substantially perpendicular to the top surface of the baseplate 8102. In one embodiment, tubes 8120 may be fastened to baseplate 8102 by welding and/or mechanical coupling such as bolting, clamping, threading, etc. Tubes 8120 include an open top end 8112 for insertion of fuel assemblies, bottom end 8114, and a plurality of longitudinally extending vertical sidewalls 8116 (“cell walls”) between the ends and defining a tube or cell height H1. Each tube 8120 defines an internal cell cavity 8118 extending longitudinally between the top and bottom ends 8112, 8114. In the embodiment shown in FIG. 75, four tube sidewalls 8116 arranged in rectilinear polygonal relationship are provided forming either a square or rectangular tube 8120 in lateral or transverse cross section (i.e. transverse or orthogonal to longitudinal axis LA) in plan or horizontal view (see also FIG. 76). Cells 8110 and internal cavities 8118 accordingly have a corresponding rectangular configuration in lateral cross section. The top ends of the tubes 8120 are open so that a fuel assembly can be slid down into the internal cavity 8118 formed by the inner surfaces of the tube sidewalls 8116. Each cell 8110 and its cavity 8118 are configured for holding only a single nuclear fuel assembly 8028. Tubes 8120 may be made of any suitable preferably corrosion resistant metal, such as without limitation stainless steel or others. Baseplate 8102 may be made of a similar or different corrosion resistant metal. It will be appreciated that each tube 8120 can be formed as a single unitary structural component that extends the entire desired height H1 or can be constructed of multiple partial height tubes that are vertically stacked and connected together such as by welding or mechanical means which collectively add up to the desired height H1. It is preferred that the height H1 of the tubes 8120 be sufficient so that the entire height of a fuel assembly may be contained within the tube when the fuel assembly is inserted into the tube. The top ends 8112 of tubes 8120 may preferably but not necessarily terminate in substantially the same horizontal plane (defined perpendicular to longitudinal axis LA) so that the tops of the tube are level with each other. The baseplate 8102 at the bottom ends 8114 of the tubes defines a second horizontal reference plane HR. As best shown in FIG. 75, tubes 8120 are geometrically arranged atop the baseplate 8102 in rows and columns along the Z-axis and X-axis respectively. Any suitable array size including equal or unequal numbers of tubes in each row and column may be provided depending on the horizontal length and width of the pool base slab 8042 and number of fuel racks 8100 to be provided. In some arrangements, some or all of the fuel racks 8100 may have unequal lateral width and lateral length as to best make use of a maximum amount of available slab surface area as possible for each installation. For convenience of reference, the outward facing sidewalls 8116 of the outboard tubes 8120A may be considered to collectively define a plurality of lateral sides 8130 of the fuel rack 8100 extending around the rack's perimeter as shown in FIG. 75. Referring to FIGS. 74-78, each fuel rack 8100 comprises a plurality of legs or pedestals 8200 which support rack from the base slab 8042 of the fuel pool 8040. Pedestals 8200 each have a preferably flat bottom end 8204 to engage the pool base slab 8042 and a top end 8202 fixedly attached to the bottom of the baseplate 8102. The pedestals 8200 protrude downwards from baseplate 8102. This elevates the baseplates 8102 of the rack off the base slab 8042, thereby forming a gap therebetween which defines a bottom flow plenum P beneath rack 8100. The plenum P allows cooling water W in the pool to create a natural convective circulation flow path through each of the fuel storage tubes 8120 (see e.g. flow directional arrows in FIG. 78). A plurality of flow holes 8115 are formed in the rack through baseplate 8102 in a conventional manner to allow cooling water to flow upwards through the cavity 8118 of each tube 8120 and outward through the open top ends 8112 of the tubes. Commonly owned U.S. patent application Ser. No. 14/367,705 filed Jun. 20, 2014 shows fuel rack baseplates with flow holes, and is incorporated herein by reference in its entirety. The pool water W flowing through the tubes is heated by the nuclear fuel in fuel assemblies, thereby creating the motive force driving the natural thermal convective flow scheme. Referring now then to FIGS. 76 and 78, flow holes 8115 create passageways from below the base plate 8102 into the cells 8110 formed by the tubes 8120. Preferably, a single flow hole 8115 is provided for each cell 8110, however, more may be used as needed to create sufficient flow through the tubes. The flow holes 8115 are provided as inlets to facilitate natural thermosiphon flow of pool water through the cells 8110 when fuel assemblies having a heat load are positioned therein. More specifically, when heated fuel assemblies are positioned in the cells 8110 in a submerged environment, the water within the cells 8110 surrounding the fuel assemblies becomes heated, thereby rising due to decrease in density and increased buoyancy creating a natural upflow pattern. As this heated water rises and exits the cells 8110 via the tube open top ends 8112 (see FIG. 74), cooler water is drawn into the bottom of the cells through the flow holes 8115. This heat induced water flow and circulation pattern along the fuel assemblies then continues naturally to dissipate heat generated by the fuel assemblies. Pedestals 8200 may therefore have a height selected to form a bottom flow plenum P of generally commensurate height to ensure that sufficient thermally-induced circulation is created to adequately cool the fuel assembly. In one non-limiting example, the height of the plenum P may be about 2 to 2.5 inches (including the listed values and those therebetween of this range). To facilitate lateral cross flow of cooling water between cells 8110 in the fuel rack 8100, a minimum of two lateral flow holes 8115A may be provided proximate to the lower or bottom end 8114 of each tube 8120 (see, e.g. FIGS. 77 and 78). Each hole defines top, bottom, and side edges in tube material. In one embodiment, the flow holes 8115A may be formed by a punching operation. Pedestals 8200 may have any suitable configuration or shape and be of any suitable type. Each fuel rack 8100 may include a plurality of peripheral pedestals 8200 spaced apart and arranged along the peripheral edges and perimeter of the baseplate 8102, and optionally one or more interior pedestals if required to provide supplemental support for the inboard fuel assemblies and tubes 8120B. In one non-limiting embodiment, four peripheral pedestals 8200 may be provided each of which is located proximate to one of the four corners 8206 of the baseplate. Additional peripheral pedestals may of course be provided as necessary between the corner pedestals on the perimeter of the baseplate. The pedestals are preferably located as outboard as possible proximate to the peripheral edges 8208 of the baseplates 8102 of each fuel rack or module to give maximum rotational stability to the modules. With continuing reference to FIGS. 74-78, each fuel rack storage tube 8120 in some embodiments may include a longitudinally-extending absorber sheath 8300 disposed on one or more tube sidewalls 8116. The sheath 8300 extends at least over the active zone or height of the fuel rack tubes 8120 where the fuel is positioned in the fuel rack 8100 (see, e.g. FIG. 78). Sheath 8300 has a raised profile or projection from the tube sidewall 8116. Sheath 8300 has a vertically elongated and generally flat body including top end 8310 defining a top lip or edge, bottom end 8311 defining a bottom lip or edge 8436, and a sidewall 8312 extending axially between the top and bottom ends. The top and bottom ends 8310, 8311 terminate at a point spaced apart from the top and bottom ends 8112, 8114 of the storage tube 8120 as shown. The sheath 8300 may be attached to the tube sidewall 8116 via welding or another suitable technique. Sheath sidewall 8312 is spaced laterally apart from the sidewall 8116 of the tube 8120 such that each “picture frame” sheath 8300 forms an envelope defining a sheathing cavity 8301 between the sheath and tube sidewall which is configured for receiving neutron absorber material 8302 therein (e.g. in sheet or panel form as represented in FIGS. 77 and 78). The sheath body is therefore configured and laterally offset from the tube sidewall 8116 by a distance commensurate with the dimensions and thickness of the absorber sheet or panel inserted therein. The boron-containing material or “poison” may be Boraflex, Tetrabor, (both previously mentioned) or another. In some existing used fuel rack installations, the absorber material 8302 may be in a degraded condition thereby requiring augmentation with a neutron absorber apparatus disclosed herein to restore fuel neutron reactivity control to the fuel rack. FIGS. 79-85 show a neutron absorber apparatus according to the present disclosure. The apparatus may be in the form of a shaped neutron absorber insert 8400 configured to be slidably insertable into one of the tubes 8120 and cells 8110 of the fuel rack 8110 shown in FIGS. 74-78 discussed above. Absorber insert 8400 includes a plurality of longitudinally-extending neutron absorber walls or plates 8402 each comprising a neutron absorber material operable to control reactivity of the fuel stored in the fuel rack cells. The absorber plates 8402 may be made of a suitable boron-containing metallic poison material such as without limitation borated aluminum. In some embodiments, without limitation, the absorber plates 8402 may be formed of a metal-matrix composite material, and preferably a discontinuously reinforced aluminum/boron carbide metal matrix composite material, and more preferably a boron impregnated aluminum. One such suitable material is sold under the tradename METAMIC™. Other suitable borated metallic materials however may be used. The boron carbide aluminum matrix composite material of which the absorption plates 8402 are constructed includes a sufficient amount of boron carbide so that the absorption sheets can effectively absorb neutron radiation emitted from a spent fuel assembly, and thereby shield adjacent spent fuel assemblies in a fuel rack from one another. The absorption plates may be constructed of an aluminum boron carbide metal matrix composite material that is about 20% to about 40% by volume boron carbide. Of course, other percentages may also be used. The exact percentage of neutron absorbing particulate reinforcement which is in the metal matrix composite material, in order to make an effective neutron absorber for an intended application, will depend on a number of factors, including the thickness (i.e., gauge) of the absorption plates 8402, the spacing between adjacent cells within the fuel rack, and the radiation levels of the spent fuel assemblies. In one configuration, absorber insert 8400 may comprise an assembly formed by two bent and chevron-shaped angled plates (designated 8402A and 8402B for convenience of reference), which are held together by metallic upper and lower stiffening bands 8404, 8406. Each plate 8402A, 8402B has the shape of a common structural angle sized to fit within the interior dimensions of each fuel rack storage tube 8120/cell 8110. Absorber plates 8402A, 8402B may each be formed of a generally flat or planar plate or sheet of neutron absorber material which is mechanically bent along a linear longitudinal bend line BL extending the plate's length L2 to form first and second half-sections 8408, 8410. The bend line BL may be located midway between the two side edges 8412 of the plates 8402A or 8402B so that each half-section 8408, 8410 has an equal width W2. In other possible embodiments, the half-sections may have unequal widths. Half-sections 8408 and 8410 may be arranged mutually perpendicular to each other at a 90-degree angle around the bend line BL in one embodiment as shown. When the absorber plates 8402A, 8402B are fastened together via the stiffening bands 8404, 8406, they collectively form a tubular box frame comprising a four-sided rectilinear absorber tube 8424 having a vertical centerline IC and defining an exterior surface 8418 and interior surface 8420. Interior surface 8420 in turn defines a longitudinally-extending and completely open central cavity 8422 configured for insertably receiving and holding a nuclear fuel assembly 8028 therein (typical fuel assembly shown in FIG. 78). Cavity 8422 extends from upper end 8414 to lower end 8416 of the absorber tube 8424. The ends 8414 and 8416 of the tube are open. Absorber tube 8424 and concomitantly cavity 8422 may have a square cross sectional shape in one embodiment as shown. Rectangular or other cross sectional tube and cavity shapes may be used in some embodiments depending on the cross sectional shape of the fuel storage tubes 8120. The mating longitudinal edges 8426 of the absorber tube plates 8402A and 8402B may laterally spaced apart in some embodiments forming an axially extending slot 8412 for the entire length of the absorber tube assembly (see, e.g. FIG. 79). The slot width is fixed by the upper and lower stiffening bands 8404, 8406 to which the absorber plates are fastened. In other embodiments, the longitudinal edges 8426 of the absorber plates 8402A, 8402B may be abutted without any appreciable gap. Upper and lower stiffening bands 8404, 8406 may be annular ring-like structures having a complementary configuration to the absorber tube 8424. Stiffening bands 8404, 8406 may have a square configuration in the non-limiting illustrated embodiment. The upper and lower bands are attached to the upper and lower extremities of the absorber tube plates 8402A, 8402B, respectively. Methods used to secure the bands 8404, 8406 to the upper and lower ends 8414, 8416 of the plates include for example without limitation welding, riveting, threaded fasteners, or other techniques. The stiffening bands may be made of a corrosion resistant metal, such as stainless steel in one embodiment. Referring to FIGS. 79-83, the upper stiffening band 8404 extends perimetrically around the upper end 8414 of the absorber tube 8424. The upper stiffening band 8404 is sized to closely fit inside the upper region of the fuel storage cell 8110/tube 8120 with a very small clearance between interior surfaces of the fuel rack storage tube sidewalls 8116 and the band, thereby giving the absorber tube 8424 structural rigidity and rotational fixity of position in the storage cell at the upper end of the absorber tube. In one embodiment, the upper stiffening band is preferably attached to the exterior surfaces 8418 of the absorber tube plates 8402A, 8402B at the upper end 8414 of absorber tube 8424. The upper stiffening band may be disposed precisely at the upper end 8414 of absorber tube 8424 as illustrated, or in other embodiments may be proximate to but spaced vertically downwards apart from the upper end 8414. In either case, upper stiffening band 8404 is preferably located at an elevation at least above the top end 8310 of the absorber sheath 8300 on storage tube 8120 to prevent interference with the sheath when inserting the absorber tube into the fuel storage cell 8110. Upper stiffening band 8404 projects laterally and transversely outwards from and beyond the exterior of the absorber tube 8424 to engage the sidewalls 8116 of the storage tube. When the absorber tube 8424 is installed in one of the fuel rack cells 8110 as shown in FIG. 78, the outwards projection of upper stiffening band 8404 laterally spaces the absorber tube 8424 apart from the interior cell side walls 8116. This creates a clearance gap G1 between the exterior surfaces 8418 of the absorber tube 8424 (formed by tube absorber plates 8402A, 8402B) and interior surfaces of the cells 8110 (formed by the sidewalls 8116 of the fuel storage tubes 8120). Gap G1 is preferably sized commensurate to the lateral projection depth D2 of the sheaths 8300 on the fuel storage tubes 8120 to receive the sheaths in the gap when installing the absorber tube 8424 in the fuel storage cell 8110. This allows the absorber tube 8424 to be slideably inserted into the fuel storage cell 8110 without interference from the projection of the sheaths 8300 outwards from the sidewalls 8116 of the storage tube 8120 (see, e.g. FIG. 78). Because the sheaths 8300 have a longitudinal length which terminates short of the upper and lower ends of the fuel storage tubes 8120 as shown in FIG. 77, the upper stiffening band 8404 may be fully seated inside the upper end of the storage tube without interference from the sheath (see, e.g. FIG. 82). To further avoid interference with the sheaths 8300 when the absorber tube 8424 is slid into the fuel storage tube 8120 through the open top end 8112 of the storage tube, the lower stiffening band 8406 is instead mounted in the interior or cavity 8422 of the absorber tube in one embodiment as best shown in FIG. 83. Lower stiffening band 8406 extends perimetrically around the lower end 8416 of the absorber tube 8424 in cavity 8422. The lower stiffening band provides structure rigidity and rotationally fixity in position to the lower end portion of the absorber tube 8424 when seated in the fuel storage cell 8110. Lower stiffening band 8406 may be completely recessed inside the absorber tube 8424 within central cavity 8422 wherein the lower end of the tube 8424 engages the baseplate 8102 of the fuel rack when the absorber insert is fully inserted therein. In alternative embodiments, the lower stiffening band may have an extended length and protrude downwards beyond the lower end 8416 of the absorber tube 8424 to engage the baseplate 8102. If the storage tube 8120 has optional lateral flow holes 8115A as shown in FIGS. 77 and 78, matching flow holes (not shown) may be provided at corresponding locations in the lower stiffening band 8406. When the absorber tube 8424 is fully seated in the storage tube 8120, the flow holes in absorber tube would become concentrically aligned with the lateral flow holes 8115A of the storage tube to preserve fuel pool cooling water cross flow between cells 8110. According to another aspect, the absorber tube 8424 may include one or more axial restraints to lock and axially fixate the tube in longitudinal position within the storage cell 8110 of the fuel rack 8100. Referring to FIGS. 79-84, the axial restraints in one non-limiting embodiment may be formed by elastically deformable locking protrusions comprised of metal leaf spring clips 8430. Spring clips 8430 each have an elongated body formed of corrosion resistant spring steel. Clips 8430 include a lower fixed end portion 8432 rigidly attached to the exterior surface 8418 of the absorber tube 8424 and an opposite resiliently movable cantilevered upper free-end locking portion 8434. Fixed end portion 8432 may be substantially flat and fixedly attached to absorber tube plates 8402A, 8402B by any suitable means, such as without limitation welding, riveting, or fasteners in some embodiments. Locking portion 8434 extends upwardly from fixed end portion 8432 and is obliquely angled thereto forming a space between the locking portion and the absorber tube 8424. Locking portion 8434 thus projects laterally outwards from the absorber tube 8424 (i.e. absorber plates 8402A, 8402B). When the absorber tube 8424 is installed in the fuel rack storage tube 8120, locking portion 8434 is also obliquely angled to the vertical longitudinal axis LA of the fuel rack (identified in FIG. 75). The locking spring clips 8430 are positioned on the lower half of absorber tube 8424 and arranged to engage an available edge disposed on the lower half of the fuel storage tubes 8120. In one embodiment, the spring clips may be positioned to engage a free bottom edge 8436 of the sheaths 8300 which is laterally spaced away from sidewall 8116 of the storage tube 8120, (see, e.g. FIGS. 77, 78, and 84). The free bottom edges 8436 are often formed near the lateral end portions 8438 of the bottom end 8430 of the sheath 8330 where the sheath is not welded or otherwise attached to the storage tube 8120. In such configurations, the spring clips 8430 may be disposed proximate to the corners 8428 of the lower half of the absorber tubes 8424 to engage the bottom edges 8436 of the sheaths 8300. Any suitable number of spring clips 8430 may be provided. In one embodiment, at least two spring clips 8430 may be provided preferably on different sides of the absorber tube 8424. In other embodiments, each of the four sides of the absorber tube may have at least one spring clip. Preferably, at least one spring clip 8430 is located to engage one available bottom edge 8436 of a sheath 8300 of the storage cell 8110 in which the absorber tube is installed to lock the absorber tube axially in place in the cell. It bears noting that at least one of the four storage tube sidewalls 8116 inside of each fuel storage cell 8110 includes a sheath 8300 for engagement by a locking spring clip 8430. This single engagement is sufficient to lock the absorber tube 8424 in position within the storage cell. The locking protrusion or spring clip 8430 is resiliently movable between an outward an inward deflected and retracted position for sliding the absorber tube 8424 into the fuel storage tube 8120 or cell 8110, and an outward undeflected and extended position for engaging the sheath 8300 and locking the absorber tube in position in the fuel rack 8100. Operation of the locking protrusion or spring clip 8430 will become evident by describing a method for installing a tubular neutron absorber insert 8400 in a storage cell 8110 of a fuel rack. A suitable cell 8110 may first be selected having at least one available absorber sheath 8300 for locking the insert in the fuel rack 8100. In one example, cell 8110A identified in FIG. 76 may be selected. The fuel rack 8100 may be still submerged in the fuel pool 40 and radioactively active. Preferably, a fuel assembly 8028 if already present in cell 8110A may be removed first. An absorber insert 8400 which may be in the form of absorber tube 8424 described above is then positioned over and axially aligned with cell 8110A. The locking spring clip or clips 8430 are initially in their outward undeflected and extended position (see, e.g. FIG. 84). An overhead hoist or crane may be used to deploy the absorber insert 8400. The insert 8400 is then slowly lowered into the cell 8110A through open top end 8112 of the cell. After the lower end 8416 of the absorber insert 8400 passes through the cell top end 8112, at least one of the locking spring clips 8430 slideably engages the top end 8310 of at least one absorber sheath 8300. The spring clip 8430 compresses and folds inward to the deflected and retracted position against the absorber tube 8424. As the absorber insert 8400 continues to be lowered farther into the cell 8110A, the locking portion 8434 of the spring clip 8430 slides along the sidewall 8312 of the sheath 8300 and remains in the compressed retracted position. When the spring clip 8430 eventually passes beneath and reaches a lower elevation in cell 8110A below the bottom end 8311 of the sheath, the spring clip 8430 will snap open via its elastic memory returning to the initial extended position of the spring clip thereby catching and lockingly engaging the bottom edge 8436 of sheath 8300 (see, e.g. FIGS. 78 and 11). This locking engagement between the sheath 8300 and locking portion 8434 of spring clip 8430 prevents the absorber insert 8400 from being axially withdrawn from the fuel rack cell 8110A, thereby locking the insert in axial position in the fuel rack. Advantageously, reactivity control to cell 8110A is fully restored despite the degraded original boron-containing neutron absorber material which may still be present in the sheath. The open cavity 8422 of the low profile absorber insert 8400 is configured to allow a fuel assembly 8028 to be inserted into cell 8110A following the absorber restoration process, and to be removed from the storage cell without requiring removal of the insert. It bears noting that while the upper stiffening band 8404 rotationally and laterally stabilizes the upper portion of the absorber insert 8400 in the storage tube 8120, the sheath 8300 on the tube sidewall and the spring clips 8430 act to rotationally and laterally stabilize lower portions of the insert by preventing excessive movement even during a seismic event. The absorber insert 8400 may also be used in some embodiments with a fuel storage tube 8120 that does not include an absorber sheath 8300 on at least one sidewall 8116 for engagement by the spring clip 8430, but instead includes an optional flow hole 8115A as shown in FIG. 77. In such a case, the spring clip 8430 may be configured and arranged on the absorber insert 8400 to engage a top edge of the flow hole 8115A for locking the insert axially in place in the tube. The insertion process and action of the spring clip 8430 is the same as described above, except that the surface of the storage tube sidewall 8116 engages the spring clip 8430 to fold the clip inwards in the retracted position until it passes below the flow hole 8115A. At that elevation, the clip springs or snaps back to the outward undeflected and extended position to lockingly engage the hole. FIG. 85 shows an alternative construction of an absorber insert 8400 according to the present disclosure. In lieu of the upper and lower stiffening bands 8404, 8406 coupling two chevron-shaped or angled absorber plates 8402A, 8402B together as shown in FIG. 79, each absorber plate 8402C, 8402D may be shaped as a structural channel. A longitudinal slot 8412 may be formed between mating edges 8426 of the plates 8402C and 8402D as shown in FIG. 85. All other element of construction including spring clips 8430 and stiffening bands 8404, 8406 may otherwise be the same as absorber plates 8402A, 8402B described herein. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims. |
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claims | 1. A process for the manufacture of a nuclear fuel cladding for a nuclear reactor, said fuel cladding comprising i) a substrate containing a zirconium-based internal layer coated or not coated with at least one interposed layer placed over said internal layer and ii) at least one external layer placed over the substrate and composed of a protective material chosen from chromium or a chromium-based alloy; the process comprising the following successive steps:a) providing a substrate containing a zirconium-based internal layer having an internal surface which delimits a closed volume for receiving the nuclear fuel and having an external surface opposite to said internal surface, said zirconium-based internal layer being coated or not coated with at least one interposed layer placed over said external face of the zirconium-based internal layer, said interposed layer having an internal face facing the external face of the zirconium-based internal layer, said interposed layer further comprising of at least one interposed material chosen from tantalum, molybdenum, tungsten, niobium, vanadium, hafnium or their alloys;b) ion etching of the external surface of the zirconium-based internal layer or, when present, of the external surface of the interposed layer; andc) deposition of said at least one external layer over the substrate and in contact with the external surface of the zirconium-based internal layer or, when present, with the external face of the interposed layer, said deposition being performed with a high power impulse magnetron sputtering (HiPIMS) process in which a magnetron cathode is composed of the protective material, said protective material being chosen from chromium or a chromium-based alloy,thereby producing a protected substrate suitable for use as a nuclear fuel cladding, wherein oxidation of the nuclear fuel cladding is inhibited on exposure of said cladding to a humid atmosphere at a temperature of 1200-1300° C. in the nuclear reactor. 2. A process for the manufacture of a nuclear fuel cladding according to claim 1, wherein the substrate provided in step (a) comprises an internal coating placed under said internal layer. 3. A process for the manufacture of a nuclear fuel cladding according to claim 1, wherein the zirconium-based internal layer is coated with at least one interposed layer placed over said internal layer. 4. A process for the manufacture of a nuclear fuel cladding according to claim 3, wherein said at least one interposed layer is placed over said internal layer by carrying out the following successive steps before the etching step a):a′) ion etching of the surface of said internal layer; andb′) production of a substrate by deposition of said at least one interposed layer over said internal layer with a high power impulse magnetron sputtering (HiPIMS) process in which the magnetron cathode is composed of the at least one interposed material. 5. A process for the manufacture of a nuclear fuel cladding according to claim 1, wherein deposition of said at least one external layer according to step (b) comprises depositing a first external layer with said HiPIMS sputtering process and depositing at least a part of an additional external layer or layers with a magnetron cathode sputtering process of a different type from the HiPIMS which is carried out simultaneously with said HiPIMS sputtering process. 6. The process according to claim 1, wherein each of said at least one external layer has a thickness of 1 μm to 50 μm. 7. The process according to claim 1, wherein each of said at least one external layer has a columnar structure. 8. The process according to claim 7, wherein constituent columnar grains of the columnar structure have a mean diameter of 100 nm to 10 μm. |
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abstract | A method for measuring an alignment of a detector is described. The method includes determining, by a processor, the alignment of the detector with respect to a collimated radiation beam. The determination of the alignment is based on a plurality of signals from a first cell of the detector and a second cell of the detector, and is independent of a shape of the collimated radiation beam. |
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claims | 1. A tool having an actuator for gripping a tubesheet of a heat exchanger having a plurality of heat exchange tubes extending at least partially through thru-holes in the tubesheet, each of the heat exchange tubes having a central axis extending along a length thereof, the actuator comprising:a first elongated finger sized to have a first end of the first elongated finger inserted at least partially within a first of the thru-holes within the tubesheet;a second elongated finger sized to have a first end of the second elongated finger inserted at least partially within a second of the thru-holes in the tubesheet, the second elongated finger being spaced from the first elongated finger to substantially align with the second of the thru-holes when the first elongated finger is substantially aligned with the first of the thru-holes;a tie rod connected between the first elongated finger and the second elongated finger at a first elevation along the first elongated finger and the second elongated finger that is spaced from the first ends, the connection of the tie rod between the first elongated finger and the second elongated finger being configured to restrain movement at the first elevation of the first elongated finger and the second elongated finger in at least a first of two lateral directions, either toward each other or away from each other; andan actuation arm connected between the first elongated finger and the second elongated finger at a second elevation along the first elongated finger and the second elongated finger that is spaced from the first elevation and spaced from the first ends, the connection of the actuation arm between the first elongated finger and the second elongated finger being configured to move the first elongated finger in at least one of the two lateral directions and cant at least one of the first elongated finger and the second elongated finger relative to the axis of a corresponding tube or thru-hole in which it is designed to be inserted to pressure the one of the first elongated finger and the second elongated finger against an inner wall of the corresponding tube or thru-hole and hold that position until the actuation arm is positively released. 2. The tool of claim 1 wherein the actuation arm cants both the first elongated finger and the second elongated finger relative to the axis of the corresponding tube or thru-hole in which it is designed to be inserted to pressure the first elongated finger and the second elongated finger against the corresponding tube in which it is inserted. 3. The tool of claim 1 wherein the actuation arm toggles between a locked position in which the at least one of the first elongated finger and the second elongated finger is canted relative to the axis of the corresponding tube or thru-hole in which it is designed to be inserted and an unlocked position in which the first elongated finger and the second elongated finger are not pressured against the inner wall of the corresponding tube or thru-hole. 4. The tool of claim 1 wherein both the first elongated finger and the second elongated finger are pressured against the inner wall of the corresponding tube or thru-hole when the actuation arm moves in the at least one of the two lateral directions. 5. The tool of claim 1 wherein the first elevation is between the first ends and the second elevation. 6. The tool of claim 5 wherein the tie rod restrains movement of the first elongated finger and the second elongated finger towards each other. 7. The tool of claim 1 wherein the tie rod restrains movement of the first elongated finger and the second elongated finger away from each other. 8. The tool of claim 1 wherein the second elevation is between the first ends and the first elevation. 9. The tool of claim 8 wherein the tie rod restrains movement of the first elongated finger and the second elongated finger towards each other. 10. The tool of claim 8 wherein the tie rod restrains movement of the first elongated finger and the second elongated finger away from each other. 11. The tool of claim 1 wherein the first elongated finger and the second elongated finger are configured to move a distance vertically independent of the actuation arm. 12. The tool of claim 1 wherein the actuation arm includes a compensator that is configured to accommodate a variation in spacing of the thru-holes while maintaining an approximately constant clamping force. 13. The tool of claim 12 wherein the compensator is an air spring. 14. The tool of claim 1 wherein a portion of the first end of either or both the first elongated finger and second elongated finger have a noncircular cross-section. 15. The tool of claim 1 including a resilient sheath over a portion of the first end of either or both the first elongated finger and second elongated finger. 16. The tool of claim 15 wherein the sheath is formed from an elastomeric material. 17. A method of supporting a tool from an underside of a heat exchange tubesheet having a plurality of openings extending through the underside comprising the steps of:inserting a portion of a first finger into a first opening in the underside of the tubesheet;inserting a portion of a second finger into a second opening in the underside of the tubesheet;leveraging the first finger off the second finger to clamp at least a part of the portion of either the first finger or the second finger that is inserted into the corresponding opening against a wall of the opening; andlocking the first finger and the second finger in their clamped position. 18. The method of claim 17 wherein the leveraging step clamps both the first finger and the second finger against the wall of the corresponding opening. 19. The method of claim 17 wherein the leveraging step cants either the first finger or the second finger relative to an axis of the corresponding opening in which it is inserted. 20. The method of claim 19 wherein the leveraging step cants both the first finger and the second finger relative to the axis of the corresponding opening in which it is inserted to clamp the first finger and the second finger against the wall of the corresponding opening. 21. The method of claim 17 including the step of suspending the tool from the first and second fingers. 22. The method of claim 17 including the step of moving the first finger and the second finger in a vertical direction independent of a mechanism for performing the leveraging step. 23. The method of claim 17 including the step of compensating for a variation in a distance between openings in the underside of the tubesheet while substantially maintaining a constant clamping force. |
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053496159 | claims | 1. A device for the retention of molten mass from a core melt-through in a light-weight reactor pressure vessel comprising: a water filled cooling basin beneath the pressure vessel; and a crucible disposed within the cooling basin, the crucible having a metal wall lined with a ceramic material and including a container with a plurality of cooling tubes extending from a lower wall of the container, a lid that forms a watertight upper seal for the container and a reinforcement structure disposed within the container for absorption of kinetic energy from the molten mass. a reactor pressure vessel having an outer wall; a containment foundation for housing the pressure vessel; a retention device disposed within the containment foundation for retaining molten mass from a core melt-through in the pressure vessel, the retention device comprising: a generally conical collecting structure disposed between the pressure vessel and the retention device, the collecting structure being adapted to protect the retention device from a section of the outer wall being blown off the pressure vessel and to direct the molten mass into the container. 2. A device according to claim 1 wherein the ceramic lining is made from a non-oxidizing ceramic, which has a thermal conductivity of more than 10 W/m K, at 800.degree. C., the ceramic lining having a wall thickness of at least 5 mm. 3. A device according to claim 2 wherein the ceramic lining is made from high-temperature isostatically pressed boron nitride. 4. A device according to claim 1, wherein the metal wall of the crucible is made from low-alloy steel. 5. A device according to claim 1 wherein the lower wall of the container is covered with a guard plate made of steel. 6. A device according to claim 1 wherein the cooling tubes have a cylindrical shape. 7. A device according to claim 1 wherein the lid is composed of a steel plate the reinforcement structure comprising a honeycomb structure of interlocking vertical plates in the container, the vertical plates being welded to the steel plate. 8. A device according to claim 1 wherein the crucible is at least 10 cm deep. 9. The device of claim 1 further including a generally conical collecting structure disposed between the pressure vessel and the retention device, the pressure vessel having an outer wall, the collecting structure being adapted to protect the retention device from a section of the outer wall being blown off the pressure vessel and to direct the molten mass into the container. 10. A core reactor plant comprising: 11. A plant according to claim 10 wherein the cooling basin of the retention device is formed from a cavity in the containment foundation. 12. A plant according to claim 10 further comprising a relief passage connected to the containment foundation for relieving pressure from the molten mass in the cooling basin. |
claims | 1. A method of assisting recovery of an injury site of an acute or chronic injury to a brain or spinal cord of a subject, the method comprising: irradiating the injury site with at least one array of microbeams comprising at least two parallel, spatially distinct microbeams in an amount and spatially arranged to deliver a therapeutic dose of X-ray radiation to said injury site, said therapeutic dose of X-ray radiation inhibiting the formation of a scar barrier and simultaneously promoting the regeneration of a microvascular and glial system at said injury site. 2. The method of claim 1, wherein said irradiating further comprises delivering the therapeutic dose with the at least one array of microbeams to the injury site repeatedly in a number n of sessions, each session being separated by a time interval. 3. The method of claim 2, wherein the at least one array comprises a number n of angle-variable intersecting microbeam arrays, the method further comprising generating the angle-variable intersecting microbeam arrays. 4. The method of claim 3, wherein said generating comprises the steps of:irradiating the injury site with one of the angle-variable intersecting microbeam arrays in one session;angularly displacing at least one of an X-ray radiation source generating the at least one array and the subject about an axis of rotation through a center of the injury site, wherein the axis of rotation is parallel to the at least two parallel, spatially distinct microbeams, to produce a second one of the angle-variable intersecting microbeam arrays;additionally irradiating the injury site with the second one of the angle-variable intersecting microbeam arrays after the time interval in a second session; and repeating said angularly displacing and additionally irradiating a number (n−1) times to generate the number n of angle-variable intersecting microbeam arrays, wherein the number n of angle-variable intersecting microbeam arrays intersect substantially only within the injury site, the injury site including a marginal volume surrounding injured tissue. 5. The method of claim 4, wherein adjacent angle-variable intersecting arrays are separated by a displacement angle, said angularly displacing comprising angularly displacing by a non-zero integer multiple of the displacement angle. 6. The method of claim 5, wherein the displacement angle is substantially equal to θ/(n−1), wherein θ is predetermined by an angular access of an X-ray source generating the angle-variable intersecting microbeam arrays to the injury site, θ being a total angular spread encompassing the angle-variable intersecting microbeam arrays. 7. The method of claim 6, wherein θ is substantially in a range of about 130 degrees to about 150 degrees. 8. The method of claim 3, wherein said generating comprises generating the angle-variable intersecting microbeam arrays for one of a horizontal, vertical, and slanted irradiation orientation of the at least two parallel, spatially distinct microbeams. 9. The method of claim 8, further comprising additionally generating a second number n of angle-variable intersecting microbeams arrays for another one of a horizontal, vertical and slanted irradiation orientation of the at least two parallel, spatially distinct microbeams, for a total number 2n of sessions, each session being separated by the time interval. 10. The method of claim 9, said additionally generating further comprising one of reorientating and replacing a multislit collimator between an X-ray source and the subject to change the irradiation orientation. 11. The method of claim 9, wherein the total number 2n of sessions is within a range of from three (3) to thirty (30) sessions. 12. The method of claim 2, wherein the time interval is substantially within a range of from about twelve (12) hours to about seven (7) days. 13. The method of claim 1, wherein the subject is positioned in one of an upright position, a side-reclined, and a slanted position, and wherein the at least one array is directed onto a subject's back, the at least one array being centered around a 90-degree angle of incidence. 14. The method of claim 1, wherein the at least one array comprises a center-to-center spacing between adjacent microbeams and a thickness of each of the at least two parallel, spatially distinct microbeams, wherein a ratio of the center-to-center spacing to the thickness is substantially in a range of about 4 to about 16. 15. The method of claim 1, wherein each of the at least two parallel, spatially distinct microbeams comprise a thickness substantially in a range of from about 0.02 mm to 1.0 mm. 16. The method of claim 1, wherein said irradiating further comprises generating said X-ray radiation with an X-ray bremsstrahlung source. 17. The method of claim 16, wherein each of the at least two parallel, spatially distinct microbeams comprise a thickness substantially in a range of from about 0.1 millimeters to 1.0 millimeter. 18. The method of claim 1, wherein said irradiating comprises generating X-ray synchrotron radiation, each of the at least two parallel, spatially distinct microbeams comprising a beam thickness substantially in a range of about 20 micrometers to about 100 micrometers. 19. The method of claim 1, wherein said irradiating further comprises generating X-ray radiation having a filtered broad beam energy spectrum, a half-power energy being substantially in a range from at least about 100 keV to about 250 keV. 20. The method of claim 1, wherein the therapeutic dose comprises an in-beam in-depth dose in each microbeam substantially in a range from about 30 Gy to about 500 Gy. 21. The method of claim 1, further comprising delivering stem cells to the injury site. 22. The method of claim 1, said irradiating further comprising delivering said at least one array in a plurality of temporally discrete pulses of said X-ray radiation. 23. The method of claim 22, wherein the plurality of temporally discrete pulses are substantially synchronized with a physiomechanical cycle of the subject. 24. The method of claim 23, wherein the physiomechanical cycle comprises at least one of a cardiac cycle and a cardiopulmonary cycle. |
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summary | ||
description | The present invention relates to shielded containers, and more particularly to a radiation-shielding container for storing and transporting a syringe holding a radiopharmaceutical. Radiation-shielding containers for storing, transporting, and dispensing radioactive drugs are known in the art. Radioactive drugs, commonly known as radiopharmaceuticals, are used to treat a variety of illnesses. However, technicians and medical personnel who handle these drugs on a regular basis must take precautions to reduce their exposure to the radiation emitted by radiopharmaceuticals. These precautions include, among other things, the use of radiation-shielding containers to store radiopharmaceuticals. Accordingly, radiation-shielding containers that are configured to hold vials of radiopharmaceutical liquid are known. Some containers provide access ports or other openings such that a liquid contained therein can be withdrawn from the vials using a syringe. Other containers exist that are configured to hold an individual syringe that contains radiopharmaceutical liquid. The syringe container is popular among hospitals and other care facilities because the radiopharmaceutical can be shipped and stored in pre-measured doses, thereby reducing the equipment and labor costs associated with storing and handling large quantities of radiopharmaceuticals. In one embodiment, the invention provides a radiation-shielding container for storing a syringe. The radiation-shielding container includes a body assembly for housing a portion of the syringe and a cap assembly for housing a portion of the syringe. The body assembly includes a body shield formed of radiation shielding material; a body shell that defines a first cavity having an open end for receiving the body shield and defines a first chamber portion for receiving a portion of the syringe, wherein an inner wall of the body shell separates the first cavity and the first chamber portion; and a body plug secured to the body shell to cover the open end of the first cavity and retain the body shield within the first cavity. The cap assembly includes a cap shield formed of radiation shielding material; a cap shell that defines a second cavity having an open end for receiving the cap shield and defines a second chamber portion for receiving a portion of the syringe, wherein an inner wall of the cap shell separates the second cavity and the second chamber portion; and a cap plug secured to the cap shell to cover the open end of the second cavity and retain the cap shield within the second cavity. The cap assembly is securable to the body assembly such that the first and second chamber portions define a chamber for storing the syringe. In another embodiment, the invention provides a radiation-shielding container for storing a syringe. The radiation-shielding container includes a body assembly for housing a portion of the syringe and a cap assembly for housing a portion of the syringe, wherein the cap assembly is securable to the body assembly. The body assembly includes a body shell having an outer wall, an inner wall spaced apart from the outer wall, an end wall connecting the outer wall to the inner wall, and an open end. The inner wall defines a first chamber portion for receiving the syringe, and the outer wall, the inner wall and the end wall define a first cavity. A body shield formed of radiation shielding material and having a substantially cylindrical shape is disposed within the first cavity of the body shell wherein an inner surface of the body shield lies adjacent the inner wall of the body shell. A body plug is received by the open end of the body shell to cover the open end and retain the body shield within the first cavity. The cap assembly includes a cap shell having an outer wall, an inner wall spaced apart from the outer wall, an end wall connecting the outer wall to the inner wall, and an open end. The inner wall defines a second chamber portion for receiving the syringe, and the outer wall, the inner wall and the end wall define a second cavity. A cap shield formed of radiation shielding material and having a substantially cylindrical shape is disposed within the second cavity of the cap shell wherein an inner surface of the cap shield lies adjacent the inner wall of the cap shell. A cap plug is received by the open end of the cap shell to cover the open end and retain the shield within the second cavity. In yet another embodiment, the invention provides a radiation-shielding container for storing a syringe. The radiation-shielding container includes a first assembly for housing a portion of the syringe. The first assembly having a shield formed of radiation shielding material, a shell that defines a first cavity having an open end for receiving the shield and defines a first chamber portion for receiving a portion of the syringe, wherein an inner wall of the shell separates the first cavity and the first chamber portion, and a plug secured to the shell to cover the open end of the first cavity and retain the shield within the first cavity. The radiation-shielding container also includes a second assembly for housing a portion of the syringe. The second assembly includes a shield formed of radiation shielding material and defining a second chamber portion. The first and second assemblies are securable together such that the first and second chamber portions define a chamber for storing the syringe. The radiation-shielding container includes a first assembly for housing a portion of the syringe and a second assembly for housing a portion of the syringe. The first assembly includes a shield formed of radiation shielding material and having a substantially cylindrical shape, and a protective coating that substantially surrounds and encases the shield. The first assembly defines a first chamber portion. The second assembly includes a shield formed of radiation shielding material and defining a second chamber portion. The first and second assemblies are securable together such that the first and second chamber portions define a chamber for storing the syringe. In another embodiment the invention provides a method of forming a radiation-shielding container for a syringe. The radiation-shielding container includes a first shell and a second shell, each shell having an outer wall, an inner wall spaced apart from the outer wall, an end wall connecting the outer wall to the inner wall, and an open end, wherein the inner wall defines a chamber portion, and the outer wall, the inner wall and the end wall define a cavity. The method includes placing a first shield in the cavity of the first shell, the first shield formed of radiation shielding material, wherein an inner surface of the first shield is positioned adjacent the inner wall of the first shell. The open end of the first shell is covered to retain the first shield within the cavity of the first shell. A second shield is placed in the cavity of the second shell, the second shield formed of radiation shielding material, wherein an inner surface of the second shield is positioned adjacent the inner wall of the second shell. The open end of the second shell is covered to retain the second shield within the cavity of the second shell. A syringe is placed within the chamber portion of the first shell and the second shell is secured to the first shell wherein the syringe is confined within the chamber portions of the first and second shells. Other aspects of the invention will become apparent by consideration of the detailed description and accompanying drawings. Before any embodiments of the invention are explained in detail, it is to be understood that the invention is not limited in its application to the details of construction and the arrangement of components set forth in the following description or illustrated in the following drawings. The invention is capable of other embodiments and of being practiced or of being carried out in various ways. Also, it is to be understood that the phraseology and terminology used herein is for the purpose of description and should not be regarded as limiting. The use of “including,” “comprising,” or “having” and variations thereof herein is meant to encompass the items listed thereafter and equivalents thereof as well as additional items. Unless specified or limited otherwise, the terms “mounted,” “connected,” “supported,” and “coupled” and variations thereof are used broadly and encompass both direct and indirect mountings, connections, supports, and couplings. Further, “connected” and “coupled” are not restricted to physical or mechanical connections or couplings. FIGS. 1 and 2 show a radiation-shielding container 10, often referred to as a radiopharmaceutical pig, for storing and transporting a syringe (not shown) containing a radiopharmaceutical. The container 10 includes radiation shielding to reduce exposure of radiation emitted by the radiopharmaceutical. The container 10 includes a body assembly 14 and a cap assembly 18 removably securable to the body assembly 14. When secured, the body assembly 14 and the cap assembly 18 define a chamber 22 for storing the syringe. The container 10 also includes a resilient O-ring 26 that provides a liquid-tight seal between the body assembly 14 and the cap assembly 18 when the two are secured to one another. In the illustrated embodiment, the body assembly 14 and the cap assembly 18 include a plurality of flats 30, such that each assembly 14, 18 preferably has a substantially octagonal shape, which provide grips to facilitate opening and closing of the container 10. Referring to FIGS. 1-3, the body assembly 14 includes a shell 34, a shield 38 formed of radiation shielding material that is received within the shell 34, and a plug 42 that retains the shield 38 within the shell 34. The shell 34 is generally cylindrical on the inside and includes an outer wall 46, an inner wall 50 spaced apart from the outer wall 46, a first end wall 54 extending between the outer wall 46 and the inner wall 50, and an open end 58 opposite the end wall 54. A second end wall 62 extends between ends of the inner wall 50 opposite the first end wall 54 and adjacent the open end 58. The outer wall 46, the inner wall 50, and the end walls 54, 62 define a cavity 66 for receiving the shield 38. The inner wall 50 and the second end wall 62 define a first chamber portion 70 for housing a portion of the syringe, wherein the inner wall 50 and the second end wall 62 separate the cavity 66 and the first chamber portion 70. In FIG. 3, an opening 74 of the first chamber portion 70 is located at a first end 78 of the body assembly 14 and the open end 58 of the shell 34 is located at a second end 82 of the body assembly 14. Proximate the first end 78 of the body assembly 14, a portion of the outer wall 46 of the shell 34 defines a cap-securing structure 86 that extends into an opening 90 of the cap assembly 18. The securing structure 86 defines a plurality of radially outwardly and circumferentially extending ribs 94. In the illustrated embodiment, four ribs are provided and are substantially equally spaced about the circumference of the first end. The cap-securing structure 86 defines a reduced diameter of the shell 34 relative to the remaining portion of the outer wall 46. Axially spaced from the ribs 94 and the first end wall 54 is an annular groove 98 that is adapted to receive the O-ring 26. The O-ring 26 engages the cap assembly 18 to substantially seal the chamber 22 when the cap assembly 18 is secured to the body assembly 14, as set forth hereafter. In the illustrated embodiment, the shell 34 is comprised of a suitable plastic, such as polycarbonate ABS blend, and is formed by injection molding. The base assembly 14 includes the shield 38 formed of radiation shielding material, such as lead, tungsten or the like. The shield 38 has a generally cylindrical shape and is received within the cavity 66 of the shell 34. An inner surface 102 of the shield 38 lies adjacent the inner wall 50 and the second end wall 62 of the shell 34, however, in the illustrated embodiment there is a clearance fit between the shield 38 and the shell 34. A gap 106 is formed between an outer surface 110 of the shield 38 and the outer wall 46 of the shell 34 such that the shield 38 does not fill the entire cavity 66, which reduces the amount of lead required for the shield 38 and thereby reduces the cost of the container 10. The shield 38 has a thickness sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. In the illustrated embodiment, the shield 38 has a thickness of about 0.20 inches, however in a further embodiment the thickness is between about 0.10 inches and about 0.25 inches. It should be readily apparent to those of skill in the art that the thickness may fall outside that range as long as the thickness is sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. The plug 42 covers the open end 58 of the shell 34 and retains the shield 38 within the shell 34. The plug 42 has a generally circular body portion 114 and includes an outer flange 118 extending axially inward from an outer circumference of the body portion 114, and an inner flange 122 spaced radially inward from the outer flange 118 and extending axially inward from the body portion 114. In use, the plug 42 is inserted into the open end 58 of the shell 34. In the illustrated embodiment, a recess 126 defined by the body portion 114 and the inner flange 122 engages a closed end 130 of the shield 30 and pushes opposite ends of the shield 38 into contact with the first end wall 54 of the shell 34, while maintaining the clearance fit between the inner surface 102 of the shield 38 and the inner wall 50 and the second end wall 54. The plug 42 retains the shield 38 within the cavity 66 and mechanically holds the shield 38 in place within the cavity 66. In the illustrated embodiment, the outer flange 118 is secured to the outer wall 46 of the shell 34, such as by friction welding, bonding, a fastener, or the like. Further, the plug 42 is comprised of a suitable plastic, such as polycarbonate ABS blend, and is formed by injection molding. Referring to FIGS. 1, 2 and 4, the cap assembly 18 includes a shell 134, a shield 138 formed of radiation shielding material that is received within the shell 134, and a plug 142 that retains the shield 138 within the shell 134. The shell 134 is generally cylindrical on the inside and includes an outer wall 146, an inner wall 150 spaced apart from the outer wall 146, a first end wall 154 extending between the outer wall 146 and the inner wall 150, and an open end 158 opposite the end wall 154. A first end of the inner wall 150 includes an angled portion 162 that extends to the first end wall 154. A second end wall 166 extends between second ends of the inner wall 150 opposite the first end wall 154 and adjacent the open end 158. The outer wall 146, the inner wall 150, and the end walls 154, 166 define a cavity 170 for receiving the shield 138. The inner wall 150 and the second end wall 166 define a second chamber portion 174 for housing a portion of the syringe, wherein the inner wall 150 and the second end wall 166 separate the cavity 170 and the second chamber portion 174. In FIG. 4, the opening 90 of the second chamber portion 174 is located at a first end 178 of the cap assembly 18 and the open end 158 of the shell 134 is located at a second end 182 of the cap assembly 18. In the illustrated embodiment, the shell is comprised of a suitable plastic, such as polycarbonate ABS blend, and is formed by injection molding. Proximate the first end 178 of the cap assembly 18, a flange 186 axially extends from the outer wall 146 and the first end wall 154 of the shell 134. The flange 186 defines a body-securing structure in the form of radially inwardly extending projections 190. The projections 190 cooperate with the ribs 94 of the cap-securing structure 86 to provide a releasable attachment between the cap assembly 18 and the body assembly 14. To couple the cap assembly 18 to the body assembly 14, the cap-securing structure 86 is inserted into the opening 90 of the second chamber portion 174 and the cap assembly 18, or the body assembly 14, is rotated approximately one-quarter turn to engage the ribs 94 with the projections 190. In the illustrated embodiment, the body assembly 14 and the cap assembly 18 are secured together by a bayonet-type interlocking means. However, it should be readily apparent to those of skill in the art that in further embodiments other securing structure may be used to secure the body assembly 14 and the cap assembly 18, such as threaded portions, or the securing structure may be reversed between the two assemblies such that the externally projecting ribs on defined by the cap assembly 18 and the internally extending projections are defined by the body assembly 14. The cap assembly 18 includes the shield 138 formed of radiation shielding material, such as lead, tungsten or the like. The shield 138 has a generally cylindrical shape and is received within the cavity 170 of the shell 134. A first end of the shield 138 defines a flange portion 194 that complements the contour of the inner wall 150 and the angled portion 162 of the shell 134. An inner surface 198 of the shield 138 lies adjacent the inner wall 150 and the second end wall 166 of the shell 134, however, in the illustrated embodiment there is a clearance fit between the shield 138 and the shell 134. A gap 202 is formed between an outer surface 206 of the shield 138 and the outer wall 146 of the shell 134 and the shield 138 does not fill the entire cavity 170, which reduces the amount of lead required for the shield 138 and thereby reduces the cost of the container 10. The shield 138 has a thickness sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. In the illustrated embodiment, the shield 138 has a thickness of about 0.15 inches, however in a further embodiment the thickness is between about 0.10 inches and about 0.25 inches. It should be readily apparent to those of skill in the art that the thickness may fall outside that range as long as the thickness is sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. The plug 142 covers the open end 158 of the shell 134 and retains the shield 138 within the shell 134. In the illustrated embodiment, the plug 142 of the cap assembly 18 is identical to the plug 42 of the body assembly 14 discussed above such that either plug 42, 142 could be used in the body assembly 14 or the cap assembly 18. The plug 142 has a generally circular body portion 210 and includes an outer flange 214 extending axially inward from an outer circumference of the body portion 210, and an inner flange 218 spaced radially inward from the outer flange 214 and extending axially inward from the body portion 210. In use, the plug 142 is inserted into the open end 158 of the shell 134. In the illustrated embodiment, a recess 222 defined by the body portion 210 and the inner flange 218 engages a closed end 226 of the shield 138 and pushes opposite ends of the shield 138 into contact with the first end wall 154 of the shell 134, while maintaining the clearance fit between the inner surface 198 of the shield 138 and the inner wall 150 and the second end wall 166. The plug 142 retains the shield 138 within the cavity 170 and mechanically holds the shield 138 in place within the cavity 170. In the illustrated embodiment, the outer flange 214 is secured to an inner surface of the outer wall 146 of the shell 134, such as by friction welding, bonding, a fastener, or the like. To fabricate the radiation-shielding container, the shells 34, 134 for the body and cap assemblies 14, 18, respectively, are formed by injection molding. The radiation shield 38 of the body assembly 14 is placed in the cavity 66 of the body assembly 14 such that the inner surface 102 of the shield 38 is positioned adjacent the inner wall 50 of the shell 34. The plug 42 is secured to the open end 58 of the body assembly shell 34 to cover the open end 58 and mechanically hold the shield 38 in place within the cavity 66. The radiation shield 138 of the cap assembly 18 is placed in the cavity 170 of the cap assembly 18 such that the inner surface 198 of the shield 138 is positioned adjacent the inner wall 150 of the shell 134. The plug 142 is secured to the open end 158 of the cap assembly shell 134 to cover the open end 158 and mechanically hold the shield 138 in place within the cavity 170. In the illustrated embodiment, preferably no adhesive is used to retain or hold the shields 38, 138 within the cavities 66, 170. A syringe (not shown) is placed in the first chamber portion 70 of the body assembly 14. The cap assembly 18 is secured to the body assembly 14 such that the syringe is confined within the chamber 22 defined by the first and second chamber portions 70, 174. FIG. 5 illustrates another embodiment of a radiation-shielding container. In FIG. 5, similar features of the radiation-shielding container 10 shown in FIGS. 1-4 are identified by similar reference numerals. A container 310 includes the body assembly 14 and the cap assembly 18 removably securable to the body assembly 14. When secured, the body assembly 14 and the cap assembly 18 define the chamber 22 for storing a syringe (not shown). In a further embodiment, the container 310 includes a resilient O-ring that provides a fluid-tight seal between the body assembly 14 and the cap assembly 18 when the two are secured together. The body assembly 14 includes the shell 34, the shield 38 formed a radiation shielding matter that is received within the shell 34, and a plug 314 that retains the shield 38 within the shell 34. The shell 34 and the radiation shield 38 are similar to the shell and the radiation shield shown and described above with respect to FIGS. 1-4. The plug 314 covers the open end 58 of the shell 34 and retains the shield 38 within the shell 34. The plug 314 has a generally cylindrical shape and is sized to securely fit within the open end 58 of the shell 34. An outer flange portion 318 of the plug 314 defines an outer circumference that is substantially equal to an inner circumference of the shell 34 at the open end 58. In the illustrated embodiment, a chamber 322 defined by the outer flange portion 328 engages the closed end 130 of the shield 38 and pushes opposite ends of the shield 38 into contact with the first end wall 54 of the shell 34, while maintaining a clearance fit between the inner surface 102 of the shield 38 and the inner wall 50 of the shell 34. The cap assembly 18 includes the shell 134, the shield 138 formed a radiation shielding matter that is received within the shell 134, and a plug 326 that retains the shield 138 within the shell 134. The shell 134 and the radiation shield 138 of the cap assembly 18 are similar to the shell and the radiation shield shown and described above with respect to FIGS. 1-4. The plug 326 covers the open end 158 of the shell 134 and retains the shield 138 within the shell 134. The plug 326 includes a generally circular body portion 330 and an outer flange 334 extending axially inward from an outer circumference of the body portion 330, such that the plug 326 has a generally cylindrical shape. In use, the plug 326 is inserted into the open end 158 of the shell 134. The outer flange portion 334 of the plug 326 defines an outer circumference that is substantially equal to an inner circumference of the shell 134 at the open end 158. In the illustrated embodiment, a recess 338 defined by the body portion 330 engages the closed end 226 of the shield 138 and pushes opposite ends of the shield 138 into contact with the first end wall 154 of the shell 134, while maintaining a clearance fit between the inner surface 198 of the shield 138 and the inner wall 150 of the shell 134. The plugs 314, 326 of the body assembly 14 and the cap assembly 18 retain the respective shields 38, 138 within the cavities 66, 170 and mechanically hold the shields 38, 138 in place within the cavities 66, 170. In one embodiment, the outer flange portion 318, 334 of each plug 314, 326 is secured to an inner surface of the outer wall 46, 146 of the respective shell 34, 134, such as by friction welding, bonding, a fastener or the like. Further, the plugs 314, 326 are comprised of a suitable plastic, such as polycarbonate ABS blend, and are formed by injection molding. FIG. 6 illustrates another embodiment of a radiation-shielding container. A container 410 includes the body assembly 14 and a cap assembly 414 removably securable to the body assembly 14. The body assembly 14 is identical to the body assembly shown and described above in FIG. 5, and similar features of the radiation-shielding container 310 shown in FIG. 5 are identified by similar reference numerals. When secured, the body assembly 14 and the cap assembly 414 define a chamber 418 for storing a syringe (not shown). The container 410 also includes the resilient O-ring 26 that provides a fluid-tight seal between the body assembly 14 and the cap assembly 414 when the two are secured together. The cap assembly 414 includes a shield 422 formed of radiation shielding material (such as lead, tungsten or the like), a protective coating 426, or shell, that surrounds and encases the shield 422, and a plug 430 that covers an exposed portion of the shield 422. The shield 422 has a generally cylindrical shape and a first end of the shield 422 defines a flange portion 434. The shield 422 has a thickness sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. The shield 422 is overmolded with a homogenous thermoplastic elastomer or thermoset, to form the protective coating 426. The protective coating 426 surrounds and encases the shield 422, except for an open portion located adjacent a second, closed end 438 of the shield 422, to substantially coat all exposed surfaces of the shield 422. The disk-shaped plug 430 is inserted into the open portion of the protective coating 426 to cover the remaining exposed portions of the shield 422. It should be readily apparent to those of skill in the art that other materials and methods may be used for fabricating the shield 422 and the protective coating 426. Proximate an open end 442 of the cap assembly 414, the protective coating 426 includes a flange 446 that axially extends away from the first end of the shield 422. The flange 446 defines a body-securing structure in the form of radially inwardly extending projections 450. The projections 450 cooperate with the ribs 94 of the cap-securing structure 86 to provide a releasable attachment between the cap assembly 414 and the body assembly 14. To couple the cap assembly 414 to the body assembly 14, the cap-securing structure 86 is inserted into the open end 442 of the cap assembly 414 and the cap assembly 414, or the body assembly 14, is rotated approximately one-quarter turn to engage the ribs 94 with the projections 450. It should be readily apparent to those of skill in the art that in further embodiments other securing structure may be used to secure the body assembly 14 and the cap assembly 414, such as threaded portions, or the securing structure may be reversed between the two assemblies such that the externally projecting ribs are defined by the cap assembly 414 and the internally extending projections are defined by the body assembly 14. FIG. 7 illustrates another embodiment of a radiation-shielding container. A container 510 includes a body assembly 514 and a cap assembly 518 removably securable to the body assembly 514. When secured, the body assembly 514 and the cap assembly 518 define a chamber 522 for storing a syringe (not shown). The container 510 also includes a resilient O-ring 526 that provides a fluid-tight seal between the body assembly 514 and the cap assembly 518 when the two are secured together. The body assembly 514 includes a shield 530 formed of radiation shielding material (such as lead, tungsten or the like) and a protective coating 534, or shell, that surrounds and encases the shield 530. The shield 530 has a generally cylindrical shape and has a thickness sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. In the illustrated embodiment, the thickness of the shield 530 is about 0.25 inches. The shield 530 is overmolded with a homogenous thermoplastic elastomer or thermoset, to form the protective coating 534. The protective coating 534 surrounds and encases the shield 530 to coat all exposed surfaces of the shield 530. Proximate an open end 538 of the body assembly 514, a portion of an outer surface 542 of the protective coating 534 defines a cap-securing structure 546 that extends into an open end 550 of the cap assembly 518. The securing structure 546 defines a plurality of radially outwardly and circumferentially extending ribs 554. The cap-securing structure 546 defines a reduced diameter of the outer surface 542 relative to the remaining portion of the outer surface. Axially spaced from the ribs 554 and the open end 538 of the body assembly 514 is an annular groove 558 that adapted to receive the O-ring 526. The O-ring 526 engages the cap assembly 518 to substantially seal the chamber 522 when the cap assembly 518 is secured to the body assembly 514. The cap assembly 518 includes a shield 562 formed of radiation shielding material (such as lead, tungsten or the like) and a protective coating 566, or shell, that surrounds and encases the shield 562. The shield 562 has a generally cylindrical shape and a first end of the shield 562 defines a flange portion 570. The shield 562 has a thickness sufficient to prevent or reduce exposure of radiation emitted by the radiopharmaceutical. In the illustrated embodiment, the thickness is about 0.15 inches. The shield 562 is overmolded with a homogenous thermoplastic elastomer or thermoset, to form the protective coating 566. The protective coating 566 surrounds and encases the shield 562 to coat all exposed surfaces of the shield 562. Proximate the open end 550 of the cap assembly 518, the protective coating 566 includes a flange 574 that axially extends away from the first end of the shield 562. The flange 574 defines a body-securing structure in the form of radially inwardly extending projections 578. The projections 578 cooperate with the ribs 554 to provide a releasable attachment between the cap assembly 518 and the body assembly 514. To couple the cap assembly 518 to the body assembly 514, the cap-securing structure 546 is inserted into the open end 550 of the cap assembly 518 and the cap assembly 518, or the body assembly 514, is rotated approximately one-quarter turn to engage the ribs 554 with the projections 578. It should be readily apparent to those of skill in the art that in further embodiments other securing structure may be used to secure the body assembly 514 and the cap assembly 518, such as threaded portions, or the securing structure may be reversed between the two assemblies such that the externally projecting ribs are defined by the cap assembly 518 and the internally extending projections are defined by the body assembly 514. In a further embodiment, the body assembly and the cap assembly may include plugs to cover an open portion of the protective coatings, as described above with respect to FIG. 6. FIGS. 8 and 9 illustrate another embodiment of a radiation-shielding container. A container 610 includes a body assembly 614 and a cap assembly 618 removably securable to the body assembly 614. When secured, the body assembly 614 and the cap assembly 618 define a chamber 622 for storing a syringe 626. The syringe 626 includes a generally cylindrical body 630, a plunger 634 that depends from one end of the body 630, and a needle (not shown) that extends from an opposite end of the body 630 and is protected by an optional, removable cap 638. The body 630 defines a radially extending flange 640 at the plunger end that facilitates movement of the plunger 634 with respect to the body 630. The container 610 includes a resilient O-ring 642 that provides a fluid-tight seal between the body assembly 614 and the cap assembly 618 when the two are secured together. In the illustrated embodiment, the body assembly 614 and the cap assembly 618 include a plurality of flats 646, such that each assembly 614, 618 preferably has a substantially rectangular shape, which provide grips to facilitate opening and closing of the container 610. Referring to FIG. 9, the body assembly 614 includes a shell 650, a shield 654 formed of radiation shielding material that is received within the shell 650, and a plug 658 that retains the shield 654 within the shell 650. The shell 650 is generally cylindrical on the inside and includes an outer wall 662, an inner wall 666 spaced apart from the outer wall 662, a stepped first end wall 670 extending between the outer wall 662 and the inner wall 666, and an open end 674 opposite the end wall 670. A second end wall 678 extends between ends of the inner wall 666 opposite the first end wall 670 and adjacent the open end 674. The outer wall 662, the inner wall 666, and the end walls 670, 678 define a cavity 682 for receiving the shield 654. The inner wall 666 and the second end wall 678 define a first chamber portion 686 for housing at least the body 630 and the needle of the syringe 626, wherein the inner wall 666 and the second end wall 678 separate the cavity 682 and the first chamber portion 686. In FIG. 3, an opening 690 of the first chamber portion 686 is located at one end of the body assembly 614 and the open end 674 of the shell 650 is located at an opposite end of the body assembly 614. In the illustrated embodiment, a removable liner 694, or bio-liner, is positioned within the first chamber portion 686 of the body assembly 614. The liner 694 has a generally cylindrical shape including an open end 696 and a closed end 700. The liner 694 extends substantially along the inner wall 666 of the shell 650, and is sized and configured to receive at least the body 630 and the needle of the syringe 626. The liner 694 prevents fluid contamination of the body assembly 614 if the protective cap 638 is removed from the syringe 626. In a further embodiment, the liner 694 may not be included in the body assembly 614, or the liner 694 may have other shapes. It should be readily apparent to those of skill in the art that other embodiments of a radiation-shielding container may include the bio-liner 694. Proximate the opening 690 of the first chamber portion 686, a portion of the outer wall 662 of the shell 650 defines a cap-securing structure that extends into an opening 704 of the cap assembly 618. In the illustrated embodiment, the securing structure is positioned rearward of the first end wall 670. The securing structure defines a plurality of radially outwardly and circumferentially extending ribs 708. In the illustrated embodiment, four ribs 708 are provided and are substantially equally spaced about the circumference of the opening 690. Axially spaced from the ribs 708 and the first end wall 670 is an annular groove 712 that is adapted to receive the O-ring 642. The O-ring 642 engages the cap assembly 618 to substantially seal the chamber 622 when the cap assembly 618 is secured to the body assembly 614, as set forth hereafter. In the illustrated embodiment, the shell 650 is comprised of a suitable plastic, such as polycarbonate ABS blend, and is formed by injection molding. The base assembly 614 includes the shield 654 formed of radiation shielding material, such as lead, tungsten, or the like. The shield 654 has a generally cylindrical shape and is received within the cavity 682 of the shell 650. An inner surface 716 of the shield 654 lies adjacent the inner wall 666 and the second end wall 678 of the shell 654, however, in the illustrated embodiment there is a clearance fit between the shield 654 and the shell 650. A gap 720 is formed between an outer surface 724 of the shield 654 and the outer wall 662 of the shell 650 such that the shield 654 does not fill the entire cavity 682. An outwardly extending radial flange 728 is axially spaced from an open end 732 of the radiation shield 54. The flange 728 and the open end 732 have a stepped configuration that complements the contour of and lies adjacent to the first end wall 670 of the body assembly 614. The plug 658 covers the open end 674 of the shell 650 and retains the shield 654 within the shell 650. The plug 658 has a generally circular body portion 736 and an outer flange 740 extending axially inward from an outer circumference of the body portion 736, such that the plug 654 has a generally cylindrical shape. An inner flange 744 is spaced radially inward from the outer flange 740 and extends axially inward from the body portion 736. In use, the plug 658 is inserted into the open end 674 of the shell 650. In the illustrated embodiment, the inner flange 744 and a recess 748, which is defined by the body portion 736 and the inner flange 744, engages a closed end of the shield 654 and pushes the open end 732 of the shield 654 into contact with the first end wall 670 of the shell 650, while maintaining the clearance fit between the inner surface 716 of the shield 654 and the inner wall 666 and the second end wall 678. The plug 658 retains the shield 654 within the cavity 682 and mechanically holds the shield 654 in place within the cavity 682. In the illustrated embodiment, the outer flange 740 is secured to the outer wall 662 of the shell 650, such as by friction welding, bonding, a fastener, or the like. Further, the plug 658 is comprised of a suitable plastic, such as polycarbonate ABS blend, and is formed by injection molding. Referring to FIG. 9, the cap assembly 618 includes a shell 752, a shield 756 formed of radiation shielding material that is received within the shell 752, and a plug 760 that retains the shield 756 within the shell 752. The shell 752 is generally cylindrical on the inside and includes an outer wall 764, an inner wall 768 spaced apart from the outer wall 764, a stepped first end wall 772 extending between the outer wall 764 and the inner wall 768, and an open end 776 opposite the end wall 772. A second end wall 780 extends between second ends of the inner wall 768 opposite the first end wall 772 and adjacent the open end 776. The outer wall 764, the inner wall 768, and the end walls 772, 780 define a cavity 784 for receiving the shield 756. The inner wall 768 and the second end wall 780 define a second chamber portion 788 for housing at least the plunger 634 portion of the syringe 626, wherein the inner wall and the second end wall 780 separate the cavity 784 and the second chamber portion 788. The opening 704 of the second chamber portion 788 is located at one end of the cap assembly 618 and the open end 776 of the shell 752 is located at an opposite end of the cap assembly 618. In the illustrated embodiment, the shell 752 is comprised of a suitable plastic, such as polycarbonate ABS blend, and is formed by injection molding. Proximate the first end of the cap assembly 618, an annular protrusion 792 axially extends from the outer wall 764 and the first end wall 772 of the shell 752. The annular protrusion 792 defines a body-securing structure in the form of radially inwardly extending projections 796. The projections cooperate with the ribs 708 of the cap-securing structure to provide a releasable attachment between the cap assembly 618 and the body assembly 614. To couple the cap assembly 618 to the body assembly 614, the cap-securing structure is inserted into the opening 704 of the second chamber portion 788 and the cap assembly 618, or the body assembly 614, is rotated approximately one-quarter turn to engage the ribs 708 with the projections 796. The flange portion 640 of the syringe 626 is caught between the first end wall 670 of the body assembly shell 650 and the first end wall 772 of the cap assembly shell 792, which holds the syringe 626 in place within the chamber 622. In the illustrated embodiment, the body assembly 614 and the cap assembly 618 are secured together by a bayonet-type interlocking means. However, it should be readily apparent to those of skill in the art that in further embodiments other securing structure may be used to secure the body assembly 614 and the cap assembly 618, such as threaded portions, or the securing structure may be reversed between the two assemblies 614, 618 such that the externally projecting ribs on defined by the cap assembly 618 and the internally extending projections are defined by the body assembly 614. The cap assembly 618 includes the shield 756 formed of radiation shielding material, such as lead, tungsten or the like. The shield 756 has a generally cylindrical shape and is received within the cavity 784 of the shell 752. A first end of the shield 756 defines an L-shaped flange portion 800 that complements the contour of and lies adjacent to the first end wall 772 of the shell 752. An inner surface 804 of the shield 756 lies adjacent the inner wall 768 and the second end wall 780 of the shell 752, however, in the illustrated embodiment there is a clearance fit between the shield 756 and the shell 752. A gap 808 is formed between an outer surface 812 of the shield 756 and the outer wall 764 of the shell 752 and the shield 756 does not fill the entire cavity 784. Referring to FIG. 9, the first end 800 of the shield 756 of the cap assembly 618 substantially surrounds the open end 732 of the shield 654 of the body assembly 614 such that the two shields overlap to prevent line of sight radiation leakage from the chamber 622. The plug 760 covers the open end 776 of the shell 752 and retains the shield 756 within the shell 752. In the illustrated embodiment, the plug 760 of the cap assembly 618 is identical to the plug 658 of the body assembly 614 discussed above such that either plug could be used in the body assembly 614 or the cap assembly 618. The plug 760 has a generally circular body portion 816 and an outer flange 820 extending axially inward from an outer circumference of the body portion 816, such that the plug 760 has a generally cylindrical shape. An inner flange 824 is spaced radially inward from the outer flange 820 and extends axially inward from the body portion 816. In use, the plug 760 is inserted into the open end 776 of the shell 752. In the illustrated embodiment, the inner flange 824 and a recess 828, which is defined by the body portion 816 and the inner flange 824, engages a closed end of the shield 756 and pushes opposite ends of the shield 756 into contact with the first end wall 772 of the shell 752, while maintaining the clearance fit between the inner surface 804 of the shield 756 and the inner wall 768 and the second end wall 780. The plug 760 retains the shield 756 within the cavity 784 and mechanically holds the shield 756 in place within the cavity. In the illustrated embodiment, the outer flange 820 is secured to an inner surface of the outer wall 764 of the shell 752, such as by friction welding, bonding, a fastener, or the like. It should be readily apparent to those of skill in the art that other suitable materials and methods may be used for fabricating the radiation shields, shells, plugs, and protective coatings described above. For example, the shells and the plugs may be formed from ABS, polycarbonate, polyethylene, PVC, CPVC, acrylic, fiberglass, polypropylene, carbon fiber, UHMW polyethylene, Nylon, polystyrene, any form of plastic or fiber filled/woven epoxy, or the like. Further, the shells and plugs may be fabricated by other known methods, such as, but not limited to, machining and two-part epoxy based applications. Various features and advantages of the invention are set forth in the following claims. |
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description | Not applicable. Not applicable. Not applicable. Not applicable. The present invention relates to a device for checking fuel pellets, used for making fuel rods for nuclear power plants, which detects if the position of the pellets with zirconium diboride is appropriate, and if any pellet is diverted. By diverted pellet it is meant that one which is located in an area not deemed for it to be. Either by having a coating in a place where it should be absent or vice versa. Likewise the present invention relates to the method applied. Zirconium diboride ZrB2 is a burnable neutron poison, used in making the fuel rods for pressurized water nuclear reactors. The use of zirconium diboride aims to: satisfy the needs of control of reactivity in new elements, the efficient use of fuel, an extension of the fuel life cycle, the power increase of the reactor and allowing higher burnup degrees. The fuel pellets with zirconium diboride, consist of fuel pellets of uranium oxide UO2 coated with a film or coating of some microns of ZrB2 (also called IFBA: Integral Fuel Burnable Absorber). The configuration of the fuel rods consists essentially of a zircaloy tube packed with fuel pellets (of about 8 or 10 mm in diameter and about 10 mm in length) and some structural components, varying depending of the product, as a spring, aluminium oxide pellets, supporting tubes, etc. The pellets with IFBA usually take up about 120 inches (about 305 cm) in length and are located in the middle of the rod. At both sides of the IFBA zone, typically there is a length of pellets of about 6 inches (about 15.2 cm) of UO2 having the same enrichment as the pellets with IFBA. In the fabrication process of these rods it is necessary to control the position of the IFBA zone, the length thereof and the detection of possible diverted pellets, (pellets without IFBA, pellets with a higher thickness of the IFBA layer, etc.). Currently the IFBA checking is carried in the factory through the combination of two inspections, one in active scanner and the other one in a passive scanner, determining the presence of IFBA by the calculation of the difference in gamma radiation emission obtained in both scanners. The invention consists of a device and method for checking fuel pellets according to claims. They have the advantage of allowing the determination of presence of IFBA through a single inspection in an active scanner, the inspection being automatable as one more parameter of those currently determined in the analysis of rods in the scanners for checking rods in fuel factories, whether active or passive. The conductivity degree of ZrB2 as compared to uranium oxide allows perceiving its presence through the variation experienced by an alternating magnetic field in the vicinity of the rod. The detection of this magnetic field variation and its analysis allows knowing the position of the pellets having a ZrB2 coating in the rod, by relating the position of the rod to each reading (at a known steady speed, time and position can be easily related). Therefore, the device for checking fuel rods (their zirconium diboride coating) comprises a variable magnetic field generator and a magnetic field pickup device, both arranged in the vicinity of the rod to be checked. Likewise, the control system can consist of a second variable magnetic field generator and a second magnetic field pickup device, both identical to the former ones and isolated form the rod to be controlled (for example, remote from the rod). So that the comparison of both magnetic fields allows detecting the variation of the electric conductivity in the rod, as a consequence of the presence or not of the conductive coating thereon. Preferably, the device further comprises a rod supplier which introduces them at a speed between 10 and 200 mm/s, more preferably 50 mm/s. In the preferred embodiment, both the generator and the pickup device shall be one or more coils. The checking method of fuel rods, in turn comprises the steps of: Arranging the rod to be measured between the generator and the pickup device. Generation of a variable magnetic field in the generator. Picking-up of the magnetic field. Comparison between the generated magnetic field and the picked-up one in order to quantify the electric conductivity of the rod. If the electric conductivity differs from the reference value, consider the rod for checking or recycling. This reference value can be calculated or found in calibration phases, previous or periodical. In the following an embodiment of the invention is briefly described, as an illustrative and non-limitative example thereof. The device of the invention comprises a variable magnetic field generator (1), which can be one or more coils, and a magnetic field pickup device (2), which generally will be another one or other coils. Alongside both devices will be located the fuel rod (3) with the pellets having the zirconium diboride coating intended to be measured. The pickup device (2) shall issue the received magnetic field to a control system (4), which shall make the comparison between the generated magnetic field and the picked-up one. Through subtraction of both signals it is possible to quantify the variation of the electric conductivity of the rod (3) due to the presence of pellets with a coating of ZrB2 in some zones of the rod (3) under measurement. Preferably, the control system (4) shall comprise a second generator (1′), identical to generator (1) and a second pickup device (2′), identical to pickup device (2), at an identical distance but remote from the rod (3) so as to not being affected by its effect. In this way, the effect of the rod (3) can be isolated, without having to refer to theoretical values of generated field. It is also possible to perform initially measurements in vacuum, without rod (3), in order to find the reference value of the generated field. The most effective solution is to arrange centred coils in the rod (3) as a generator (1) and as a pickup device (2) and connected through a Wheatstone bridge. The control system (4) would be identical. In order to perform the process continuously throughout the length of the rod (3), the device shall have a bar (3) supplier (5) thereof, with variable speed between 10 and 200 mm/s, generally 50 mm/s. The advancement speed depends on the configuration of the remaining checking or inspection operations in the scanner where the device is integrated. Considering the difference in electric conductivity between pellets with ZrB2 and without ZrB2 when the rod (3) passes, a differential signal shall be observed which allows identifying where begins and where ends the zone with pellets with IFBA (FIG. 2). If the value of conductivity is also analysed, if the coating is enough or if the rod (3) must be removed for a more deep examination, can be checked. The method for checking the fuel rod comprises therefore the steps of: Arranging the rod (3) to be measured between the generator (1) and the pickup device (2). Generation of a variable magnetic field in the generator (1). Picking-up of the magnetic field. Comparison between the generated magnetic field and the picked-up one in order to quantify the electric conductivity of the rod (3) If the electric conductivity differs from the expected reference value, consider the rod (3) for checking, recycling, or the corresponding procedure for non-compliant rods (3). Preferably all these steps shall be performed continuously, through the aforesaid bar (3) supplier (5). The reference value can be calculated or found out in a previous calibration phase, introducing rods (3) whose coating is known. This calibration can be repeated periodically in order to check the reference value. The magnetic field shall be typically created in a coil with low frequency (between 1 and 30 kHz, generally, although it can reach 100 kHz or more), with an excitation voltage of a few volts (2-15 V). However, both frequency and voltage shall depend on the characteristics of the generator (1) and of the pickup device (2). The frequency of the magnetic field shall depend also of external factors. For example, if the pellets are within a metal sheath (for example, circaloy), which is currently the standard, the sheath makes a screening when the frequency is very high (in the order of megahertz or higher). If the sheath is made of other material, such as composites or ceramics, this phenomenon does not occur and those frequencies can be used. In FIG. 2 can be seen an example of the application of the final reading for a dummy rod, composed of a circaloy tube, of 9.5 mm in outer diameter, 8.4 mm in inner diameter, packed with alumina pellets coated with ZrB2 (6) and alumina pellets (7) (non-conductive). The vertical axis is intensity, in decibels, whereas the horizontal one can be distance or time, because these area directly connected by the known steady speed. |
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description | This application claims the priority benefits of U.S. provisional application Ser. No. 62/148,755, filed on Apr. 17, 2015. The entirety of the above-mentioned patent application is hereby incorporated by reference herein and made a part of specification. Field of the Invention The invention relates to a target and an X-ray tube, and particularly relates to a composite target and an X-ray tube with the composite target. Description of Related Art An X-ray tube can be broadly divided into a transmission type X-ray tube and a reflection type X-ray tube, which is suitable for medical image, industrial testing, and other technical fields. FIG. 1 is a schematic diagram of a conventional transmission type X-ray tube. Referring to FIG. 1, a transmission type X-ray tube 100 includes a cathode 110, a focusing mechanism 120, an end window anode 130, a target 140, a power source supply 150, and a vacuum casing 160. An electron from the cathode 110 is accelerated along an electron beam path 170, so as to hit the target 140 to generate an X-ray 180. Referring to FIG. 1, the focusing mechanism 120 is used to focus on the electron, so as to control the position where the electron hits the target 140. The power source supply 150 is connected between the cathode 110 and the end window anode 130 to provide the energy for the electron, so as to accelerate the electron. The generated X-ray 180 penetrates through the end window anode 130 and emits to the outside of the vacuum casing 160. FIG. 2 is a schematic diagram of a conventional reflection type X-ray tube. Referring to FIG. 2, a reflection type X-ray tube 200 includes a cathode 210, an anode 220, a power source supply 230, a side window 240, and a vacuum casing 250. An electron from the cathode 210 is accelerated along an electron beam path 260, so as to hit the target (not shown) contained on the anode 220, thereby generating an X-ray 270 at the anode 220. Referring to FIG. 2, the power source supply 230 is connected between the cathode 210 and the anode 220 to provide the energy for the electron, so as to accelerate the electron. The generated X-ray 270 is reflected toward the side window 240 at the anode 220, and then emits to the outside of the vacuum casing 250. The transmission type X-ray tube 100 using the conventional single target as shown in FIG. 1 has the following issues which are illustrated by FIG. 3. FIG. 3 illustrates an output spectrum of a transmission type X-ray tube. In FIG. 3, a fixed tube voltage of 120 kV is used. It shows the relationship between the energy band of the X-ray photon on the horizontal axis and the amount of the X-ray photon on the vertical axis under the condition that tantalum with different thickness (13 μm, 50 μm, and 100 μm) is used as the target. Referring to FIG. 3, it can be learned that each target with different thickness (13 μm, 50 μm, and 100 μm) has different bremsstrahlung distribution. When it is desired to use different bremsstrahlung distribution, it is required to correspondingly replace the target with different thickness, so that the use is quite inconvenient. The amount of the X-ray photon of the conventional transmission type X-ray tube can be adjusted by using the methods such as adjusting the thickness, tube voltage, and tube current of the target. However, it is still difficult to obtain the required X-ray energy spectrum distribution. The reflection type X-ray tube 200 as shown in FIG. 2 has the following issues which are illustrated by FIG. 4. FIG. 4 is a diagram illustrating the comparison between an output spectrum of a transmission type X-ray tube and an output spectrum of a reflection type X-ray tube. The transmission type X-ray tube uses tantalum with a thickness of 25 μm as the target, and a filter layer is not provided; and the reflection type X-ray tube uses tungsten (W)+aluminum (Al) (1.6 mm) as the target. In FIG. 4, the horizontal axis represents the energy band of the X-ray photon, and the vertical axis represents the amount of the X-ray photon. Also, FIG. 4 shows the distribution of the X-ray photon of the two when the tube voltage is set at 120 kV. Referring to FIG. 4, at the same tube voltage (120 kV), the amount of the high energy photon of the X-ray of the reflection type X-ray tube is far more than the amount of the high energy photon of the X-ray of the transmission type X-ray tube. For the reflection type X-ray tube, the energy spectrum distribution ratio can be changed by increasing the tube voltage. Corresponding to the object with different thickness, the number of the X-ray photon which can penetrate the object to be tested can be increased by changing the tube voltage, so as to improve the image contrast ratio. However, as shown in FIG. 4, the amount of the low energy photon of the X-ray of the reflection type X-ray tube is usually too much, which causes unnecessary radiation absorbed dose for human body. Additionally, if the target is bombarded by the electron for a long time, the loss of the surface material of the target is generated. Also, since the target is hit by the electron, it becomes a high-temperature target, and the temperature which is close to the melting temperature of the target is often achieved. Thus, the target is subjected to high evaporation rate at the melting temperature, so as to shorten the life of the target. In view of this, the invention provides a composite target, which can generate a variety of X-ray energy spectrum distributions and with a sufficient long service life. The invention provides an X-ray tube with the above composite target, which can provide a variety of X-ray energy spectrum distributions and with a sufficient long service life. Based on the above, the invention provides a composite target interacted with an electron to generate an X-ray, and an energy of the electron can be changed by controlling a tube voltage at least. The composite target includes a target body and an interposing layer. The interposing layer is connected with the target, wherein the interposing layer moves a highest peak of an energy spectrum of the X-ray toward a high energy direction, and a low energy photon of the X-ray is filtered by the interposing layer. Also, the low energy photon of the X-ray can be increased by increasing a thickness of the interposing layer. As the tube voltage is enhanced, an amount of a high energy photon of the X-ray generated is increased. The invention also provides an X-ray tube including a casing, an anode, a cathode, and a power source. The anode is disposed at the casing, and a composite target is disposed on the anode. The composite target is interacted with an electron to generate an X-ray, and an energy of the electron can be changed by controlling a tube voltage at least. The composite target includes a target body and an interposing layer. The interposing layer is connected with the target body, wherein the interposing layer moves a highest peak of an energy spectrum of the X-ray toward a high energy direction, and a low energy photon of the X-ray is filtered by the interposing layer. Also, the low energy photon of the X-ray can be increased by increasing a thickness of the interposing layer. As the tube voltage is enhanced, an amount of a high energy photon of the X-ray generated is increased. The cathode is disposed in the casing to provide the electron. The power source is connected between the cathode and the anode. Based on the above, the composite target of the invention has the interposing layer. By the interaction between the electron and the target body and between the electron and the interposing layer, and the interaction generated from the X-ray and the interposing layer, the X-ray energy spectrum distribution can be changed. Also, by the protective layer which can protect the composite target from excessive bombardment of the electron, the composite target has a sufficient long life. Additionally, by the setting method of a plurality of film layers of the target body and by changing the position where the electron enters the target, the X-ray with a designated energy spectrum distribution can be chosen. In order to make the aforementioned features and advantages of the disclosure more comprehensible, embodiments accompanied with figures are described in detail below. The composite target and the X-ray tube with the composite target provided by the invention can provide a variety of X-ray energy spectrum distributions to meet various actual needs. Embodiments of the invention are illustrated referring to the following figures. FIG. 5 is a schematic diagram of an X-ray tube with a composite target of a preferred embodiment of the invention. Here, the composite target is used in the transmission type X-ray tube as an example to illustrate. However, the composite target is not limited to be used in the transmission type X-ray tube, it can be used in the reflection type X-ray tube, or any other type of X-ray tube. Referring to FIG. 5, an X-ray tube includes a casing 310, an anode 320, a cathode 330, and a power source 340. The anode 320 is disposed at the casing 310. A composite target 350 is disposed on the anode 320. The composite target 350 is interacted with an electron 360 to generate an X-ray 370, and an energy of the electron 360 can be changed by controlling a tube voltage at least. The composite target 350 includes a target body 352 and an interposing layer 354. The interposing layer 354 is connected with the target body 352. The interposing layer 354 may move a highest peak of an energy spectrum of the X-ray 370 toward a high energy direction. A low energy photon of the X-ray 370 is filtered by the interposing layer 354, and the low energy photon of the X-ray can be increased by increasing a thickness of the interposing layer 354. Also, as the tube voltage is enhanced, an amount of a high energy photon of the X-ray 370 generated is increased. The cathode 330 is disposed in the casing 310 to provide the electron 360. The power source 340 is connected between the cathode 330 and the anode 320 to provide such as the tube voltage and tube current, so as to adjust the energy of the electron 360. Referring to FIG. 5, the X-ray tube 300 may further include an electron track moving device 380 to control the position where the electron 360 enters the composite target 350. FIG. 6A to FIG. 6B are schematic diagrams of two types of electron track moving devices of embodiments of the invention. Referring to FIG. 6A, in an embodiment, an electron track moving device 380A may include a main body 380A1, and at least four electromagnets 380A2 correspondingly disposed on the main body 380A1. Specifically, as shown in FIG. 5, the electron track moving device 380A of FIG. 6A may be disposed inside the X-ray tube 300. In an embodiment not shown, the electron track moving device 380A may also be disposed outside the X-ray tube 300. By adjusting the size and direction of magnetic force of the electromagnet 380A2, the electron 360 can be moved to any position on the composite target 350. Referring to FIG. 6B, in another embodiment, the electron track moving device 380 may also be a magnet 380B. By rotating the magnet 380B (the arrow A as shown in FIG. 6B), the magnet performs a uniaxial movement (the arrow B as shown in FIG. 6B). Thereby, the electron 360 performs a uniaxial movement, so as to adjust the position where the electron 360 enters the composite target 350. As shown in FIG. 6B, the magnet 380B may be disposed outside the X-ray tube 300 to control the traveling track of the electron 360. For the material of the target body 352, in an embodiment, the material of the target body 352 may be selected from the group consisting of scandium, titanium, vanadium, chromium, manganese, iron, cobalt, nickel, copper, zinc, germanium, yttrium, niobium, molybdenum, ruthenium, rhodium, palladium, silver, tin, barium, lanthanum, cerium, neodymium, gadolinium, terbium, dysprosium, holmium, erbium, thulium, ytterbium, lutetium, hafnium, tantalum, tungsten, rhenium, iridium, platinum, gold, thorium, uranium, and a combination thereof. Additionally, the material of the interposing layer 354 may be selected from the group consisting of copper, silver, gold, indium, nickel, tin, aluminum, diamond, bismuth, antimony, tungsten, molybdenum, tantalum, zinc, cobalt, and a combination thereof. Hereinafter, the effect of the X-ray energy spectrum distribution adjusted by the composite target is illustrated by FIG. 7 and FIG. 8. FIG. 7 illustrates an output spectrum of a transmission type X-ray tube of an embodiment of the invention. The composite target 350, having a thickness of 25 μm, without providing a filter layer, the target body 352 using tantalum, and the interposing layer 354 using copper is used. In FIG. 7, the horizontal axis represents the energy hand of the X-ray photon, and the vertical axis represents the amount of the X-ray photon. FIG. 7 also illustrates the distribution of the X-ray photon when the tube voltage is different (80 kV, 100 kV, and 120 kV). Referring to FIG. 7, as the tube voltage is enhanced (80 kV→100 kV→120 kV, as shown in FIG. 7), the amount of the high energy photon of the X-ray (equal to or more than 60 KeV, the circle C part as shown in FIG. 7) of the X-ray tube 300 using the composite target 350 is significantly increased. From the comparison between the conventional FIG. 3A (without the interposing layer) and FIG. 7 of the invention (with the interposing layer 354, copper), it can be seen that, the interposing layer 354 changes the X-ray energy spectrum distribution as shown in FIG. 7. That is, the amount of the overall photon of the X-ray 370 (including the low energy photon of the X-ray and the high energy photon of the X-ray) occupied by the high energy photon of the X-ray is significantly increased, especially helpful to enhance the penetration for the object. FIG. 8 illustrates an output spectrum of a transmission type X-ray tube of another embodiment of the invention. Another composite target 350, having a thickness of 25 μm, without providing a filter layer, the target body 352 using tantalum, and the interposing layer 354 using the composite material of copper and silver is used. In FIG. 8, the horizontal axis represents the energy band of the X-ray photon, and the vertical axis represents the amount of the X-ray photon. FIG. 8 also illustrates the distribution of the X-ray photon when the tube voltage is different (80 kV, 100 kV, and 120 kV). Similarly, from the comparison between the conventional FIG. 3B (without the interposing layer) and FIG. 8 of the invention (with the interposing layer 354, copper and silver), it can be seen that, the amount of the overall photon of the X-ray 370 (including the low energy photon of the X-ray and the high energy photon of the X-ray) occupied by the high energy photon of the X-ray (equal to or more than 60 KeV, the circle D part as shown in FIG. 8) is significantly increased, especially helpful to enhance the penetration for the object. According to the above, the X-ray energy spectrum distribution of the transmission type X-ray tube 300 using the composite target (with the interposing layer 354) can be adjusted to dramatically increase the high energy photon of the X-ray. Therefore, the issue of the conventional transmission type X-ray tube 100 using the single target can be avoided. FIG. 9A to FIG. 9E are diagrams illustrating the comparison between an output spectrum of a transmission type X-ray tube of an embodiment of the invention and an output spectrum of a conventional reflection type X-ray tube. The transmission type X-ray tube 300 of an embodiment of the invention uses the composite target 350, having a thickness of 25 μm, the target body 352 using tantalum, the interposing layer 354 using copper, and without providing a filter layer; and the conventional reflection type X-ray tube 200 uses tungsten (W) as the target, and with 1.6 mm of aluminum as a filter material. FIG. 9A to FIG. 9E respectively represents an output spectrum when the tube voltage is 80 kV, 90 kV, 100 kV, 110 kV, and 120 kV. It can be learned from FIG. 9A to FIG. 9E that, since the interposing layer 354 is adopted by the transmission type X-ray tube 300 using the composite target 350 of the invention, the high energy photon of the X-ray can be dramatically increased, thereby obtaining the X-ray energy spectrum distribution which is similar to that of the conventional reflection type X-ray tube 200. Particularly, compared to the conventional reflection type X-ray tube 200, the interposing layer 354 of the invention may further filter the low energy photon of the X-ray which is harmful to the human body. FIG. 10A to FIG. 10E are diagrams illustrating the comparison of an output spectrum of a transmission type X-ray tube having different interposing layers. The composite target used by one transmission type X-ray tube has a thickness of 25 μm, the target body 352 using tantalum, the interposing layer 354 using copper, and without providing a filter layer. The composite target used by another one transmission type X-ray tube has a thickness of 25 μm, the target body 352 using tantalum, the interposing layer 354 using the composite material of copper and silver. FIG. 10A to FIG. 10E respectively represents an output spectrum when the tube voltage is 80 kV, 90 kV, 100 kV, 110 kV, and 120 kV. Referring to FIG. 10A to FIG. 10E, it can be seen that, the amount of the X-ray photon generated from the interposing layer 354 using the composite material of copper and silver is more than the amount of the X-ray photon generated from the interposing layer 354 using copper. Also, as the tube voltage is enhanced, the above trend is more obvious. It can be seen that, by changing the material composition of the interposing layer 354, the production efficiency of the X-ray photon can be adjusted. FIG. 11 to FIG. 15 are schematic cross-sectional diagrams of composite targets of several embodiments of the invention. The X-ray with a designated energy spectrum distribution chosen by the setting method of the film layer of the composite target is illustrated by FIG. 11 to FIG. 15. Referring to FIG. 11, the target body 352 at least includes a first film layer 352a, and a second film layer 352b (a number n of film layers are shown in FIG. 11). The second film layer 352b is disposed at one side of the first film layer 352a. The electron e can be interacted with the first film layer 352a and the second film layer 352b, so as to choose the X-ray with a designated energy spectrum distribution. Specifically, as shown in FIG. 11, the electron e can pass the first film layer 352a and the second film layer 352b to perform the mechanism for generating the X-ray, so as to generate the X-ray with a designated energy spectrum distribution. Actually, the target body 352 may include a number n of film layers, such as the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n. In an embodiment of FIG. 11, the electron e can pass the film layer composed of any combination of the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n, so as to obtain the required X-ray energy spectrum distribution. Additionally, when setting the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n, if the electron penetration depth is less than the thickness of the first film layer 352a, the first film layer 352a is a main film layer (as the target body), and other film layers 352b, . . . , 352n are as the interposing layers. If the electron penetration depth is less than the thickness of the first film layer 352a plus the second film layer 352b, both the first film layer 352a and the second film layer 352b are the main film layers (as the target body), and other film layers . . . , 352n are as the interposing layers. The following several basic settings are usually performed. The first setting: more high energy photon part is left. The atomic number of the film layer material of the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n is that, the first film layer 352a> the second film layer 352b> . . . the nth film layer 352n in order, and the thickness of the main film layer (i.e. target body) is more than the total thickness of other film layers (i.e. interposing layer). The second setting: the low energy photon and the high energy photon are filtered. The thickness of the main film layer (i.e. target body) is less than the total thickness of other film layers (i.e. interposing layer), and the atomic number of the nth film layer 352n is more than the atomic number of the main film layer. In industrial applications, the preferred setting of the thickness of the main film layer (i.e. target body) is 1/7 to ⅓ times a maximum electron penetration depth of the material. In medical diagnosis, the preferred setting of the thickness of the main film layer (i.e. target body) is 3 to 10 times a maximum electron penetration depth of the material, which is the preferred thickness setting. In medical treatment, the thickness of the main film layer (i.e. target body) for enhanced generation of characteristic radiation energy spectrum is 10 to 30 times a maximum electron penetration depth of the material, which is the preferred thickness setting. The maximum depth D of the high energy electron penetrating the target changes with different target, and formula is as follows: D ( µm ) = 2.76 × 10 - 2 AE 1.67 ρ Z 0.89 wherein, ρ=target density; Z=atomic number; A=atomic mass; E=incident electron voltage. Referring to FIG. 12, the target body 352 at least includes a first film layer 352a, and a second film layer 352b. The second film layer 352b is disposed at one side of the first film layer 352a. It can be noted that, the second film layer 352b and the first film layer 352a are staggered stacked, and electrons e1 and e2 can be interacted with the first film layer 352a, and a stacking location of the first film layer 352a and the second film layer 352b respectively, so as to choose the X-ray with a designated energy spectrum distribution. Specifically, as shown in FIG. 12, there is the first film layer 352a at part of the position, and there is a stacked structure of the first film layer 352a and the second film layer 352b at another part of the position. When the electron e1 enters the first film layer 352a, the X-ray with the first setting energy spectrum distribution can be generated; and when the electron e2 enters the stacked structure of the first film layer 352a and the second film layer 352b, the X-ray with the second setting energy spectrum distribution can be generated. It can be learned that, by changing the stacking ways of the film layer of the target body 352 and the position where the electrons e1 and e2 enter the target body 352, the X-ray with a variety of designated energy spectrum distributions can be provided and chosen according to the required conditions. Actually, as shown in FIG. 12, the target body 352 may include a number n of film layers, such as the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n. That is, the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n can be staggered stacked to obtain various film layer stacked structures. When the electron e enters different film layer stacked structure, the X-ray with a variety of designated energy spectrum distributions can be provided. Referring to FIG. 13, the target body 352 at least includes the first film layer 352a, and the second film layer 352b. The first film layer 352a and the second film layer 352b have a tilt interface S therebetween. The position of the electron e relative to the tilt interface S can be adjusted, such that the electron e can be interacted with the first film layer 352a and the second film layer 352b, so as to choose the X-ray with a designated energy spectrum distribution. As shown in FIG. 13, the position where the electron c enters the target body 352 can be moved from the left side to the right side of the target body 352. When the electron e only enters the first film layer 352a disposed at the left side, the X-ray completely generated from the first film layer 352a can be obtained; and when the electron e only enters the second film layer 352b disposed at the right side, the X-ray completely generated from the second film layer 352b can be obtained. Particularly, when the electron e enters the position of the tilt interface S, the X-ray co-generated from the first film layer 352a and the second film layer 352b can be obtained. Additionally, by moving the position of the electron e relative to the tilt interface S, the X-ray energy spectrum distribution generated from the first film layer 352a and the second film layer 352b composed in different proportions can be obtained. Referring to FIG. 14, the target body 352 at least includes the first film layer 352a, and the second film layer 352b. The second film layer 352b is disposed at one side of the first film layer 352a. The first film layer 352a and the second film layer 352b are stepped stacked. The electrons e1 and e2 can be interacted with the stacking location of the first film layer 352a and the second film layer 352b, and the second film layer 352b respectively, so as to choose the X-ray with a designated energy spectrum distribution. Specifically, as shown in FIG. 14, the thickness of the target body 352 is different from each other from the left side to the right side of the target body 352. When two film layers are adopted, that is, the first film layer 352a and the second film layer 352b are stepped stacked, the electron e1 enters the stacking location of the first film layer 352a and the second film layer 352b of the target body 352, and the electron e2 enters the second film layer 352b only. In other words, by adjusting the position where the electrons e1 and e2 enter the target body 352, the X-ray with different energy spectrum distribution can be generated from the target body 352 with different thickness respectively. Actually, as shown in FIG. 14, the target body 352 may include a number n of film layers, such as the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n. That is, by the stepped stacked method to stack the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n, the film layer stacked structure with various thicknesses can be obtained. When the electron e enters the film layer stacked structure with different thickness, the X-ray with a variety of designated energy spectrum distributions can be provided. Referring to FIG. 15, the target body 352 at least includes the first film layer 352a, and the second film layer 352b. The second film layer 352b is disposed at one side of the first film layer 352a. In the first film layer 352a and the second film layer 352b, a groove G with a designated shape is formed. The electron e can penetrate the groove G, and the electron e can be interacted with the stacking location of the first film layer 352a and the second film layer 352b, so as to choose the X-ray with a designated energy spectrum distribution. Specifically, in the target body 352 as shown in FIG. 15, the groove G may be a conical groove which causes the target body 352 having different thickness and different stacked structure at different position. When the electron e enters the target body 352 at different thickness or different stacked structure, the X-ray with a variety of designated energy spectrum distributions can be provided. Actually, as shown in FIG. 15, the target body 352 may include a number n of film layers, such as the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n. After the first film layer 352a, the second film layer 352b, . . . , and the nth film layer 352n are stacked, an etching process may be performed to form the groove G. When the electron e enters the target body 352 at different position, the X-ray with a designated energy spectrum distribution corresponding to the film layer structure of the position can be provided. By using the thickness gradient distribution, or stepped type distribution of the target body 352 of FIG. 11 to FIG. 15 coordinated with controlling the position where the electron e enters the target body 352, the X-ray with different energy spectrum distribution can be generated at the same tube voltage, so as to be conducive to adjust the image contrast ratio and adjust to the best image. In other words, in a single X-ray tube 300 using the above composite target 350, the X-ray with various energy spectrum distributions can be generated by the single X-ray tube 300. Therefore, the single X-ray tube 300 can be used in a variety of technical fields. Also, the trouble of replacing the X-ray tube according to different conditions can be avoided in the prior art. Referring to FIG. 5, and FIG. 11 to FIG. 15 at the same time, in the composite target 350, a protective layer 400 may also be further provided. The protective layer 400 is disposed at an upstream side of the composite target 350, and the protective layer 400 faces the electron 370. The critical energy of electron sputtering of the protective layer 400 is more than the critical energy of electron sputtering of the target body 352. Although FIG. 5 shows that the protective layer 400 and the composite target 350 are separately disposed, the protective layer 400 may be integrally disposed with the composite target 350. For example, the protective layer 400 is formed on a surface of the composite target 350. By the protective layer 400, the situation of material loss of the composite target 350 caused by electron bombardment may be reduced, and the situations of volatilization generated from high temperature of the composite target 350 or sublimation phenomenon generated from the vacuum being too high in the tube can be avoided. FIG. 16A to FIG. 16C are schematic top views of composite targets of several embodiments of the invention. Referring to FIG. 16A first, the target body 352 is divided into at least a first region R1 and a second region R2, and the first region R1 and the second region R2 have an interface therebetween. The electron e can be interacted with the first region R1 and the second region R2 respectively, so as to choose the X-ray with a designated energy spectrum distribution. Actually, the target body 352 in FIG. 16A can be divided into the first region R1 to the third region R3, or can be divided into an infinite number of parts. Additionally, the position where the electron e hits the first region R1 to the third region R3 or the interface S of the target body 352 can be controlled, so as to choose the X-ray with a designated energy spectrum distribution. FIG. 16B is another method to divide the target body 352. It can also be divided into the first region R1 to the third region R3. FIG. 16C is further another method to divide the target body 352 (circular distribution). It can also be divided into the first region R1 to the third region R3. Additionally, the electron e can be controlled, such that the electron e has a designated spot size. By adjusting the spot size of the electron e, the area range of the target body 352 of FIG. 11 to FIG. 15 and FIG. 16A to FIG. 16C hit by the electron e can be controlled, so that the X-ray with different energy spectrum distribution can be generated. Referring to FIG. 5 again, a filter layer 500 may be further disposed, which is disposed at a downstream side of the composite target 350. The filter layer 500 has a k-edge absorption energy, and the k-edge absorption energy is higher than the energy of the low energy photon of the X-ray and lower than the energy of the high energy photon of the X-ray. Therefore, the low energy photon part which is harmful to the human body can be filtered. Additionally, by using the filter layer 500 of which the atomic number is close to the atomic number of the composite target 350, the k-edge absorption energy close to the composite target 350 can be filtered, and the low energy photon of the X-ray can be filtered at the same time. Thus, the effect of the filter layer 500 on the characteristic radiation spectrum of the X-ray can be reduced. If the filter layer 500 is used with a thin target, a sharper single spectrum can be generated. In summary, the composite target and the X-ray tube with the composite target of the invention at least have the following advantages: The composite target has the interposing layer. By the interaction between the electron and the target body, and between the electron and the interposing layer, and the interaction generated from the X-ray and the interposing layer, the X-ray energy spectrum distribution is changed. Also, by the protective layer which can protect the composite target from excessive bombardment of the electron, the composite target has a sufficient long life. Additionally, by the setting method of a plurality of film layers of the target body and by changing the position where the electron enters the target, the X-ray with a designated energy spectrum distribution can be chosen. The X-ray tube with the above composite target can provide a variety of X-ray energy spectrum distributions. Additionally, the composite target can also be used to the related applications of the X-ray, not limited to the X-ray tube of the invention. Although the invention has been described with reference to the above embodiments, it will be apparent to one of ordinary skill in the art that modifications to the described embodiments may be made without departing from the spirit of the invention. Accordingly, the scope of the invention is defined by the attached claims not by the above detailed descriptions. |
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abstract | Example embodiments include a vapor forming apparatus, system and/or method for producing vapor from radioactive decay material. The vapor forming apparatus including an insulated container configured to enclose a nuclear waste container. The nuclear waste container includes radioactive decay material. The insulated container includes an inlet valve configured to receive vapor forming liquid. The radioactive decay material transfers heat to the vapor forming liquid. The insulated container also includes an outlet valve configured to output the vapor forming liquid heated by the radioactive decay material. |
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045089698 | description | DETAILED DESCRIPTION FIGS. 1 and 2 show a device 2 having a hollow-cylindrical container 6 open at the top 4 for holding, transporting and final storing of fuel elements 8 and 10. The interior 11 of the container can be circular-cylindrical (FIG. 2a) or rectangular or polygonal (FIG. 2b) in cross section. The container 6 is closed with a cover 12. The wall of the container 6 preferably is made in one piece, but it can also be made of several pieces. Carbon steel or high-grade steel is used as the fabricating material if the container walls are not too thick. With greater wall thickness carbon steel or spheroidal graphite iron is used. The wall thickness is selected such that gamma radiation is absorbed; thus, for instance, a thickness of 200 mm is sufficient to meet transportation limit values of 200 mrem/h on the surface. Spheroidal graphite iron has the advantage of a favorable price combined with ductility and a good shielding effect. The wall thickness also depends on the formation of the final storage place and on the corrosion induced by the environment on the container. Beyond that, of course, economic considerations are also important. A separate removable temporary outer shielding of spheroidal graphite iron can be used during transportation of the container and thus minimize the wall thickness of the final storage container. Such a design may have double walls and consist of an outer container and an inner container; this will be described in more detail with reference to FIG. 14. The peripheral shape of the container 6 is preferably circular because circular boreholes into which the containers are placed for the purpose of final storing are simpler to prepare. Surrounding the container 6 proper a cylindrical shielding layer 13, for instance, of a hydrocarbon such as polyethylene, in order to absorb the residual neutron radiation in the burned-out reactor elements. As a rule, 3 to 4 cm wall thickness are sufficient. This shielding is connected to the container in such a way that after transporting the container into the final storage place it can be removed for reuse. The free volume in the hollow space of the container can be filled by pouring in a filling material to improve the stability and the shielding against gamma radiation. Lead is especially suited for this purpose. The free volume to be filled in this manner, for compressed water reactor fuel elements, totals approximately 300 liters per fuel element in the case of a Biblis fuel element and to approximately the same amount in the case of four boiling water reactor fuel elements. The cover 12 is gas-tight and firmly connected to the container 6. For this purpose, the upper surface 14 of the wall of the container in the area surrounding the opening 4 terminates in a circular projection 16 having a profile as shown in FIGS. 3 and 4. In FIG. 3 the projection 16 is dovetailed and formed integrally with the wall. The cover 12 is cast around the projection 16 producing a complementary recess 18 whereby a very firm and tight connection of cover and container is achieved. To produce this connection a hollow mold is placed on the container after the fuel elements have been placed in the container and the hollow space was closed with a flat shielding cover 20 of high-grade steel (the shielding cover is drawn only schematically; details regarding its arrangement and special design will be given below in a more detailed manner with reference to FIG. 8). The mold is filled by pouring in a molten material, preferably the same material of which the container proper consists, whereby after the hardening of the poured material an intimate connection with the container is produced which is so firm that lifting of the container is possible, for instance, by means of a hook 22 which is cast into the cover. FIGS. 4 and 7 show variations of the contruction of FIG. 3 in which opposed dovetailed recesses are provided in opposed mating surfaces of the cover and container. In FIG. 7 the underside of the cover 12 is also provided with a center extension 23 insertable into the container 6 according to FIGS. 6 and 7. In these modifications, the cover 12 is already prefabricated. In the sealing surface 24, 26 of FIGS. 4 and 6 respectively, the cover is provided with dovetailed recesses 28 and 30 into which channels 32, 34 open. The recesses 28 and 30 are located opposite dovetailed recesses 36, 38 formed in the opposite sealing surfaces 40, 42 of the container 6. The channels 48 and 50 of FIGS. 5 and 7 respectively open directly into the sealing surfaces 44 and 46 of the cover at a point opposite recesses 52, 54 in the sealing surfaces 56, 58 of the con- tainer. In order to connect the cover 12 and the container 6, casting material is fed through the channels into the recesses. When the molten material fills and hardens in the channels and recesses a firm and gas-tight connection is produced. Screw connections and sealing elements can also be provided additionally or alternatively. The projections need not be dovetailed; they can have also other suitable shapes which preferably are narrower at the sealing edge than at the base. FIG. 8 shows in detail a preferred design for the cover zone of the device. The container 6, just as the container according to FIGS. 1 to 7, consists of a jacket 70, the bottom of which is not shown, and of a shielding cover 72. The shielding cover 72 has a protruding circumferential edge flange 74 which fits into a stepped recess 76 in the mouth of the jacket 70. An extension 78 of the shielding cover 72 protrudes into the hollow space 11 of the container 6. The edge flange 74 of the shielding cover 12 is secured to the jacket 70 by means of screws 80. A gasket 84 is provided for sealing the gap 82 between the shielding cover 72 and the stepped recess 76. The shielding cover preferably is made from spheroidal graphite iron. A relatively thin plate 86 covers the shielding cover 72 as well as the screws 80 and the gap 82. The cover plate 86 is welded flush to the top surface of the jacket wall. Above the cover plate 86 a final cover 12, as described before in connection with FIGS. 1 to 7, is cast onto the container by means of a suitable casting mold. Instead of the arched shape illustrated in FIGS. 1 to 7 the top cover can also be made flat as it is illustrated in FIG. 8. For the casting of the cover 12, the container 6 including the shielding cover and possibly the cover plate 86 is heated to a suitable temperature, for instance, 500.degree. to 600.degree. C. in order to preclude rapid cooling and thus obtain a uniform grain structure at the connection between cover and container jacket and prevent the development of a martensitic structure in the cast metal. The cover plate 86 prevents connecting the cover 12 with the shielding cover 72 and the screws 80. Thereby the container remains accessible in a simple manner. The cover 12 may be removed together with the cover plate 86. Then the opening of the container is possible after the loosening of the screws and the removal of the shielding cover. The jacket 70 is provided on its top edge with a projection 88 which may take the form of dovetailed individual segmental projections or of a dovetailed annular rib. These projections may also take other suitable shapes. After the cover 12 has been put on or cast on, the projections guarantee a firm and secure connection between the container 6 and the cover 12. For a better handling of the container, lifting lugs 90 can be attached to the side of the jacket 70. These lifting lugs are preferably detachable. Also to facilitate handling the cover 12 can be provided with a hook 92 which is preferably detachable. In place of projection 88, it is possible to provide in the top edge of the jacket a recess 94 (shown in broken lines) into which the casting material is fed during the casting of the cover. A mold (not shown) is placed on top of the container, into which the casting material is fed and which produces the shape of the cover. FIGS. 9 and 10 show two further variations for the cover of the container 6. In both types, the jacket 110 of the container is provided inside with a stepped recess 112 and a shielding cover 114 of similar construction to FIG. 8. A top cover 116 is recessed so that its top surface 118 is spaced slightly above the top surface of the jacket wall. For this type, the cover 116 is prefabricated and has channels 120 which open into the lateral surfaces of the cover opposite channels 122. As illustrated, parts of the channels can be dovetailed as described before in connection with FIGS. 4 to 6. After the prefabricated cover has been put on, casting material is fed into those channels and into dovetailed recesses 30 by way of filling orifices 124 and 126. Upon solidification the solid metal results in a firm connection between cover and container. FIG. 11 shows another modification in the cover zone of the container 6 where the shielding cover 114 is designed approximately like the shielding cover according to FIG. 8 and is connected with the container. The cover 128 is also prefabricated and provided with casting channels 130 and filling orifice 132 approximately as shown in FIG. 5. It has a shape arched outwardly, for instance, like the cover according to FIGS. 3 to 5. In this construction, dovetailed recesses 134 are provided in the top edge of the jacket with which channels 130 communicate as described in connection with FIG. 5. In FIG. 12a, b, c, d, some examples of cross section shapes suitable for the projections on the top surface of the container jacket are shown. The shapes according to FIG. 12a and 12d result in a firmer connection because of the undercut design and are preferred. FIG. 13 shows recesses 136, formed in the wall of jacket 70 of the container 6, with air bleed ducts 138 in order to ensure that the recess is completely filled with casting material. FIG. 14a and 14b show a separate inner container 140 for holding fuel elements. The inner container consists of a jacket 142, a cover 144 and a bottom 146. Cover and bottom are welded to the jacket at 148 and 150. The bottom can be cast in one piece with the jacket, or cast on separately. The cover can be put on by casting or by welding. The cover and the bottom can be arched inward (FIG. 14a), arched outward (FIG. 14b), or also be straight (shown by broken lines in FIG. 14b). During transporting, the inner container is inserted into an outer container or transport container which is designed like the container according to FIGS. 1 to 13; compare especially FIGS. 1, 2a and 2b in which the inside container 140 is shown in broken lines and the outer container comprises the container 6. Such a double-container has several advantages. In connection with the final storing, only the inner container is lost. The outer container can be reused; it can be salvaged during the transfer at the borehole of the final storage site. The inner container and the outer container can be constructed from the same materials and in the same manner. High-grade steel or casting material is also preferable for the inner container. If carbon steel is used, ceramic material or another corrosion-protecting layer is put on. Preferably the outer shape of the inner container corresponds to the inner shape of the outer container. The thickness of the material for the inner container is selected in such a way that the minimum requirements regarding the shielding effect and the stability are met. The outer container must be constructed so that transportation specifications are met and protection against corrosion is guaranteed. For protection against corrosion the container can be provided with a ceramic layer. This can be carried out, for instance, by the spraying on the appropriate material. For reasons of completeness there may also be mentioned that a lock system can be provided in the zone of the cover in order to make it possible to take a sample from the container and to carry out supervisory tasks. |
summary | ||
claims | 1. A radiographic apparatus for obtaining a radiographic image, comprising:a radiation source for emitting radiation;a radiation detecting device having a plurality of detecting elements arranged two-dimensionally in rows and columns for detecting the radiation;a radiation grid with absorbing foil strips extending in a direction of the rows and arranged in a direction of the columns for removing scattered radiation;a physical quantity acquiring device for calculating predetermined physical quantities to determine pixel values of pixels arranged two-dimensionally;a physical quantity map generating device for generating a physical quantity map by mapping the predetermined physical quantities; anda physical quantity map smoothing device for smoothing the physical quantities arranged on the physical quantity map in the direction of extension of the absorbing foil strips, thereby to generate an average value map. 2. The radiographic apparatus according to claim 1, further comprising:a pixel specifying device for specifying certain pixels among pixels forming the radiographic image; andan intensity estimating device for estimating at least one of scattered radiation intensity at the certain pixels specified by the pixel specifying device, and direct radiation intensity at the certain pixels;wherein(A) a rate of change calculating device is provided as a component corresponding to the physical quantity acquiring device, for determining a rate of change for each pixel relative to an average value or a value of each pixel obtained by smoothing and interpolating calculations as a reference intensity for the pixels relating to the radiation intensity, using the radiation intensity estimated by the intensity estimating device based on actual measurement carried out in the presence of a subject;(B) a rate of change map generating device is provided as a component corresponding to the physical quantity map generating device, for generating a rate of change map by mapping the rate of change for each pixel; and(C) a rate of change map smoothing device is provided as a component corresponding to the physical quantity map smoothing device, for smoothing rates of changes arranged on the rate of change map in the direction of extension of the absorbing foil strips, thereby to generate the average value map. 3. The radiographic apparatus according to claim 2, wherein the intensity estimating device is arranged to estimate radiation intensity at the certain pixels specified by the pixel specifying device, based on the average value map, direct radiation transmittance calculated by the transmittance calculating device, and actual measurement intensity which is a radiation intensity after transmission through the scattered radiation removing device in actual measurement carried out in the presence of a different subject. 4. The radiographic apparatus according to claim 3, wherein the physical quantity map smoothing device is arranged to remove influences of statistical noise superimposed on the physical quantity map. 5. The radiographic apparatus according to claim 4, wherein spacing between the absorbing foil strips of the radiation grid adjoining in the direction of the columns is synchronized with an integral multiple of spacing between the detecting elements of the radiation detecting device adjoining in the direction of the columns. 6. The radiographic apparatus according to claim 4, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 7. The radiographic apparatus according to claim 3, wherein spacing between the absorbing foil strips of the radiation grid adjoining in the direction of the columns is synchronized with an integral multiple of spacing between the detecting elements of the radiation detecting device adjoining in the direction of the columns. 8. The radiographic apparatus according to claim 3, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 9. The radiographic apparatus according to claim 2, wherein the physical quantity map smoothing device is arranged to remove influences of statistical noise superimposed on the physical quantity map. 10. The radiographic apparatus according to claim 9, wherein spacing between the absorbing foil strips of the radiation grid adjoining in the direction of the columns is synchronized with an integral multiple of spacing between the detecting elements of the radiation detecting device adjoining in the direction of the columns. 11. The radiographic apparatus according to claim 9, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 12. The radiographic apparatus according to claim 2, wherein spacing between the absorbing foil strips of the radiation grid adjoining in the direction of the columns is synchronized with an integral multiple of spacing between the detecting elements of the radiation detecting device adjoining in the direction of the columns. 13. The radiographic apparatus according to claim 12, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 14. The radiographic apparatus according to claim 2, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 15. The radiographic apparatus according to claim 1, wherein the physical quantity map smoothing device is arranged to remove influences of statistical noise superimposed on the physical quantity map. 16. The radiographic apparatus according to claim 15, wherein spacing between the absorbing foil strips of the radiation grid adjoining in the direction of the columns is synchronized with an integral multiple of spacing between the detecting elements of the radiation detecting device adjoining in the direction of the columns. 17. The radiographic apparatus according to claim 15, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 18. The radiographic apparatus according to claim 1, wherein spacing between the absorbing foil strips of the radiation grid adjoining in the direction of the columns is synchronized with an integral multiple of spacing between the detecting elements of the radiation detecting device adjoining in the direction of the columns. 19. The radiographic apparatus according to claim 18, further comprising a C-arm for supporting the radiation source and the radiation detecting device. 20. The radiographic apparatus according to claim 1, further comprising a C-arm for supporting the radiation source and the radiation detecting device. |
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060193263 | abstract | A video enhancement kit is provided for use with a support, a video camera, a magnification lens assembly, and a plurality of couples. The kit preferably includes an elongated base strip removably coupled to the support with the video camera removably coupled thereon at a user selected position along a length of the base strip. Also included is a mount removably coupled to the elongated base strip via one of the couples. The mount is further removably coupled to the magnification lens assembly via one of the couples for aligning the magnification lens assembly with the video camera. |
040594831 | claims | 1. In a nuclear reactor fuel assembly, the fuel assembly including a plurality of zircaloy spacer grids which locate and support elongated fuel elements which pass therethrough, said fuel assembly further including a plurality of guide tubes which extend through the fuel assembly parallelly of the fuel elements and which receive movable neutron absorber elements, the improvement comprising: at least a first seismic grid positioned intermediate a pair of spacer grids, said seismic grid having a greater resistance to deformation when compared to the spacer grids and having larger external dimensions than said spacer grids, said seismic grid including: a central portion, said central portion having said openings and projections formed therein; and oppositely disposed edge portions, said edge portions being bent inwardly toward the fuel assembly at an angle of less than . tube segments mounted in said seismic grid so as to be in alignment with the fuel assembly guide tubes, said tube segments having inner diameters greater than the outer diameters of aligned guide tubes; means joining divider members which intersect said tube segments to said tube segments; and means attaching said tube segments to the exterior of said guide tubes. tube segments mounted in aid seismic grid so as to be in alignment with the fuel assembly guide tubes, said tube segments having inner diameters greater than the outer diameters of aligned guide tubes; means joining divider members which intersect said tube segments to said tube segments; and means attaching said tube segments to the exterior of said guide tubes. tube segments mounted in aid seismic grid so as to be in alignment with the fuel assembly guide tubes, said tube segments having inner diameters greater than the outer diameters of aligned guide tubes; means joining divider members which intersect said tube segments to said tube segments; and means attaching said tube segments to the exterior of said guide tubes. 2. The apparatus of claim 1 wherein said divider member define grid sectors which each receive a plurality of fuel elements. 3. The apparatus of claim 1 wherein said perimeter strip openings are generally aligned with fuel elements disposed adjacent to said perimeter strip in the outer row of grid sectors defined by said perimeter strip and divider members. 4. The apparatus of claim 3 wherein said perimeter strip projections comprise inwardly extending dimples formed in said perimeter strip, said dimples being located above and below each of said openings. 5. The apparatus of claim 4 wherein said divider members are provided with projections extending into each of said grid sectors so as to limit motions of the fuel elements passing therethrough. 6. The apparatus of claim 5 wherein said divider member define grid sectors which each receive a plurality of fuel elements. 7. The apparatus of claim 1 wherein said perimeter strip comprises: 8. The apparatus of claim 6 wherein said perimeter strip openings are generally aligned with fuel elements disposed adjacent to said perimeter strip in the outer row of grid sectors defined by said perimeter strip and divider members. 9. The apparatus of claim 8 wherein said perimeter strip projections comprise inwardly extending dimples formed in said perimeter strip, said dimples being located above and below each of said openings. 10. The apparatus of claim 9 wherein said divider member define grid sectors which each receive a plurality of fuel elements. 11. The apparatus of claim 10 wherein said perimeter strip edge portions are serrated. 12. The apparatus of claim 1 wherein said means attaching divider members to guide tubes comprises: 13. The apparatus of claim 6 wherein said means attaching divider members to guide tubes comprises: 14. The apparatus of claim 11 wherein said means attaching divider members to guide tubes comprises: |
abstract | Apparatus and methods for determining a system matrix for pinhole collimator imaging systems are provided. One method includes using a closed form expression to determine a penetration term for a collimator of the medical imaging system and determining a point spread function of the collimator based on the penetration term. The method further includes calculating the system matrix for the medical imaging system based on the determined point spread function. |
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description | This application claims benefit of PCT/CN2003/000985 filed Nov. 20, 2003. The present invention relates to a collimator and a radiation irradiator. Particularly, the invention is concerned with a collimator for a radiation, e.g., X-ray, and a radiation irradiator provided with the collimator. In a radiation irradiator there is used a collimator for controlling the irradiation range of a radiation. The collimator has an aperture which permits a radiation to pass therethrough, and the radiation cannot pass through other than the aperture. Thus, the irradiation range of the radiation is controlled by the aperture. The degree of opening of the aperture can be changed to adjust the irradiation range of the radiation. As shown in FIG. 10, the aperture-adjustable collimator has a pair of movable plate members having a radiation shielding property, i.e., blades 901 and 901′. The blades 901 and 901′ are disposed so that respective end faces are opposed to each other, and are movable in directions opposite to each other in a plane parallel to their surfaces. For widening the aperture, the pair of blades 901 and 901′ are moved away from each other, while for narrowing the aperture, the both blades are moved to close to each other. Thus, the aperture becomes maximum when the pair of blades are remotest from each other, and becomes minimum when both blades are closest to each other. In a collimator having a large aperture adjusting range, there are used blades of a large area. If the blade area is large, an external form of the collimator comes to have a large size corresponding to a maximum moving distance of blade outer edges, as shown in FIG. 11. In a mammography apparatus which makes fluoroscopy of the breast with X-ray, a collimator assumes a position confronting the face of a subject and so it is preferable that its external form be as small as possible. However, in the case of a mammography apparatus capable of making tomography, an increase in external form of the collimator has heretofore been unavoidable because a large aperture is needed. Therefore, it is an object of the present invention to provide a collimator capable of being reduced in its external form without sacrificing an aperture, as well as a radiation irradiator provided with such a collimator. (1) In one aspect of the present invention for achieving the above-mentioned object there is provided a collimator comprising a pair of first plate members having a shielding property against a radiation and movable in a direction parallel to surfaces thereof, the pair of first plate members defining a radiation passing apertures by a spacing between respective opposed end faces, a pair of second plate members having a shielding property against a radiation and parallel to the pair of first plate members and movable in a direction parallel to surfaces thereof, the pair of second plate members having end faces opposed to each other, the pair of second plate members overlapping the pair of first plate members at least partially so as to block any other radiation than the radiation passing through the aperture, a pair of third plate members having a shielding property against a radiation and parallel to the pair of second plate members, the pair of third plate members having respective end faces opposed to each other with a predetermined spacing, the pair of third plate members overlapping the pair of second plate members at least partially so as to block any other radiation than the radiation passing through the aperture, an adjusting mechanism which adjusts the aperture by moving the pair of first plate members, and a follow-up mechanism which causes the pair of second plate members to move following the pair of first plate members with movement of the first plate members. (2) In another aspect of the present invention for achieving the above-mentioned object there is provided a radiation irradiator having a radiation source and a collimator for applying a radiation from the radiation source to an object through an aperture, the collimator comprising a pair of first plate members having a shielding property against a radiation and movable in a direction parallel to surfaces thereof, the pair of first plate members defining a radiation passing aperture by a spacing between respective opposed end faces, a pair of second plate members having a shielding property against a radiation and parallel to the pair of first plate members and movable in a direction parallel to surfaces thereof, the pair of second plate members having end faces opposed to each other, the pair of second plate members overlapping the pair of first plate members at least partially so as to block any other radiation than the radiation passing through the aperture, a pair of third plate members having a shielding property against a radiation and parallel to the pair of second plate members, the pair of third plate members having respective end faces opposed to each other with a predetermined spacing, the pair of third plate members overlapping the pair of second plate members at least partially so as to block any other radiation than the radiation passing through the aperture, an adjusting mechanism which adjusts the aperture by moving the pair of first plate members, and a follow-up mechanism which causes the pair of second plate members to move following the pair of first plate members with movement of the first plate members. In the above aspects of the present invention, since movable blades are constituted by two sets of plate members so that corresponding ones overlap each other, it is possible to reduce the external form of the collimator without sacrificing the aperture. For changing the aperture symmetrically with respect to the center of the collimator, it is preferable that the adjusting mechanism be able to move the pair of first plate members so as to be closed to and away from each other. For permitting an appropriate follow-up motion of the pair of second plate members with respect to the pair of first plate members, it is preferable that the follow-up mechanism comprise a rack provided in the first plate member, a gear provided in the second plate member rotatably and engaging with the rack, and a fixed rack provided in the moving direction of the second plate member and engaging with the gear. For the simplification of construction, it is preferable that the follow-up mechanism comprise an arm member mounted at an intermediate portion thereof to the second plate member and rotatable about the mounting portion in a plane parallel to the plate surface, a groove formed in the first plate member and with which one end of the arm member is engaged, the groove permitting movement of the end of the arm member in a direction perpendicular to the moving direction of the first plate member, and a groove formed in the third plate member and with which an opposite end of the arm member is engaged, the groove permitting movement of the opposite end of the arm member in a direction perpendicular to the moving direction of the second plate member. For facilitating the adjustment of dose, it is preferable that the radiation be X-ray. According to the present invention, it is possible to realize a collimator capable of being reduced in its external size without sacrificing the aperture, as well as a radiation irradiator provided with such a collimator. Embodiments of the present invention will be described in detail hereinunder with reference to the drawings. FIG. 1 illustrates a schematic construction of a radiation irradiator according to an embodiment of the present invention. The construction of this radiation irradiator shows an example of how to carry out the present invention. As shown in the same figure, the radiation irradiator has a radiation source 1. As the radiation source 1 there is used an X-ray tube for example. The radiation source 1 is not limited to the X-ray tube, but may be any other radiation source capable of emitting a suitable radiation such as β ray or γ ray. The radiation source 1 is an example of the radiation source used in the present invention. A radiation 3 emitted from the radiation source 1 passes through an aperture of a collimator 5 which embodies the present invention and is applied to an object 7. The object 7 is an object for fluoroscopy using the radiation 3 or an object for therapy using the radiation 3. In fluoroscopy, the radiation which has passed through the object 7 is received by a suitable light receiving means, e.g., a photosensitive plate. FIG. 2 schematically illustrates the collimator 5, which embodies the present invention. Through the constitution of this, one embodiment of the present invention is described. As shown in the same figure, the collimator 5 has three pairs of blades 501, 501′, 503, 503′, and 505, 505′. Each blade is constituted by a quadrangular plate member. As the material of the plate members there is used a material of a high radiation absorbance such as, for example, lead (Pb) or tungsten (W), whereby each blade comes to have a radiation shielding property. The blades 501 and 501′ are an example of the pair of first plate members in the present invention. The blades 503 and 503′ are an example of the pair of second plate members in the present invention. The blades 505 and 505′ are an example of the pair of third plate members in the present invention. In FIG. 2, three directions perpendicular to one another assumed to be x, y, and z. The x direction is a direction of one side of each blade, and the direction will hereinafter be referred to also as the width direction. The y direction is a direction of another side of each blade, and the direction will hereinafter be referred to also as the length direction. The z direction is the thickness direction of each blade. The radiation source 1 lies in the z direction. The three pairs of blades 501, 501′, 503, 503′, and 505, 505′ overlap one another in the thickness direction. The blades 501 and 501′ are upper blades, the blades 503 and 503′ are intermediate blades, and the blades 505 and 505′ are lower blades. The lower blades 505 and 505′ are supported by a pair of cross beams 507 and 507′. The cross beams 507 and 507′ are also constituted by a material of a high radiation absorbance. The cross beams 507 and 507′ constitute a picture frame-like frame together with the blades 505 and 505′. The upper blades 501 and 501′ are made movable in the x direction by means of a drive mechanism which will be described later. The blades 501 and 501′ are movable so as to be close to and away from each other. The intermediate blades 503 and 503′ are made movable following the upper blades 501 and 501′ by means of a follow-up mechanism to be described later. The movement of the blades 503 and 503′ is done while maintaining their overlap with the blades 501, 501′ and the blades 505, 505′ constantly. The size of the aperture through which the radiation passes is determined in the x direction by the spacing between opposed end faces of the blades 501 and 501′ and in the y direction by the spacing between opposed end faces of the cross beams 507 and 507′. The aperture size in the x direction varies with movement of the blades 501 and 501′, while the aperture size in the y direction is fixed. That is, the collimator 5 has an aperture whose size in the x direction can be changed. A fully closed state of the aperture is shown in FIG. 3 and a fully open state thereof is shown in FIG. 4. Throughout the whole process of aperture changes, the blades 503 and 503′ are kept overlapped with the blades 501, 501′ and the blades 505, 505′. Consequently, the passage of any other radiation than the radiation passing through the aperture is blocked. FIG. 5 schematically illustrates the construction of a drive mechanism for the blades 501 and 501′. This drive mechanism is an example of the adjusting mechanism in the present invention. As shown in the same figure, the blades 501 and 501′ have arms 601 and 601′, respectively, which extend in the y direction. End portions of the arms 601 and 601′ are in engagement with shafts 603 and 603′, respectively. The shafts 603 and 603′ are parallel shafts extending in the x direction. Both shafts are spaced a predetermined distance in the z direction. The arm 601′ is bent to equalize the height in the z direction of the blade 501′ to that of the blade 501. The shafts 603 and 603′ are threaded throughout the overall lengths thereof. The arms 601 and 601′ are internally threaded at their portions engaged with the shafts 603 and 603′. Gears 605 and 605′ are provided coaxially at one ends of the shafts 603 and 603′ respectively. The gears 605 and 605′ are in mesh with each other at a gear ratio of 1:1. The gear 605 is rotated by means of a motor 607. The motor 607, which is a reversible motor, is controlled by a control means (not shown). The control means controls both rotational direction and rotational quantity of the motor 607. Since the gears 605 and 605′ are in mesh with each other, the shafts 603 and 603′ rotate in directions opposite to each other. Consequently, the arms 601 and 601′ engaged with the shafts 603 and 603′ move reverse to each other in the x direction. That is, as the motor 607 rotates in one direction, both arms move to be close to each other, while as the motor 607 rotates in the opposite direction, both arms move away from each other. Their movement quantity is determined by the amount of rotation of the motor 607. By such movements of the arms 601 and 601′ there is adjusted the spacing between the blades 501 and 501′, i.e., the degree of opening of the aperture. In this way the degree of opening of the aperture can be changed symmetrically with respect to the center of the collimator. There is provided a follow-up mechanism for allowing the blade 503 and 503′ to follow such movements of the blades 501 and 501′. The follow-up mechanism is constituted so as to span the three pairs of blades and the cross beams. As a part of the follow-up mechanism, the blade 503 is constituted as shown in FIG. 6, in which (a) is a plan view and (b) is a sectional view taken on line A-A in (a). As shown in the same figure, the blade 503 has a gear 701. The gear 701 is mounted rotatably on a shaft 703 which is provided in the blade 503. To be more specific, the shaft 703 is disposed within a cutout portion 705 formed in one end portion in the y direction of the blade 503. The cutout portion 705 is formed in the x direction and the shaft 703 is mounted so as to cross the cutout portion 705 in the y direction. The cutout portion 705 is formed on one side in the x direction of the blade 503. The side where the cutout portion 705 is formed confronts the blade 503′ which makes a pair with the blade 503. The blade 503 has a similar gear also at its opposite end portion in the y direction. That is, the blade 503 has gears at both ends thereof in the y direction. The blade 503′ is also of the same construction, provided the blades 503 and 503′ are in a relation of specular symmetry. FIG. 7 illustrates the construction of the follow-up mechanism schematically. This follow-up mechanism is an example of the follow-up mechanism defined in the present invention. As shown in the same figure, the follow-up mechanism is composed of the gear 701 provided in the blade 503, a rack 707 provided in the blade 501 and meshing with the gear 701, and a rack 709 provided in both blade 505 and cross beam 507 and meshing with the gear 701. The racks 707 and 709 extend in the x direction in parallel with each other. The gear 701 is an example of the gear defined in the present invention. The rack 707 is an example of the rack defined in the present invention. The rack 709 is an example of the fixed rack defined in the present invention. As the blade 501 is moved in the x direction, the gear meshing with the rack 707 moves in the same direction while rotating on the rack 709. The blade 503 also moves together with the gear 701. As a result, the blade 503 moves following the blade 501. The distance of the movement of the blade 503 is a half of that of the blade 501. Such a follow-up mechanism is provided at both end portions of the blades 501, 503, and 505. This is also the case with the mating blades 501′, 503′, and 505′. With the follow-up mechanism, the follow-up motion of the blades 503 and 503′ for the blades 501 and 501′ can be done appropriately. Consequently, it becomes possible to make such aperture adjustment as shown in FIGS. 2 to 4. FIG. 8 schematically shows another constructional example of a follow-up mechanism. This follow-up mechanism is an example of the follow-up mechanism defined in the present invention. In the same figure, (a) is a plan view and (b) is a sectional view taken on line B-B in (a). As shown in the same figure, the blade 503 is provided with a shaft 803 at its center. The shaft 803 extends through the blade 503 perpendicularly to the plate surface. The shaft 803 is rotatable. Arms 801 and 805 are fixed respectively to both ends of the shaft 803. The arms 801 and 805 are perpendicular to the shaft 803 and extend in directions opposite to each other. The shaft 803 and the arms 801, 805 form a crank. The extending directions of the arms 801 and 805 in the crank are not coincident with the x direction. The arms 801 and 805 are an example of the arm member defined in the present invention. The arms 801 and 805 are formed with lugs 811 and 851 at respective free ends. The lugs 811 and 851 extend in the z direction so as to face reverse to each other. The lug 811 is loosely fitted in a groove 511 formed in the blade 501. The groove 511 is positioned on the blade 503 side of the blade 501 and extends in the y direction. The lug 851 is loosely fitted in a groove 551 formed in the blade 505. The groove 551 is positioned on the blade 503 side of the blade 505 and extends in the y direction. The grooves 511 and 551 are an example of the grooves defined in the present invention. In this construction, when the blade 501 is moved in the x direction, the lugs 811 and 851 move in directions opposite to each other along the grooves 511 and 551, with the result that the crank rotates about the shaft 803. The position of the blade 505 is fixed, so with the rotation of the crank, the shaft 803 moves in the x direction following the blade 501 and the blade 503 moves in the same direction together with the shaft 803. By setting the lengths of the arms 801 and 805 equal to each other, the distance of the movement of the blade 503 becomes half of that of the blade 501. Thus, this follow-up mechanism becomes simple in construction. Such a follow-up mechanism is provided also on the side of the mating blades 501′, 503′, and 505′. As a result, it becomes possible to make such aperture adjustment as shown in FIGS. 2 to 4. In the fully open condition of the aperture shown in FIG. 4, corresponding ones in the three pairs of blades 501, 501′, 503, 503′, 505, and 505′ overlap each other completely. FIG. 9 shows this state in terms of a sectional view. As shown in the same figure, the movable blades 501, 501′, 503, and 503′ overlap the fixed blades 505 and 505′ completely. In this state, outer edges of the movable blades are in alignment with outer edges of the fixed blades, not protruding therefrom. Consequently, an external form of the collimator becomes constant irrespective of the degree of opening of the aperture. Therefore, if the maximum degree of opening of the aperture is set equal to that in the prior art shown in FIG. 11, it is possible to reduce the external form of the collimator to about three fourths. Alternatively, if the external form of the collimator is made about the same as in the prior art, it is possible to enlarge the maximum value of the aperture to approximately 4/3 time. Many widely different embodiments of the invention may be configured without departing from the spirit and the scope of the present invention. It should be understood that the present invention is not limited to the specific embodiments described in the specification, except as defined in the appended claims. |
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description | This application is the national phase of PCT application PCT/KR2014/009348 having an international filing date of 2 Oct. 2014, which claims the benefit of earlier filing date and right of priority to Korean Application No. 10-2013-0118741, filed on Oct. 4, 2013, and Korean Application No. 10-2014-0132166, filed on Oct. 1, 2014, the contents of which are incorporated by reference herein in their entirety. 1. Field of the Invention The present disclosure relates to a passive safety facility to enhance cooling performance and to reduce radioactive materials within the containment during an accident by introducing a fluid circulation increasing device and filter facility, and a nuclear power plant including the same. 2. Description of the Related Art Reactor is divided into a loop type reactor and an integrated type reactor according to the installation location of main components. When the main components (a steam generator, a pressurizer, a pump, etc.) are installed outside a reactor vessel, it is classified as a loop type reactor (for example, commercial reactor: Korea). When the main components are installed within a reactor vessel, however, it is classified as an integrated type reactor (for example, SMART reactor: Korea). Furthermore, reactor is divided into an active type reactor and a passive type reactor according to the implementation method of a safety system. The active type reactor uses an active component such as a pump operated by electric power from an emergency diesel generator or the like to drive the safety system. On the other hand, the passive type reactor uses a passive component operated by passive power such as gravity, gas pressure or the like to drive the safety system. When an accident occurs in the passive type reactor, the passive safety system maintains the reactor in a safe condition for at least a period of time (72 hours) according to the regulatory requirements by using only natural forces integrated in the system even without any operator actions or alternating current (AC) power of a safety class such as an emergency diesel. The operator action or non-safety system involvement is allowable in the passive safety system operation after 72 hours following the design basis accidents. In the related art, a passive residual heat removal system and a passive containment cooling system have been configured using a steel containment and a secondary side of a steam generator (Korean Patent Publication No. 10-2013-0047871). However, it is preferable to adopt the reinforced concrete-type containment building rather than the steel containment applied to the related art due to difficulties in manufacturing and maintenance, low economic efficiency, and the like. Hereinafter, a passive residual heat removal system and a passive containment cooling system will be described, respectively. The passive residual heat removal system is employed to remove heat in a reactor coolant system (sensible heat in the reactor coolant system and residual heat in the core) during an accident in various reactors including the integrated type reactor. For a cooling water circulation method of the passive residual heat removal system, two types are mainly used, such as a primary fluid circulating method to cool the reactor coolant system (AP1000: United States Westinghouse Company) and a secondary fluid circulating method using a steam generator to cool the reactor coolant system (SMART reactor: Korea), and a method of injecting primary fluid to a condensation tank to directly condensate it (CAREM: Argentina) is also partly used. Furthermore, for a cooling method of the heat exchanger (condensation heat exchanger) in the passive residual heat removal system, a water-cooled type (AP1000) that is used most frequently, an air-cooled type (WWER 1000: Russia), and a water-air hybrid cooled type (IMR: Japan) are used. The heat exchanger in the passive residual heat removal system transfers heat received from the reactor coolant system to an outside (ultimate heat sink) through an emergency cooling water storage section or the like. For a method of heat exchanger, a condensation heat exchanger with high heat transfer efficiency using steam condensation phenomenon is mainly used. The passive containment cooling system is one of several safety systems for reducing pressure, temperature, and a concentration of radioactive materials. The passive containment cooling system is employed to suppress the increase of pressure and to remove the heat within the containment (reactor building, containment or safeguard vessel) during an accident in various nuclear reactors including the integrated type reactor. For a configuration method of the function of the passive containment cooling system, a method of using a suppression tank for condensing the steam discharged to the containment building (Commercial BWR, CAREM: Argentina, IRIS: United States Westinghouse Company, etc.), a method of applying a steel containment and cooling an outer wall (water spray, air cooling) (AP1000: United States Westinghouse Company), and a method of using a heat exchanger (SWR1000: France Framatome ANP, AHWR: India, SBWR: United States GE), and the like are used. As described above, in general, the passive residual heat removal system using the secondary side of the steam generator mainly uses a method of installing a heat sink (emergency cooling water storage section) outside the containment using a natural circulation. The cooling water that is cooled within the condensation heat exchanger is supplied to the steam generator by gravitational force, and the steam that is formed within the steam generator while removing the heat of the reactor coolant system is circulated to the condensation heat exchanger continuously (Korean Patent Publication No. 10-2013-0047871). On the other hand, the passive containment cooling system generally uses a method of removing heat within the containment by natural circulation flow formed within the containment without any other flow inducing means. Furthermore, the passive residual heat removal system and passive containment cooling system are typically designed independently. The performance of the passive containment cooling system in the related art has been determined only by natural circulation flow without any additional devices for forming the flow thereof. In such a configuration where natural convection is dominant, the heat transfer coefficient on a surface over which the atmosphere (air and steam) flows is very small, thereby causing a problem in which the size of the heat exchanger should be greatly increased. Meanwhile, the heat exchanger may be a structure of forming a pressure boundary of the containment, thus a possibility of the pressure boundary break may increase when the size of the heat exchanger increases, thereby causing a problem of decreasing the safety. An object of the present disclosure is to propose a passive safety facility with an enhanced cooling performance within the containment using a facility which can overcome the limitation of the natural circulation, and a nuclear power plant including the same. Another object of the present disclosure is to provide a passive safety facility capable of performing both functions of a passive residual heat removal system and a passive containment building cooling system, and discharging heat transferred from a reactor coolant system to the external environment, and a nuclear power plant including the same. Still another object of the present disclosure is to disclose a passive safety facility capable of enhancing a heat transfer performance without increasing a size of the heat exchanger, and a nuclear power plant including the same. Yet still another object of the present disclosure is to disclose a passive safety facility capable of enhancing circulation flow within the containment to reduce a concentration of radioactive materials within the containment at an early stage, and a nuclear power plant including the same. Still yet another object of the present disclosure is to provide a passive safety facility capable of collecting radioactive materials to reduce the concentration of the radioactive materials within the containment at an early stage, and a nuclear power plant including the same. In order to accomplish an object of the foregoing aspects, a passive safety facility associated with an embodiment of the present disclosure may include a cooling section formed to cool a first fluid discharged from a reactor coolant system or steam generator along with a second fluid within a containment; and a circulation inducing jet device formed to jet the first fluid discharged from the reactor coolant system or the steam generator to the cooling section, at least part of which is open toward an inside of the containment to entrain the second fluid by a pressure drop caused while jetting the first fluid so as to jet the entrained second fluid along with the first fluid. The circulation inducing jet device may include a first fluid jetting section connected to the reactor coolant system or the steam generator to receive the first fluid, and formed to jet the received first fluid; a second fluid entraining section formed in an annular shape around the first fluid jetting section to entrain the second fluid within the containment; and a circulating fluid jetting section configured to surround the first fluid jetting section with a portion having an inner diameter larger than that of the first fluid jetting section to form the second fluid entraining section, and supply the first fluid and the second fluid to the cooling section. The first fluid jetting section may include a nozzle configured to jet the first fluid to the circulating fluid jetting section, and the circulating fluid jetting section may include a throat formed with an inner diameter smaller than that of an inlet of the nozzle to cause a local pressure drop while jetting the first fluid; and a diffuser configured to induce the first fluid and the second fluid that have passed through the throat to the cooling section. The circulation inducing jet device may include a turbine blade rotatably installed at an outlet of the first fluid jetting section to induce the jetting of the first fluid; and a pump impeller connected to the turbine blade to rotate along with the turbine blade, and induce the entrainment of the second fluid through the second fluid entraining section. The passive safety facility may further include a filter facility connected to an outlet of the cooling section to filter out noncondensible gas discharged from the cooling section, and collect radioactive materials filtered out from the noncondensible gas. The filter facility may include a filter or absorbent configured to separate the radioactive materials from the noncondensible gas; a gas discharge section configured to discharge noncondensible gas filtered out while passing through the filter or absorbent to an inside of the containment; and a gas line connected to the outlet of the cooling section to supply the noncondensible gas to the filter or absorbent. The cooling section may cool the first fluid and the second fluid to form condensate, and the passive safety facility may further include a cooling water storage section formed to store cooling water therein, and the cooling water storage section may be installed below the cooling section to collect the condensate discharged from the cooling section. The cooling water storage section may include a first cooling water storage section configured to store pure cooling water to be supplied to the steam generator so as to remove sensible heat within the reactor coolant system and residual heat in a core; and a second cooling water storage section configured to store borated water to be injected into the reactor coolant system so as to maintain a level of the reactor coolant system. The passive safety facility may further include an additive injection section configured to inject an additive into the condensate for suppressing the revolatilization of condensate collected in the cooling water storage section, and the additive may be formed to maintain a pH of the condensate above a preset value. The cooling water storage section may be configured to collect the condensate in the first cooling water storage section, and flow the condensate collected in the first cooling water storage section into the second cooling water storage section when a level of the collected condensate exceeds a reference level, and the additive injection section may be installed at a flow path connected from the first cooling water storage section to the second cooling water storage section to inject the additive into the condensate flowing into the second cooling water storage section. The additive injection section may be installed at a flow path connected from the cooling section to the cooling water storage section to inject the additive to condensate collected in the cooling water storage section. The passive safety facility may further include a condensate holding section installed between the cooling section and the cooling water storage section to collect the condensate falling from the cooling section so as to return it to the cooling water storage section. The passive safety facility may further include a return line extended from an outlet of the cooling section or the condensate holding section to the cooling water storage section to return condensate generated during the cooling process of the cooling section to the cooling water storage section. The passive safety facility may further include a filter facility connected to an outlet of the cooling section to filter out noncondensible gas discharged from the cooling section so as to collect radioactive materials filtered out from the noncondensible gas, and the return line may include a first return line connected to a lower portion of the casing to form a flow path of the condensate collected in the casing; and a second return line connected to a lower portion of the filter facility to form a flow path of the condensate collected in the filter facility. The passive safety facility may further include a fluid circulation section configured to circulate the cooling water of the cooling water storage section to the circulation inducing jet device through the reactor coolant system or the steam generator, and the fluid circulation section may include a fluid supply line connected to the cooling water storage section to supply cooling water within the cooling water storage section to the reactor coolant system or the steam generator; and a steam discharge line connected to the reactor coolant system or the steam generator and the circulation inducing jet device to supply the first fluid discharged from the reactor coolant system or the steam generator to the circulation inducing jet device. The cooling section may include a heat exchanger installed within the containment to allow the cooling water of the emergency cooling water storage section or atmosphere outside the containment to pass therethrough so as to exchange heat with the first fluid and the second fluid jetted from the circulation inducing jet device; and a casing configured to surround the heat exchanger to allow at least part thereof to protect the heat exchanger and accommodate the first fluid and the second fluid jetted from the circulation inducing jet device. The connected line may include a first connected line connected to the emergency cooling water storage section and the heat exchanger to form a flow path for supplying the cooling water of the emergency cooling water storage section to the heat exchanger; a second connected line extended from the heat exchanger to an outside of the containment to discharge the cooling water of the emergency cooling water storage section that has passed through the heat exchanger to an outside thereof; a third connected line branched from the second connected line and extended to an outside of the containment to form a flow path for supplying atmosphere outside the containment to the heat exchanger; and a fourth connected line branched from the first connected line and extended to an outside of the containment to discharge atmosphere heated while passing through the heat exchanger to an outside thereof, wherein the passive safety facility further comprises isolation valves installed at the first connected line through the fourth connected line, respectively, and the isolation valves are open or closed by a preset signal to switch between water-cooled type cooling using the coolant and air-cooled type cooling using the atmosphere when the cooling water of the emergency cooling water storage section is exhausted. The passive safety facility may further include a filter facility connected to an outlet of the cooling section to filter out noncondensible gas discharged from the cooling section so as to collect radioactive materials filtered out from the noncondensible gas, and the cooling section may include a heat exchanger installed outside the containment to connect to a connected line passing through the containment, and allow the first fluid and the second fluid flowing in from the circulation inducing jet device to pass through the connected line to exchange heat with the cooling water of the emergency cooling water storage section or atmosphere outside the containment, and the filter facility may be installed within the containment, and connected to the heat exchanger by the connected line to receive noncondensible gas and condensate from the heat exchanger. The cooling section may include a first heat exchanger installed within the containment to cool the first fluid and second fluid jetted from the circulation inducing jet device; and a second heat exchanger installed outside the containment, and connected to the first heat exchanger by a connected line passing through the containment to form a closed flow path to transfer heat that has transferred to a fluid circulating the closed flow path to cooling water within the emergency cooling water storage section or atmosphere outside the containment. The containment may include a containment vessel formed of steel to surround the reactor coolant system; and a containment building formed of concrete to surround the containment vessel at a position separated from the containment vessel to form an air circulating flow path, and the circulation inducing jet device may be configured to jet the first fluid and the second fluid to an inner wall surface of the containment vessel, and the cooling section may cool the containment vessel using air that circulates through the air circulating flow path and the spraying of a passive containment vessel spray system. According to the present disclosure having the foregoing configuration, it may be possible to promote natural circulation using a circulation inducing jet device to increase an efficiency of cooling an inside of the containment, thereby overcoming the technical limitation of the related art depending on natural circulation within the containment. Furthermore, the present disclosure may induce the second fluid within the containment at the same time to the heat exchanger by the discharge flow of the first fluid, thereby solving the problems of size increase, cost increase and safety degradation in the heat exchanger for cooling the containment in a nuclear power plant. In addition, the present disclosure may reduce a concentration of radioactive materials within the containment at an early stage using a filter facility. As the concentration of radioactive materials is reduced, the present disclosure may decrease the exclusion area boundary. Hereinafter, a passive safety facility and a nuclear power plant including the same associated with the present disclosure will be described in more detail with reference to the accompanying drawings. Even in different embodiments according to the present disclosure, the same or similar reference numerals are designated to the same or similar configurations, and the description thereof will be substituted by the earlier description. Unless clearly used otherwise, expressions in the singular number used in the present disclosure may include a plural meaning. FIG. 1 is a conceptual view illustrating when a normal operation is carried out on a passive safety facility 1100 and a nuclear power plant 110 including the same associated with an embodiment of the present disclosure. The nuclear power plant 110 may include a reactor coolant system 111, a containment 112 and a passive safety facility 1100. A core 111a and a steam generator 111b are provided within the reactor coolant system 111, and a lower portion of the steam generator 111b is connected to a feedwater system 113 by a feedwater line 113a, and an upper portion of the steam generator 111b is connected to a turbine system 114 by a steam line 114a. The containment 112 surrounds the reactor coolant system 111 to prevent radioactive materials from being leaked to an external environment. During the occurrence of an accident, such as a loss of coolant accident or non-loss of coolant accident, there is a concern of leaking radioactive materials from the reactor coolant system 111, and thus the containment 112 is formed to surround the reactor coolant system 111 outside the reactor coolant system 111 to prevent the leakage of radioactive materials. The containment 112 performs the role of a final barrier for preventing the leakage of radioactive materials to an external environment from the reactor coolant system 111. The containment 112 is divided into a containment building (or reactor building) configured with reinforced concrete and a containment vessel and a safeguard vessel configured with a steel vessel according to a material constituting a pressure boundary. The containment vessel is a large vessel designed at a low pressure such as a containment building, and the safeguard vessel is a small vessel designed in a small size at an increased design pressure. According to the present disclosure, unless otherwise noted, the containment 112 commonly refers to a containment building, a reactor building, a containment vessel, a safeguard vessel or the like. The containment 112 illustrated in FIG. 1 is illustrated as a containment building formed with reinforced concrete. Various fluids for maintaining the safety of the nuclear power plant 110 exist within the containment 112. A fluid for cooling the core 111a is filled in the reactor coolant system 111. Furthermore, fluids for making preparations for various accidents are also filled within the containment 112. Hereinafter, it will be described that among fluids within the containment 112, a fluid discharged from the reactor coolant system 111 and a fluid existing in a space between the reactor coolant system 111 and the containment 112 are divided into a first fluid and a second fluid, respectively. However, such a division of fluids is irrelevant to the properties of a fluid or materials constituting a fluid. Accordingly, the first fluid and second fluid may be the same type of fluid. The passive safety facility 1100 is configured to circulate a fluid to the reactor coolant system 111 so as to remove the heat of the reactor coolant system 111 using a primary cooling water circulation method or secondary cooling water circulation method, and cool a first fluid discharged from the reactor coolant system 111 and a second fluid within the containment 112 at the same time to discharge heat within the containment 112 to an external environment. The passive safety facility 1100 is configured to increase a heat and pressure reduction efficiency within the containment 112 and a removal efficiency of radioactive materials in a passive method using a structure formed to accelerate circulation flow by getting out of a conventional method using pure natural convection flow that occurs within the containment 112. The passive safety facility 1100 may include a cooling section 1110 and a circulation inducing jet device 1120. The cooling section 1110 is formed to cool a first fluid discharged from the reactor coolant system 111 along with a second fluid within the containment 112. The cooling section 1110 is configured to discharge heat received from the first fluid and second fluid to an external environment as the first fluid and second fluid are cooled, and return the cooled first fluid and second fluid to a cooling water storage section 1130. The cooling section 1110 may include an emergency cooling water storage section 1111 and a heat exchanger 1112. The emergency cooling water storage section 1111 is formed to store cooling water therein, and receives heat from the first fluid and second fluid, and when the temperature of the cooling water increases, the emergency cooling water storage section 1111 evaporates the cooling water to discharge the received heat to an external environment. At least part of an upper portion of the emergency cooling water storage section 1111 is open to allow cooling water to be evaporated to an external environment. The heat exchanger 1112 is installed within the emergency cooling water storage section 1111, and connected to a connected line 1113 passing through the containment 112 to communicate with an inside of the containment 112. The heat exchanger 1112 allows a fluid flowing in from the containment 112 through the connected line 1113 to pass therethrough so as to exchange heat with cooling water stored within the emergency cooling water storage section 1111. A fluid flowing in from the containment 112 through the connected line 1113 may include a first fluid discharged from the reactor coolant system 111 and a second fluid within the containment 112. According to the characteristics of the nuclear power plant 110, the cooling section 1110 may be configured with an air-cooled type by exposing the heat exchanger 1112 to atmosphere and installing a duct (not shown) without installing the emergency cooling water storage section 1111. An inlet header 1112a for distributing the first fluid and second fluid to an internal flow path of the heat exchanger 1112 is installed at an inlet of the heat exchanger 1112. An outlet header 1112b for collecting heated cooling water from the internal flow path is installed at an outlet of the heat exchanger 1112. The connected line 1113 connects an inside of the containment 112 and the heat exchanger 1112 through the containment 112 and the emergency cooling water storage section 1111. At least one isolation valve 1114 for closing and isolating the isolation valve 1114 when the system is damaged during an accident or switching at the time point when the maintenance is required may be installed at the connected line 1113. The circulation inducing jet device 1120 is formed to jet the first fluid and second fluid to the cooling section 1110. The second fluid within the containment 112 is entrained and jetted along with the first fluid by a pressure drop caused while jetting the first fluid. The detailed structure and operation mechanism of the circulation inducing jet device 1120 will be described with reference to FIG. 2. FIG. 2 is an enlarged conceptual view illustrating the circulation inducing jet device 1120 illustrated in FIG. 1. The circulation inducing jet device 1120 may include a first fluid jetting section 1121, a second fluid entraining section 1122b, and a circulating fluid jetting section 1122. The first fluid jetting section 1121 is connected to the reactor coolant system 111 (refer to FIG. 1) or the steam generator 111b to jet a first fluid provided from the reactor coolant system 111 or the steam generator 111b. The first fluid jetting section 1121 may be connected to the steam generator 111b to receive the first fluid from the steam generator 111b as illustrated in FIG. 1. The first fluid jetting section 1121 may include a nozzle 1121a formed to jet the first fluid. The first fluid discharged through the nozzle 1121a rapidly increases the speed thereof, and decreases the pressure thereof while passing through a throat 1122a with a small flow path area. Accordingly, a pressure drop is locally caused within the circulation inducing jet device 1120. The second fluid entraining section 1122b is formed in an annular shape around the first fluid jetting section 1121 to entrain the second fluid within the containment. A pressure difference is formed between an inside and an outside of the circulation inducing jet device 1120 by a pressure drop caused by the jetting of the first fluid. Since a pressure within the circulation inducing jet device 1120 is lower than that of the outside thereof, the second fluid existing outside the circulation inducing jet device 1120 is entrained into the circulation inducing jet device 1120 through the second fluid entraining section 1122b. The circulating fluid jetting section 1122 as a portion having an inner diameter larger than that of the first fluid jetting section 1121 to form the second fluid entraining section 1122b surrounds the first fluid jetting section 1121. Accordingly, the second fluid entraining section 1122b having an annular shape is formed between an outer circumferential surface of the first fluid jetting section 1121 and an inner circumferential surface of the circulating fluid jetting section 1122. The circulation inducing jet device 1120 supplies the first fluid and second fluid to the cooling section 1110 at the same time. The circulation inducing jet device 1120 may include a throat 1122a and a diffuser 1122c. The throat 1122a is formed with an inner diameter smaller than that of the surroundings to cause a local pressure drop during the jetting of the first fluid. As illustrated in FIG. 2, the throat 1122a has an inner diameter smaller than that of the second fluid entraining section 1122b and diffuser 1122c. The diffuser 1122c naturally induces the first fluid and second fluid to the cooling section 1110 without generating a large pressure loss to the first fluid and second fluid that have passed through the throat 1122a. If the flowing of the first fluid and second fluid that have passed through the throat 1122a is not naturally diffused, a flow path resistance may increase to decrease circulation flow. The diffuser 1122c decreases a flow path resistance by naturally changing a dynamic pressure to a static pressure to efficiently supply the first fluid and second fluid to the cooling section 1110. As illustrated in FIG. 2, the throat 1122a and the diffuser 1122c are sequentially connected to each other. The inner diameter thereof is formed to gradually decrease as moving from the second fluid entraining section 1122b to the throat 1122a, and increase again as moving from the throat 1122a to the diffuser 1122c. The circulating fluid jetting section 1122 is connected to the connected line 1113 passing through the containment 112. The circulating fluid jetting section 1122 jets the first fluid and second fluid to an inside of the connected line 1113. The jetted first fluid and second fluid exchanges heat with cooling water within the emergency cooling water storage section 1111 while passing through the heat exchanger 1112. Due to such a structural feature of the circulation inducing jet device 1120, the present disclosure may overcome the limitation of the related art depending on pure natural convection within the containment 112, and promote the circulation flow of the first fluid and second fluid to enhance cooling efficiency within the containment 112. Referring to FIG. 1 again, the passive safety facility 1100 may include a cooling water storage section 1130, a fluid circulation section 1140, a condensate holding section 1150 and a return line 1160. The cooling water storage section 1130 is formed to store cooling water to be injected into the reactor coolant system 111 therein. The cooling water storage section 1130 is installed at a position higher than that of the reactor coolant system 111 to allow the injection of cooling water due to a gravity water head. According to the present disclosure, the passive safety system 1110 collectively refers to i) a function of a passive residual heat removal system and ii) a latter safety injection function of a passive safety injection system. The cooling water circulation method of the passive safety system 1110 may use a secondary cooling water circulation method using the steam generator 111b and a primary cooling water circulation method for directly injecting cooling water to the reactor coolant system 111. Cooling water stored in the cooling water storage section 1130 may be used for at least one of the residual heat removal of the reactor coolant system 111 and safety injection to the reactor coolant system 111 according to the usage. Since sensible heat and residual heat generated from the core 111a exist within the reactor coolant system 111 during the occurrence of an accident, the sensible heat and residual heat should be removed to safely maintain the core 111a. A method of circulating cooling water to the steam generator 111b to remove sensible heat within the reactor coolant system 111 and residual heat in the core 111a is applied to the embodiment of FIG. 1. The cooling water storage section 1130 may be connected to the feedwater line 113a to use cooling water stored therein for the removal of residual heat. Furthermore, since a water level of the reactor coolant system 111 decreases during the occurrence of an accident, cooling water should be injected into the reactor coolant system 111 to maintain the water level. The cooling water storage section 1130 may be connected to the safety injection line 115a by a line 1115 to use cooling water stored therein for safety injection. The cooling water storage section 1130 may be separately provided with a first cooling water storage section 1130a and a second cooling water storage section 1130b to store pure cooling water to be used for residual heat removal and borated water to be used for safety injection, respectively, in a separate manner as illustrated in FIG. 1. The first cooling water storage section 1130a stores pure cooling water to be supplied to the steam generator 111b to remove sensible heat within the reactor coolant system 111 and residual heat in the core 111a. The second cooling water storage section 1130b stores borated water to be directly injected into the reactor coolant system 111 to maintain the water level of the reactor coolant system 111. The fluid circulation section 1140 is formed to circulate the first fluid from the cooling water storage section 1130 to the circulation inducing jet device 1120 through the reactor coolant system 111 or the steam generator 111b. The fluid circulation section 1140 may include a fluid supply line 1141 and a steam discharge line 1142. The fluid supply line 1141 is connected to the cooling water storage section 1130 to supply cooling water within the cooling water storage section 1130 to the reactor coolant system 111 or steam generator 111b. The fluid supply line 1141 may be directly or indirectly connected to the reactor coolant system 111. The fluid supply line 1141 may be connected to the feedwater line 113a to supply cooling water to the steam generator 111b within the reactor coolant system 111 as illustrated in the drawing. The steam discharge line 1142 is connected to the circulation inducing jet device 1120 to supply the first fluid that has passed through the reactor coolant system 111 or steam generator 111b to the circulation inducing jet device 1120. The steam discharge line 1142 may supply is connected to the steam generator 111b to supply the first fluid discharged from the steam generator 111b to the circulation inducing jet device 1120 as illustrated in the drawing. The condensate holding section 1150 collects condensate formed by cooling the first fluid and second fluid in the cooling section 1110. At least part of an upper portion of the condensate holding section 1150 is open to collect condensate falling from the cooling section 1110 and installed between the cooling section 1110 and the cooling water storage section 1130. The return line 1160 is extended from the condensate holding section 1150 to the cooling water storage section 1130 to return condensate collected in the condensate holding section 1150 again to the cooling water storage section 1130. A flow regulator 1160a for controlling the flow of the condensate may be installed at the return line 1160. When the condensate is collected in the condensate holding section 1150, and returned to the cooling water storage section 1130 through the return line 1160, the circulation of flow started from the cooling water storage section 1130 is completed. The present disclosure induces the circulation of flow in a passive method by natural forces, and thus the circulation of flow does not end by one circulation. The circulation of flow may continue to be sustained while a sufficient passive force capable of inducing the circulation of flow within the containment 112 is maintained by generating steam from the reactor coolant system 111 or steam generator 111b. Hereinafter, the operation of the passive safety facility 1100 and the nuclear power plant 110 including the same during the occurrence of an accident will be described. FIG. 3 is a conceptual view illustrating when an event occurs on the passive safety facility 1100 and the nuclear power plant 110 including the same illustrated in FIG. 1. When an accident such as a loss of coolant accident occurs, isolation valves 113b, 114b installed at the feedwater line 113a and steam line 114a, respectively, are closed. Then, valves 115b installed at the safety injection line 115a are open to implement safety injection from the safety injection facility 115 to the reactor coolant system 111. When an internal pressure of the reactor coolant system 111 decreases as the isolation valve 1115a and check valve 1115b installed at the line 1115 connected to the safety injection line 115a are open, borated water stored in the second cooling water storage section 1130b is also injected into the reactor coolant system 111 by a gravity water head. The borated water stored in the second cooling water storage section 1130b performs the role of maintaining a water level within the reactor coolant system 111 along with the safety injection facility 115. The isolation valve 1141a and check valve 1141b installed at the fluid supply line 1141 for connecting the first cooling water storage section 1130a to the feedwater line 113a are also open to start the supply of cooling water due to a gravity water head. The cooling water of the first cooling water storage section 1130a is supplied to a lower portion of the steam generator 111b through the feedwater line 113a to remove sensible heat within the reactor coolant system 111 and residual heat in the core 111a in the steam generator 111b, and is discharged to an upper portion of the steam generator 111b. The isolation valve 1142a installed at the steam discharge line 1142 is also open, and the first fluid discharged to an upper portion of the steam generator 111b is evaporated through the steam discharge line 1142. The first fluid is supplied to the circulation inducing jet device 1120, and the circulation inducing jet device 1120 jets the first fluid supplied through the steam discharge line 1142 and the second fluid entrained from an inside of the containment 112 by a pressure drop to the connected line 1113. The first fluid and second fluid jetted from the circulation inducing jet device 1120 transfer heat cooling water within the emergency cooling water storage section 1111 while passing through the heat exchanger 1112, and cool and condense. The cooling water within the emergency cooling water storage section 1111 evaporates and discharges heat to an external environment when the temperature increases. The condensate formed by cooling and condensing the first fluid and second fluid in the heat exchanger 1112 is injected into the containment 112 again through the connected line 1113, and collected into the condensate holding section 1150 installed at a lower portion of the connected line 1113. Noncondensible gas discharged along with condensate from the connected line 1113 is discharged into the containment 112. The condensate collected in the condensate holding section 1150 is returned to the cooling water storage section 1130 again through the return line 1160, and the circulation of flow is carried out in a continuous and consistent manner. However, when it is configured to directly collect condensate into the cooling water storage section 1130 according to the characteristics of the nuclear power plant 110, the condensate holding section 1150 may not be separately provided therein. Hereinafter, another embodiment of a passive safety facility and a nuclear power plant including the same will be described. FIG. 4 is a conceptual view illustrating when a normal operation is carried out on a passive safety facility 1200 and a nuclear power plant 120 including the same associated with another embodiment of the present disclosure. The passive safety facility 1200 may include a cooling section 1210 and a circulation inducing jet device 1220. The passive safety facility 1200 may include a cooling section 1210 and a circulation inducing jet device 1220. An emergency cooling water storage section 1211 is installed outside the containment 122, and cooling water is stored therein. A heat exchanger 1212 is installed within the containment 122 other than an inside of the emergency cooling water storage section 1211, and connected to the emergency cooling water storage section 1211 by a connected line 1213 passing through the containment 122. A sparger 1213′ for jetting cooling water may be installed at an end portion of the connected line 1213. The heat exchanger 1212 is configured to allow cooling water within the emergency cooling water storage section 1211 to pass therethrough so as to exchange heat with the first fluid and second fluid jetted from the circulation inducing jet device 1220. However, a duct (not shown) may be installed to configure the nuclear power plant 120 in an air-cooled type without installing the emergency cooling water storage section 1211 according to the characteristics of the nuclear power plant 120. The structure and operation of the circulation inducing jet device 1220 illustrated in FIG. 4 will be described with reference to FIGS. 5A and 5B. FIGS. 5A and 5B are enlarged conceptual views illustrating the circulation inducing jet device 1220 illustrated in FIG. 4. FIG. 5A is a conceptual view in which the circulation inducing jet device 1220 is seen from a front side, and FIG. 5B is a conceptual view in which the circulation inducing jet device 1220 is seen from a lateral side. A circulating fluid jetting section 1222 jets the first fluid and second fluid to the heat exchanger 1212. The jetted first fluid and second fluid exchange heat with cooling water in the heat exchanger 1212. An inlet header 1212a for distributing cooling water supplied from the emergency cooling water storage section 1211 (refer to FIG. 4) to an internal flow path of the heat exchanger 1212 is installed at an inlet of the heat exchanger 1212. An outlet header 1212b for collecting heated cooling water from the internal flow path is installed at an outlet of the heat exchanger 1212. A tube 1212c is installed between the inlet header 1212a and the outlet header 1212b. A casing 1215 for protecting the tube 1212c from missiles (fragments) during an accident is installed on a circumference of the tube 1212c. The circulation inducing jet device 1220 is formed to jet the first fluid and second fluid to a surface of the heat exchanger 1212. However, when a shell-and-tube type heat exchanger is employed according to the design characteristics of the nuclear power plant 120, shell and tube side flow paths may be configured in an opposite manner. The first fluid and second fluid jetted from the circulation inducing jet device 1220 are cooled and condensed by exchanging heat with the cooling water of the emergency cooling water storage section 1211 while passing through the internal flow path of the heat exchanger 1212. Referring to FIG. 4 again, the passive safety facility 1200 may include a cooling water storage section 1230′, 1230″. The cooling water storage section 1230′, 1230″ may be formed with a tank or cistern in which a first cooling water storage section and a second cooling water storage section are integrally formed without including the first cooling water storage section and the second cooling water storage section in a separate manner. A plurality of cooling water storage sections 1230′, 1230″ may be provided therein, and any part of the cooling water storage section 1230′ may be connected to the feedwater line 123a, and another part of the cooling water storage section 1230″ may be connected to the safety injection line 125a. Hereinafter, the operation of the passive safety facility 1200 and the nuclear power plant 120 including the same as illustrated in FIG. 4 during the occurrence of an accident will be described. FIG. 6 is a conceptual view illustrating when an event occurs on the passive safety facility 1200 and the nuclear power plant 120 including the same illustrated in FIG. 4. When an accident occurs on the nuclear power plant 120, isolation valves 123b, 124b installed at the feedwater line 123a and steam line 124a, respectively, are closed. Then, an isolation valve 1241a and a check valve 1241b installed at a fluid supply line 1241 are open. An isolation valve 1215a and a check valve 1251b installed at a line 1215 for connecting between a cooling water storage section 1230″ and a safety injection line 125a are also open. The cooling water of the cooling water storage section 1230″ is safely injected into a reactor coolant system 121 along with the cooling water of the safety injection facility 125. The cooling water of the cooling water storage section 1230′ is supplied to a steam generator 121b through the fluid supply line 1241 to remove sensible heat within the reactor coolant system 121 and residual heat in the core 121a. The first fluid discharged from the steam generator 121b is evaporated through a steam discharge line 1242 in which an isolation valve 1242a is open, and supplied to the circulation inducing jet device 1220. The first fluid is jetted to the heat exchanger 1212 by the circulation inducing jet device 1220. The second fluid is also entrained into the circulation inducing jet device 1220 and jetted to the heat exchanger 1212 along with the first fluid. The first fluid and second fluid exchange heat with cooling water supplied from the emergency cooling water storage section 1211 on a surface of the heat exchanger 1212, and cool and condense. Then, condensate formed by condensing the first fluid and second fluid falls by gravity. The falling condensate is collected into a condensate holding section 1250, and returned to the cooling water storage section 1230 through a return line 1260, and the circulation of flow is consistently carried out. Hereinafter, still another embodiment of a passive safety facility and a nuclear power plant including the same will be described. FIG. 7 is a conceptual view illustrating when a normal operation is carried out on a passive safety facility 1300 and a nuclear power plant 130 including the same associated with still another embodiment of the present disclosure. The passive safety facility 1300 may include a cooling section 1310 and a circulation inducing jet device 1320. The passive safety facility 1300 may include an emergency cooling water storage section 1311, a first heat exchanger 1312′ and a second heat exchanger 1312″. The description of the emergency cooling water storage section 1311 will be substituted by the earlier description of FIGS. 1 and 4. The heat exchanger 1312′ is installed within a containment 132 to exchange heat a fluid within the containment 132. The second heat exchanger 1312″ is installed within the emergency cooling water storage section 1311, and connected to the first heat exchanger 1312′ by a connected line 1313 to form a closed flow path with the second heat exchanger 1312″. A fluid circulates within the closed flow path independently from cooling water within the emergency cooling water storage section 1311 or a fluid within the containment 132. The second heat exchanger 1312″ transfers heat transferred to a fluid that circulates the closed flow path in the first heat exchanger 1312′ to cooling water within the emergency cooling water storage section 1311. The heat transferred to the emergency cooling water storage section 1311 is discharged to an external environment by the evaporation of cooling water. The connected line 1313 is connected to the first heat exchanger 1312′ and the second heat exchanger 1312″, respectively, through the containment 132 and emergency cooling water storage section 1311. A makeup tank 1316 is formed to store a makeup fluid therein, and connected to the connected line 1313 to make up the makeup fluid to the closed flow path. A first cooling water storage section 1330a may be used for the purpose of removing sensible heat within a reactor coolant system 131 and residual heat in a core 131a, and a safety injection facility (not shown) may be separately provided from the first cooling water storage section 1330a. FIG. 8 is a conceptual view illustrating when an event occurs on a passive safety facility 1300 and a nuclear power plant 130 including the same illustrated in FIG. 7. When an accident occurs on the nuclear power plant 130, isolation valves 133b, 134b installed at the feedwater line 133a and steam line 134a, respectively, are closed. Furthermore, an isolation valve 1341a and a check valve 1341b installed at the fluid supply line 1341 are open to supply cooling water to a steam generator 131b. Cooling water supplied to the steam generator 131b is evaporated by receiving sensible heat and residual heat. The first fluid discharged from the steam generator 131b is supplied to the circulation inducing jet device 1320 through the steam discharge line 1342. The circulation inducing jet device 1320 jets the first fluid supplied through the steam discharge line 1342 and the second fluid entrained from the containment 132 to a surface of the first heat exchanger 1312′. The first fluid and second fluid jetted to a surface of the first heat exchanger 1312′ exchange heat with cooling water flowing through an inside of the closed flow path formed by the first heat exchanger 1312′, the second heat exchanger 1312″ and the connected line 1313 to cool and condense, and falls as condensate. The falling condensate is collected into a condensate holding section 1350. A fluid flowing through an inside of the closed flow path receives heat from an inside of the containment 132 while persistently circulating the closed flow path, and transfers the received heat to cooling water within an emergency cooling water storage section 1311. When a fluid within the closed flow path is insufficient, a makeup fluid is made up from the makeup tank 1316 to continue the circulation. Cooling water within the emergency cooling water storage section 1311 increases the temperature as receiving heat, and evaporates to discharge heat to an external environment. Hereinafter, a passive safety facility 1400 and a nuclear power plant 140 including the same according to yet still another embodiment of the present disclosure will be described. FIG. 9 is a conceptual view illustrating when a normal operation is carried out on a passive safety facility 1400 and a nuclear power plant 140 including the same associated with yet still another embodiment of the present disclosure. The nuclear power plant 140 may include a reactor coolant system 141, a containment 142, a passive containment vessel spray system 146, and a passive safety facility 1400. The containment 142 may include a containment vessel 142a and a containment building 142b contrary to the foregoing containment. The containment vessel 142a formed of steel, and formed to surround the reactor coolant system 141. The containment building 142b is formed of concrete, and formed to surround the containment vessel 142a at a position separated from the containment vessel 142a so as to form an air circulation flow path 142c between the containment vessel 142a and the containment building 142b. The containment building 142b may include at least one air inlet 142b′ to flow external air for cooling the containment vessel 142a thereinto while circulating the air circulation flow path 142c. The passive containment vessel spray system 146 may include a spray cooling water storage section 146a, a spray line 146b, a spray isolation valve 146c and a spray nozzle 146d. The spray cooling water storage section 146a is formed to store cooling water, and installed at an upper portion of the containment building 142b. The spray line 146b may form a flow path to flow the cooling water of the spray cooling water storage section 146a, and the spray isolation valve 146c may be installed at the spray line 146b. Furthermore, the spray nozzle 146d may be installed at an end portion of the spray line 146b. The passive containment vessel spray system 146 sprays cooling water to an outer surface of the containment vessel 142a to cool the containment vessel 142a. The passive safety facility 1400 may include a cooling water storage section 1430 and a circulation inducing jet device 1420. The cooling water storage section 1430 may include a first cooling water storage section 1430a and a second cooling water storage section 1430b. The first cooling water storage section 1430a is connected to a feedwater line 143a to inject cooling water into a steam generator 141b. The first fluid discharged from the steam generator 141b is supplied to the circulation inducing jet device 1420 through a steam discharge line 1442. The second cooling water storage section 1430b is connected to a safety injection line 145a to inject borated water into the reactor coolant system 141. Cooling water safely injected from the second cooling water storage section 1430b circulates the reactor coolant system 141, and the first fluid discharged from the reactor coolant system 141 is supplied to the circulation inducing jet device 1420 through the steam discharge line 1442. The circulation inducing jet device 1420 will be described with reference to FIGS. 10A and 10B. FIGS. 10A and 10B are enlarged conceptual views illustrating the circulation inducing jet device 1420 illustrated in FIG. 9. FIG. 10A is a conceptual view in which the circulation inducing jet device 1420 is seen from a front side, and FIG. 10B is a conceptual view in which the circulation inducing jet device 1420 is seen from a lateral side. The circulation inducing jet device 1420 is formed to jet the first fluid and the second fluid to an inner wall surface of the containment vessel 142a. An outlet of the circulation inducing jet device 1420 is installed toward an inner wall surface of the containment vessel 142a. The first fluid supplied from the steam discharge line 1442 (refer to FIG. 9) is jetted from a first fluid jetting section 1421. The second fluid is also entrained into the circulation inducing jet device 1420. The first fluid and second fluid is jetted to an inner wall surface of the containment vessel 142a. The first fluid and second fluid jetted to the containment vessel 142a are cooled and condensed on the inner wall surface of the containment vessel 142a. Referring to FIG. 9 again, the cooling section is not installed as an additional device on the nuclear power plant 140, but the containment vessel 142a functions as the cooling section. The first fluid and second fluid jetted from the circulation inducing jet device 1120 transfer heat to the containment vessel 142a. Air that circulates the air circulation flow path 142c through the air inlet 142b′ and cooling water sprayed from the passive containment vessel spray system 146 consistently cool the containment vessel 142a. Heat is discharged to an external environment by air that circulates the air circulation flow path 142c and cooling water sprayed on an outer surface of the containment vessel 142a. A condensate holding section 1450 is installed at a lower portion of an outlet of the circulation inducing jet device to collect condensate condensed on an inner wall surface of the containment vessel 142a to fall. On the contrary, the cooling water storage section 1430 may be also installed at a lower portion of an outlet of the circulation inducing jet device 1420 on the nuclear power plant 140 without installing the condensate holding section 1450 in a separate manner to collect condensate condensed on the inner wall surface of the containment vessel 142a to fall. Hereinafter, the operation of the passive safety facility 1400 and the nuclear power plant 140 including the same as illustrated in FIG. 9 during the occurrence of an accident will be described. FIG. 11 is a conceptual view illustrating when an event occurs on the passive safety facility 1400 and the nuclear power plant 140 including the same illustrated in FIG. 9. When an accident occurs on the nuclear power plant 140, isolation valves 143b, 144b installed at the feedwater line 143a and steam line 144a, respectively, are closed. An isolation valve 145b installed at a safety injection line 145a for connecting between the safety injection facility 145 and the reactor coolant system 141 are open. An isolation valve 1415a and a check valve 1451b installed at a line 1415 for connecting between a safety injection line 145a and a second cooling water storage section 1430b are also open. When a pressure within the reactor coolant system 141 decreases, safety injection from the safety injection facility 145 or the second cooling water storage section 1430b into the reactor coolant system 141 is carried out by a gravity water head. The first fluid that has received heat within the reactor coolant system 141 is evaporated through the steam discharge line 1442, and jetted to an inner wall surface of the containment vessel 142a along with the second fluid through the circulation inducing jet device 1420. Accordingly, heat is transferred from the first fluid and second fluid to the containment vessel 142a. External air flows on an outer surface of the containment vessel 142a through the air inlet 142′ to cool the containment vessel 142a while circulating the air circulation flow path 142c. The air that has received heat from the outer surface of the containment vessel 142a is ascended and discharged to an outside through an opening portion at an upper portion of the containment building 142b. Furthermore, the passive containment vessel spray system 146 sprays cooling water to an outer surface of the containment vessel 142a to cool the containment vessel 142a as the spray isolation valve 146c is open. The containment vessel 142a is cooled by air circulation and spraying. As the circulation inducing jet device 1420 sprays the first fluid and second fluid to the containment vessel 142a, heat transferred to the containment vessel 142a may be discharged to an external environment by air circulation and spraying. Condensate formed by cooling and condensing the first fluid and second fluid on an inner wall surface of the containment vessel 142a falls, and returns to the cooling water storage section 1430 through the condensate holding section 1450. According to the foregoing embodiments, noncondensible gas is discharged to an inside of the containment. Hereinafter, a passive safety facility including a filter facility for filtering out noncondensible gas and a nuclear power plant including the same will be described. FIG. 12 is a conceptual view illustrating a passive safely system 2100 and a nuclear power plant 210 including the same associated with still yet another embodiment of the present disclosure. The nuclear power plant 210 may include a containment 212, a reactor coolant system 211, a core 211a, a steam generator 211b, a reactor coolant pump 211c and a pressurizer 211d. The nuclear power plant 210 may include systems for the normal operation of the nuclear power plant 210 and various systems for securing the safety of the nuclear power plant 210 in addition to the constituent elements illustrated in FIG. 12. The reactor coolant system 211 is installed within the containment 212. The reactor coolant system 211 is a coolant system for transferring and transporting thermal energy generated by the fission of the core 211a. The first fluid is filled into the reactor coolant system 211. During the occurrence of an accident such as a loss of coolant accident, steam may be discharged from the reactor coolant system 211, and the containment 212 blocks radioactive materials contained in the steam from being leaked to an outside thereof. The steam generator 211b generates steam using heat transferred from the core. A lower inlet of the steam generator 211b is connected to a feedwater system 213 by a feedwater line 213a, and an upper outlet of the steam generator 211b is connected to a turbine system 214 by a steam line 214a. Feedwater supplied to the steam generator 211b through the feedwater line 213a evaporates in the steam generator 211b to become steam. The steam is supplied to the turbine system 214 through the steam line 214a. The reactor coolant pump 211c induces the circulation of the first fluid, and the pressurizer 211d maintains a pressurized state that exceeds a saturation pressure to prevent the boiling of coolant in the core 211a of a pressurized water reactor. The containment 212 surrounds the reactor coolant system 211 to prevent radioactive materials from being leaked to an external environment. During the occurrence of an accident, such as a loss of coolant accident or non-loss of coolant accident, there is a concern of leaking radioactive materials from the reactor coolant system 211, and thus the containment 212 is formed to surround the reactor coolant system 211 at an outside of the reactor coolant system 211 to prevent the leakage of radioactive materials. Various fluids for maintaining the safety of the nuclear power plant 210 exist within the containment 212. A fluid for cooling the core 211a is filled in the reactor coolant system 211. Furthermore, fluids for making preparations for various accidents are also filled within the containment 212. Hereinafter, it will be described that among fluids within the containment 212, a fluid discharged from the reactor coolant system 211 and a fluid existing in a space between the reactor coolant system 211 and the containment 212 are divided into a first fluid and a second fluid, respectively. However, such a division of fluids is irrelevant to the properties of a fluid or materials constituting a fluid. Accordingly, the first fluid and second fluid may be the same type of fluid. Furthermore, the first fluid and second fluid should be distinguished from a primary fluid and a secondary fluid. The primary fluid and the secondary fluid may be a first fluid or second fluid. Referring to FIG. 12, a method of circulating the second fluid using the steam generator 211b is applied to the passive safely system 2100. Accordingly, both the first fluid and second fluid in the passive safely system 2100 illustrated in FIG. 12 indicate a secondary fluid. If it is a passive safety facility with a method of circulating the first fluid, then both the first fluid and second fluid indicate a primary fluid. The passive safely system 2100 removes the sensible heat of the reactor coolant system 211 and the residual heat of the core 211a using a circulation method of the primary fluid or a circulation method of the secondary fluid. In case of using the circulation of the primary fluid, the passive safely system 2100 circulates the primary fluid to the reactor coolant system 211. In case of using the circulation of the secondary fluid, the passive safely system 2100 circulates the secondary fluid to the steam generator 211b. The passive safely system 2100 is configured to cool the first fluid discharged from the reactor coolant system 211 or steam generator 211b and the second fluid within the containment 212 at the same time to discharge heat within the containment 212 to an external environment. FIG. 12 illustrates the passive safely system 2100 using the circulation method of the secondary fluid. The passive safety facility 2100 uses a facility formed to accelerate circulation flow by getting out of a conventional method using pure natural convection flow. The passive safely system 2100 is configured to increase a heat and pressure reduction efficiency within the containment 212 and a removal efficiency of radioactive materials in a passive method. Referring to FIG. 12, the passive safely system 2100 may include a cooling section 2110, a circulation inducing jet device 2120 and a filter facility 2170. The cooling section 2110 is formed to cool the first fluid discharged from the steam generator 211b along with the second fluid within the containment 212. The cooling section 2110 is configured to discharge heat received from the first fluid and second fluid to an external environment of the containment 212 as the first fluid and second fluid are cooled. The cooling section 2110 may include an emergency cooling water storage section 2111, a heat exchanger 2112, a connected line 2113 and an isolation valve 2114. The emergency cooling water storage section 2111 is formed to store cooling water therein. The cooling water filled in the emergency cooling water storage section 2111 receives heat from the first fluid and second fluid by the heat exchanger 2112. When the temperature of the cooling water filled in the emergency cooling water storage section 2111 increases, the cooling water evaporates to discharge heat that has transferred to the cooling water to an external environment. At least part of an upper portion of the emergency cooling water storage section 2111 is open to allow cooling water to be evaporated to an external environment. The heat exchanger 2112 is configured such that the cooling section of the emergency cooling water storage section 2111 exchanges heat with the first fluid and second fluid. The heat exchanger 2112 may be installed within the containment 212, and connected to the emergency cooling water storage section 2111 by the connected line 2113 passing through the containment 212. The heat exchanger 2112 allows cooling water flowing in from the emergency cooling water storage section 2111 through the connected line 2113 to pass therethrough to cool the first fluid and second fluid. An inlet header 2112a for distributing cooling water supplied from the emergency cooling water storage section 2111 to an internal flow path of the heat exchanger 2112 is installed at an inlet of the heat exchanger 2112. An outlet header 2112b for collecting heated cooling water from the internal flow path of the heat exchanger 2112 is installed at an outlet of the heat exchanger 2112. The connected line 2113 is connected to the heat exchanger 2112 and emergency cooling water storage section 2111 to form a circulating flow path of cooling water stored in the emergency cooling water storage section 2111. A plurality of connected lines 2113 are provided therein and connected to the inlet header 2112a and outlet header 2112b, respectively, of the heat exchanger 2112. The connected line 2113 passes through at least part of the containment 212, and extended to an inside of the emergency cooling water storage section 2111. A sparger 2113′ for jetting cooling water may be installed at an end portion of the connected line 2113. Cooling water returned from the heat exchanger 2112 to the emergency cooling water storage section 2111 by the connected line 2113 may be jetted to an inside of the emergency cooling water storage section 2111 by the sparger 2113′. The isolation valve 2114 may be installed at each connected line 2113. The isolation valve 2114 may be closed and isolated when the system is damaged during an accident or switched for maintenance at the time point when the maintenance is required. The cooling section 2110 may be divided according to the operation mechanism of the emergency cooling water storage section 2111 and heat exchanger 2112. As illustrated in FIG. 12, the cooling section 2110 in such a type that the heat exchanger 2112 is connected to the emergency cooling water storage section 2111 by the connected line 2113, and the cooling water of the emergency cooling water storage section 2111 consistently circulates the heat exchanger 2112 may be divided into a circulation type. The cooling section 2110 with a circulation type uses natural circulation based on a density difference due to a difference between cooling water temperatures or phases. The cooling section 2110 may be also divided into an immersion type or injection type in addition to the circulation type, and the configuration of the immersion type and injection type will be described later. The circulation inducing jet device 2120 is formed to jet the first fluid discharged from the reactor coolant system 211 or steam generator 211b to the cooling section 2110. When the first fluid is jetted, a pressure drop is locally caused in the circulation inducing jet device 2120. At least part of the circulation inducing jet device 2120 is open toward an inside of the containment 212 to entrain the second fluid by a pressure drop caused while jetting the first fluid. The circulation inducing jet device 2120 jets the entrained second fluid along with the first fluid to the cooling section 2110. The detailed structure and operation mechanism of the circulation inducing jet device 2120 will be substituted by the earlier description. The circulation inducing jet device 2120 jets the first fluid and second fluid to the heat exchanger 2112. The jetted first fluid and second fluid exchange heat with cooling water in the heat exchanger 2112. An inlet header 2112a for distributing cooling water supplied from the emergency cooling water storage section 2111 to an internal flow path of the heat exchanger 2112 is installed at an inlet of the heat exchanger 2112. An outlet header 2112b for collecting heated cooling water from the internal flow path is installed at an outlet of the heat exchanger 2112. A casing 2115 for protecting the heat exchanger 2112 from missiles (fragments) during an accident is installed on a circumference of the heat exchanger 2112. The circulation inducing jet device 2120 is formed to jet the first fluid and second fluid to a surface of the heat exchanger 2112. However, when a shell-and-tube type heat exchanger is employed according to the design characteristics of the nuclear power plant 210, shell and tube side flow paths may be configured in an opposite manner. The first fluid and second fluid jetted from the circulation inducing jet device 2120 are cooled and condensed on a surface of the heat exchanger 2112 by exchanging heat with the cooling water passing through the internal flow path of the heat exchanger 2112. Due to such a structural feature of the circulation inducing jet device 2120, the present disclosure may overcome the limitation of the related art depending on pure natural convection within the containment 212, and promote the circulation flow of the first fluid and second fluid to enhance cooling efficiency within the containment 212. The casing 2115 surrounds the heat exchanger 2112 to protect the heat exchanger 2112. Furthermore, the casing 2115 is configured to accommodate the first fluid and second fluid jetted from the circulation inducing jet device 2120. An upper portion 2115a of the casing 2115 is connected to the circulation inducing jet device 2120. A lower portion 2115b of the casing 2115 is sealed excluding a gas line 2173 and a return line 2160. An intermediate portion 2115c connecting between the upper portion 2115a and the lower portion 2115b surrounds the heat exchanger 2112. Condensate formed by cooling the first fluid and second fluid may be collected into the lower portion 2115b of the casing 2115. The filter facility 2170 is connected to an outlet of the cooling section 2110 to filter out noncondensible gas discharged from the cooling section 2110. In the passive safely system 2100 illustrated in FIG. 12, the outlet of the cooling section 2110 refers to the lower portion 2115b of the casing 2115. The filter facility 2170 collects radioactive materials filtered out from the noncondensible gas. The filter facility 2170 may include a filter or absorbent 2171, and a gas discharge section 2172 and a gas line 2173. The filter or absorbent 2171 is configured to separate the radioactive materials from the noncondensible gas. The filter may use a high efficiency particulate air filter (2HEPA filter). Radioactive materials in a gas phase contained in the noncondensible gas are removed while passing through the filter. For example, when the radioactive material is iodine, the iodine is combined with silver nitrate (2silver nitrate) and converted to iodic silver while passing through the filter. Iodic silver is a form that is separable from noncondensible gas. The filter is configured to form iodic silver by reacting silver nitrate with iodine contained in noncondensible gas. Furthermore, the filter is formed to remove iodic silver from the fluid. The absorbent may use charcoal. Iodine organic compounds are combined with materials impregnated into charcoal and converted to a form of quaternary ammonium salt, and adsorbed into the charcoal. Iodine in a molecular form is combined with charcoal through chemical absorption. Charcoal is used as an absorbent material since it has a large internal contact area due to its porous structure. Accordingly, the absorbent is formed to remove iodine contained in noncondensible gas through chemical absorption that is carried out by charcoal. However, the foregoing filter and absorbent are merely an example, and the types of the filter and absorbent may not be necessarily limited to them. The gas discharge section 2172 is configured to discharge noncondensible gas filtered out while passing through the filter or absorbent 2171 to an inside of the containment 212. Radioactive materials are mostly collected by the filter or absorbent 2171, and thus there hardly exist radioactive materials in noncondensible gas discharged from the gas discharge section 2172. The gas line 2173 is connected to the outlet of the cooling section 2110 to supply the noncondensible gas to the filter or absorbent 2171. As illustrated in FIG. 12, the gas line may be connected to the lower portion 2115b of the casing 2115. The passive safely system 2100 may further include a cooling water storage section 2130, a fluid circulation section 2140, a return line 2160 and an additive injection section 2180. The cooling water storage section 2130 is formed to store cooling water to be injected into the reactor coolant system 211 or steam generator 211b therein. The cooling water storage section may be installed below cooling water to collect condensate collected into the lower portion 2115b of the casing 2115. The cooling water storage section 2130 may be installed at a position higher than that of the reactor coolant system 211 or steam generator 211b to allow the injection of cooling water due to a gravity water head. The cooling water stored in the cooling water storage section 2130 may be used for the purpose of removing sensible heat within the reactor coolant system 211 and residual heat in the core 211a according to the design. Furthermore, the cooling water stored in the cooling water storage section 2130 may be used for the purpose of being injected into the reactor coolant system 211. Since sensible heat and residual heat generated from the core 111a exist within the reactor coolant system 211 during the occurrence of an accident, the sensible heat and residual heat should be removed to safely maintain the core 211a. A method of circulating cooling water to the steam generator 211b to remove sensible heat and residual heat is applied to the present embodiment. The cooling water storage section 2130 may be connected to the feedwater line 213a to use cooling water stored therein for the removal of residual heat. The cooling water storage section 2130 illustrated in the right side of the FIG. 12 will be referred to such a structure. Furthermore, since a water level of the reactor coolant system 211 decreases during the occurrence of an accident such as a loss of coolant accident, cooling water should be injected into the reactor coolant system 211 to maintain the water level. The cooling water storage section 2130 may be connected to the safety injection line 215a by a line 2115 to use cooling water stored therein for safety injection. The cooling water storage section 2130 illustrated in the left side of the FIG. 12 will be referred to such a structure. A safety injection facility 215, which is another safety system of the nuclear power plant 210, injects cooling water to the reactor coolant system 211 to maintain a water level of the reactor coolant system 211. The safety injection facility 215 may include various tanks (not shown) for storing safety injection water, a safety injection line 215a, a valve 215b, and the like. The safety injection line 215a connects the tanks to the reactor coolant system 211, and the valve 215b may be installed at the safety injection line 215a. The cooling water storage section 2130 may be connected to the safety injection line 215a for safety injection. The cooling water storage section 2130 may include a first cooling water storage section 2130a and a second cooling water storage section 2130b to store pure cooling water to be used for residual heat removal and borated water to be used for safety injection, respectively, in a separate manner as illustrated in FIG. 12. The first cooling water storage section 2130a stores pure cooling water to be supplied to the steam generator 211b to remove sensible heat within the reactor coolant system 211 and residual heat in the core 211a. The second cooling water storage section 2130b stores borated water to be directly injected into the reactor coolant system 211 to maintain the water level of the reactor coolant system 211. The fluid circulation section 2140 is formed to circulate the cooling water of the cooling water storage section 2130 to the circulation inducing jet device 2120 through the reactor coolant system 211 or the steam generator 211b. The fluid circulation section 2140 may include a fluid supply line 2141 and a steam discharge line 2142. The fluid supply line 2141 is connected to the cooling water storage section 2130 to supply cooling water within the cooling water storage section 2130 to the reactor coolant system 211 or steam generator 211b. The fluid supply line 2141 may be directly or indirectly connected to the reactor coolant system 211. For example, the fluid supply line 2141 may be connected to the feedwater line 213a to supply cooling water to the steam generator 211b within the reactor coolant system 111 as illustrated in FIG. 12. An isolation valve 2141a and a check valve 2141b may be installed at the fluid supply line 2141. The steam discharge line 2142 is connected to the reactor coolant system 211 or steam generator 211b and circulation inducing jet device 2120 to supply the first fluid discharged from the reactor coolant system 211 or steam generator 211b to the circulation inducing jet device 2120. The steam discharge line 2142 is connected to the steam generator 211b to supply the first fluid discharged from the steam generator 211b to the circulation inducing jet device 2120 as illustrated in the drawing. An isolation valve 2142a may be installed at the steam discharge line 2142. The return line 2160 is extended from the casing 2115 to the cooling water storage section to supply condensate collected in the lower portion 2115b of the casing 2115 to the cooling water storage section 2130. A flow regulator 2160a for controlling the flow of the condensate may be installed at the return line 2160. When the condensate is returned to the cooling water storage section 2130 through the return line 2160, one circulation of cooling water started from the cooling water storage section 2130 is completed. The circulation of flow is induced in a passive method by natural forces, and thus the circulation of flow does not end by one circulation. The circulation of flow may continue to be sustained while a sufficient passive force capable of inducing the circulation of flow within the containment 212 is maintained by generating steam from the reactor coolant system 211 or steam generator 211b. An additive injection section 2180 injects an additive into condensate for suppressing the revolatilization of condensate collected in the cooling water storage section 2130. The additive is formed to maintain a pH of the condensate above a preset value. Radioactive iodine dissolved in cooling water exists in the form of negative ions, and when a pH of cooling water dissolved therein is low, the revolatilization amount of radioactive iodine may greatly increase. The reason is because an amount of radioactive iodine being converted to the form of volatile elemental iodine (2I2) greatly increases in cooling water below pH 7. In addition, the amount of being converted to elemental iodine is also related to a temperature of soluble cooling water, a concentration of iodine in the solution, and the like. The converted elemental iodine may be revolatilized into atmosphere according to a separation factor defined as a ratio of a concentration of iodine in cooling water to a concentration of iodine in atmosphere. According to the related regulatory requirements, when a pH of soluble cooling water is above 7.0, the amount of being converted to elemental iodine is sharply reduced to ignore revolatilization. Trisodium phosphate may be used for the additive, for example. Trisodium phosphate controls the pH of cooling water to prevent corrosion within the containment 212 and the revolatilization of radioactive nuclides during an accident. However, according to the present disclosure, the type of the additive may not be necessarily limited to this. Boric acid for suppressing the reactivity of the core 211a and other additives for suppressing the corrosion of the device or the like may be added to the additive to manage the water quality of the cooling water storage section 2130 in a passive manner. The cooling water storage section 2130 is configured to flow the condensate collected in the first cooling water storage section 2130a into the second cooling water storage section 2130b when a level of condensate collected in the first cooling water storage section 2130a exceeds a reference level. For example, when the level of the first cooling water storage section 2130a gradually increases and exceeds the reference level by the collection of the condensate, the condensate collected in the first cooling water storage section 2130a may overflow and flow into the second cooling water storage section 2130b. The additive injection section 2180 may be installed at a flow path connected from the first cooling water storage section 2130a to the second cooling water storage section 2130b to inject the additive to condensate flowing into the second cooling water storage section 2130b. Accordingly, the additive injection section 2180 may inject an additive into condensate flowing into the second cooling water storage section 2130b. Hereinafter, the operation of the passive safety facility 2100 and the nuclear power plant 210 including the same during the occurrence of a virtual event will be described with reference to FIG. 13. FIG. 13 is a conceptual view illustrating when an event occurs on the passive safety facility 2100 and the nuclear power plant 210 including the same illustrated in FIG. 12. When an accident such as a loss of coolant accident occurs, isolation valves 213b, 214b installed at the feedwater line 213a and steam line 214a, respectively, are closed, and valves 215b installed at the safety injection line 215a are open. Then, safety injection is implemented by the safety injection facility 215 to the reactor coolant system 211. An isolation valve 2115a and a check valve 2115b are installed at a line 2115 connecting between the second cooling water storage section 2130b and the safety injection line 215a, and the isolation valve 2115a and check valve 2115b are also open when an event occurs. Accordingly, when an internal pressure of the reactor coolant system 211 decreases, borated water stored in the second cooling water storage section 2130b is also injected into the reactor coolant system 211 by a gravity water head. The borated water stored in the second cooling water storage section 2130b performs the role of maintaining a water level within the reactor coolant system 211 along with the safety injection facility 215. The isolation valve 2141a and check valve 2141b installed at the fluid supply line 2141 are also open to start the supply of cooling water due to a gravity water head from the first cooling water storage section 2130a. The cooling water is supplied to a lower inlet of the steam generator 211b through the feedwater line 213a. The cooling water removes sensible heat within the reactor coolant system 211 and residual heat in the core 211a in the steam generator 211b, and is discharged to an upper outlet of the steam generator 211b. The isolation valve 2142a installed at the steam discharge line 2142 is also open, and the first fluid discharged to an upper portion of the steam generator 211b is evaporated through the steam discharge line 2142. The first fluid is supplied to the circulation inducing jet device 2120, and the circulation inducing jet device 2120 jets the first fluid supplied through the steam discharge line 2142 and the second fluid entrained from an inside of the containment 212 by a pressure drop to an inside of the casing 2115. The first fluid and second fluid jetted from the circulation inducing jet device 2120 exchange heat with the cooling water of the emergency cooling water storage section 2111 in the heat exchanger 2112. The cooling water of the emergency cooling water storage section 2111 is supplied to the heat exchanger 2112 through the connected line 2113, and exchanges heat with the first fluid and second fluid while passing through an internal flow path of the heat exchanger 2112. The cooling water of the emergency cooling water storage section 2111 continuously circulates the emergency cooling water storage section 2111 and heat exchanger 2112 through the connected line 2113. Heat is transferred to cooling water from the first fluid and second fluid. The first fluid and second fluid are cooled and condensed, and the cooling water is heated. The cooling water within the emergency cooling water storage section 2111 evaporates when the temperature increases. Accordingly, heat is discharged to an external environment. The condensate formed by cooling and condensing the first fluid and second fluid in the heat exchanger 2112 is collected into a lower portion 2115b of the casing 2115. The collected condensate is guided through the return line 2160 and returned to the first cooling water storage section 2130a. The circulation of flow is carried out in a continuous and consistent manner. When condensate is continuously collected in the first cooling water storage section 2130a, the level of the first cooling water storage section 2130a gradually increases. Then, as the level of the first cooling water storage section 2130a increases, the condensate flows into the second cooling water storage section 2130b from the first cooling water storage section 2130a. During the process, the additive injection section 2180 injects an additive to the condensate. Accordingly, the revolatilization of the condensate flowing into the second cooling water storage section 2130b may be suppressed. Noncondensible gas discharged along with the condensate is flowed into the filter facility 2170 and filtered out. The noncondensible gas is supplied to the filter or absorbent 2171 through a flow path formed by the gas line 2173. The filter or absorbent 2171 separates radioactive materials from the noncondensible gas. The filtered noncondensible gas is discharged to an inside of the containment 212 through the gas discharge section 2172. The present disclosure has an effect of reducing a concentration of radioactive materials within the containment 212 at an early stage by the filter facility 2170. Assuming an accident at a nuclear power plant, an exclusion area boundary (EAB) is set to the nuclear power plant for the safety of the general public during an accident to limit the residence of the general public. The present disclosure may reduce the exclusion area boundary by the filter facility 2170. Among the related arts, a filtered containment ventilation system (FCVS) has been developed to prevent the damage of the containment and reduce a concentration of radioactive materials discharged to an external environment during the occurrence of an accident in which a pressure within the containment greatly increases (AREVA in France, Westinghouse in United States, etc.). The concept in which a filter facility is installed at a boundary between an inside and an outside of the containment, and the boundary is open (using a rupture disc, a valve, etc.) when an accident occurs in which a pressure within the containment greatly increases, and atmosphere within the containment is discharged through the filter facility is applied to the filtered containment ventilation system (FCVS). When a design basis exceeding accident (here, design basis exceeding accident denotes an accident in which an internal pressure of the containment greatly increases above a design pressure) occurs at a nuclear power plant employing a filtered containment ventilation system in the related art, a rupture disc or valve installed between an inside of the containment and the filter facility is open, and a flow is formed by a pressure difference formed at an inside and an outside of the containment (a difference between a high pressure formed within the containment and an external atmospheric pressure). Then, atmosphere (air and steam) within the containment is passed through the filter facility and then discharged to an outside thereof by the flow. However, the filtered containment ventilation system in the related art does not operate during the occurrence of a design basis accident (here, design basis accident denotes an accident in which an internal pressure of the containment is within a design pressure range). Accordingly, the filtered containment ventilation system in the related art is unable to reduce the concentration of radioactive materials within the containment during the occurrence of a design basis accident, thereby causing a problem in which the amount of radioactive materials being leaked out of the containment cannot be greatly suppressed. On the contrary, the filter facility 2170 according to the present disclosure is configured to operate during the occurrence of all accidents including a design basis accident as well as a design basis exceeding accident. The filter facility 2170 is configured to discharge noncondensible gas having a very low radioactive material concentration into the containment 212. Radioactive materials are collected into the filter facility 2170 while passing through the filter or absorbent 2171. The present disclosure may very effectively reduce a radioactive material concentration within the containment 212, thereby significantly reducing the amount of radioactive materials being leaked out of the containment 212. Furthermore, the present disclosure may decrease the exclusion area boundary. FIG. 14 is a conceptual view illustrating when a normal operation is carried out on a passive safely system 2200 and a nuclear power plant 220 including the same associated with yet still another embodiment of the present disclosure. An emergency cooling water storage section 2211 is installed outside the containment 222, and cooling water is stored within the emergency cooling water storage section 2211. A heat exchanger 2212 is installed within the emergency cooling water storage section 2211. An inlet header 2212a and an outlet header 2212b of the heat exchanger 2212 are respectively connected to a connected line 2213 passing through the containment 222. The connected line 2213 is connected to a circulation inducing jet device 2220 to supply the first fluid and second fluid jetted from the circulation inducing jet device 2220 to the heat exchanger 2212. Furthermore, another connected line 2213 is connected to a filter facility 2270 to supply condensate and noncondensible gas formed by the cooling of the first fluid and second fluid to the filter facility 2270. The heat exchanger 2212 illustrated in FIG. 14 is immersed in the cooling water of the emergency cooling water storage section 2211. In this aspect, the cooling section 2210 of FIG. 14 may be divided into an immersion type. However, the present embodiment illustrates an immersion type as an example, but may be also configured with an air-cooled type by exposing the heat exchanger 2212 to atmosphere and installing a duct (not shown) without installing the emergency cooling water storage section 2211. The circulation inducing jet device 2220 is connected to a steam discharge line 2242 to supply the first fluid from a steam generator 221b. The circulation inducing jet device 2220 is connected to the connected line 2213 to jet the first fluid received from the steam discharge line 2242 and the second fluid entrained by a pressure drop to the connected line 2213. The heat exchanger 2212 allows the first fluid and second fluid entrained through the connected line 2213 to pass through an internal flow path to exchange heat with the cooling water of the emergency cooling water storage section 2211. The filter facility 2270 is installed within the containment 222. The containment 222 is connected to the heat exchanger 2212 by the connected line 2213 to receive condensate and noncondensible gas from the heat exchanger 2212. A return line 2260 is extended from a lower portion of the filter facility 2270 to a first cooling water storage section 2230a to transfer condensate supplied to the filter facility 2270. FIG. 15 is a conceptual view illustrating when an event occurs on a passive safely system 2200 and a nuclear power plant 220 including the same illustrated in FIG. 14. When an accident occurs on the nuclear power plant 220, isolation valves 223b, 224b installed at the feedwater line 223a and steam line 224a, respectively, are closed. Then, an isolation valve 2241a and a check valve 2241b installed at a fluid supply line 2241 are open, and an isolation valve 2215a and a check valve 2251b installed at a line 2215 for connecting between a second cooling water storage section 2230b and a safety injection line 225a are also open. The cooling water of the second cooling water storage section 2230b is safely injected into a reactor coolant system 221 along with the safety injection facility 225. The cooling water of the first cooling water storage section 2230a is supplied to a steam generator 221b through the fluid supply line 2241 to remove sensible heat within the reactor coolant system 221 and residual heat in the core 221a. During an accident, a steam discharge line 2242 and an isolation valve 2242a are open. The first fluid discharged from the steam generator 221b is evaporated through the steam discharge line 2242, and supplied to the circulation inducing jet device 2220. The second fluid is entrained into the circulation inducing jet device 2220 and jetted to the connected line 2213 along with the first fluid. The first fluid and second fluid are supplied to the heat exchanger 2212 through the connected line 2213. The first fluid and second fluid are cooled and condensed by the cooling water of the emergency cooling water storage section 2211 while passing through an internal flow path of the heat exchanger 2212. Condensate and noncondensible gas discharged from the heat exchanger 2212 are supplied to the filter facility 2270 through the connected line 2213. The condensate in the filter facility 2270 is returned to the first cooling water storage section 2230a through the return line 2260. The noncondensible gas in the filter facility 2270 is filtered out while passing through the filter or absorbent 2271. The filtered noncondensible gas is discharged to an inside of the containment 222 through a gas discharge section 2272. FIG. 16 is a conceptual view illustrating when a normal operation is carried out on a passive safely system 2300 and a nuclear power plant 230 including the same associated with still yet another embodiment of the present disclosure. FIG. 17 is a conceptual view illustrating when an event occurs on the passive safely system 2300 and the nuclear power plant 230 including the same illustrated in FIG. 16. The passive safely system 2300 is distinguished from that of the foregoing embodiment in that it includes a first cooling water storage section 2330a but does not include a second cooling water storage section. The passive safely system 2300 removes sensible heat within a reactor coolant system 231 and residual heat in a core 231a using a secondary system. The pure cooling water of the first cooling water storage section 2330a is supplied to a steam generator 231b for a passive residual heat removal function. A cooling section 2310 illustrated in FIGS. 16 and 17 is the same as the cooling section 2110 illustrated in FIG. 12, and thus may be divided into a circulation type. The remaining configuration and operation will be substituted by the description of FIGS. 12 and 13. FIG. 18 is a conceptual view illustrating when a normal operation is carried out on a passive safely system 2400 and a nuclear power plant 240 including the same associated with yet still another embodiment of the present disclosure. FIG. 19 is a conceptual view illustrating when an event occurs on the passive safely system 2400 and the nuclear power plant 240 including the same illustrated in FIG. 18. The passive safely system 2400 uses a primary system. A second cooling water storage section 2430b stores borated water to be injected into a reactor coolant system 241 to maintain a water level of the reactor coolant system 241. The second cooling water storage section 2430b is connected to a safety injection line 245a by a fluid supply line 2441. A steam discharge line 2442 is connected to the reactor coolant system 241 to supply the first fluid discharged from the reactor coolant system 241 to a circulation inducing jet device 2420. An additive injection section 2480 may be installed at a line connected from a lower portion 2415b of a casing 2415 to the second cooling water storage section 2430b to inject an additive to condensate collected into the second cooling water storage section 2430b. For example, the additive injection section 2480 may be installed at an outlet of a return line 2460. The additive injection section 2480 may inject an additive to condensate returned to the second cooling water storage section through the return line 2460. When a loss of coolant accident occurs in this embodiment, both the first fluid and second fluid are a primary fluid. However, in case of a steam line break, the first fluid is a primary fluid, and the second fluid is a secondary fluid. The primary fluid circulates the passive safely system 2400 from the time of being started from the second cooling water storage section 2430b to the time of being returned to the second cooling water storage section 2430b. The passive safely system 2400 is different from the foregoing embodiment in that it uses a primary system without using a steam generator 241b. FIG. 20 is a conceptual view illustrating when a normal operation is carried out on a passive safely system 2500 and a nuclear power plant 250 including the same associated with still yet another embodiment of the present disclosure. FIG. 21 is a conceptual view illustrating when an event occurs on the passive safely system 2500 and the nuclear power plant 250 including the same illustrated in FIG. 20. A cooling section 2510 may include a first heat exchanger 2512′ and a second heat exchanger 2512″. The first heat exchanger 2512′ is installed within a containment 252 to cool the first fluid and second fluid jetted from a circulation inducing jet device 2520. The present embodiment illustrates an immersion type as an example. However, according to the characteristics of the nuclear power plant 250, the cooling section 2510 may be also configured with an air-cooled type by exposing the second heat exchanger 2512″ to atmosphere and installing a duct (not shown) without installing the emergency cooling water storage section 2511. The second heat exchanger 2512″ is installed within the emergency cooling water storage section 2511. Furthermore, an isolation valve 2514a or check valve 2514b is installed at a connected line 2513. The second heat exchanger 2512″ transfers heat that has transferred to a fluid circulating a closed flow path to cooling water within the emergency cooling water storage section 2511. When the first fluid and second fluid are jetted to a casing 2515 from a circulation inducing jet device 2520, the first fluid and second fluid are cooled and condensed by a fluid circulating a closed flow path of the first heat exchanger 2512′. Condensate formed by the condensation of the first fluid and second fluid is collected into a lower portion 2515b of the casing 2515, and returned to a first cooling water storage section 2530a through a return line 2560. The fluid within the closed flow path that has received heat from the first fluid and second fluid flows to the second heat exchanger 2512″ through the connected line 2513. Heat in the second heat exchanger 2512″ is transferred to the cooling water of the emergency cooling water storage section 2511 from the fluid within the closed flow path. The emergency coolant storage section evaporates cooling water to discharge heat to an outside thereof. The fluid continuously receives the heat of the first fluid and second fluid while circulating the closed flow path. The cooling section 2510 may further include a makeup tank 2516. The makeup tank 2516 is formed to store makeup water. The makeup tank 2516 is connected to the connected line 2513 to supply makeup water to the closed flow path or accommodate the excess water of the closed flow path. An isolation valve 2516b is installed at a line 2516a connecting between the makeup tank and the connected line 2513. The isolation valve 2516b may be configured to be open at a time point at which the supply of makeup water is required or configured to be open in advance. A fluid within the closed flow path may receive makeup water to maintain a sufficient water level. FIG. 22 is a conceptual view illustrating when a normal operation is carried out on a passive safely system 2600 and a nuclear power plant 260 including the same associated with yet still another embodiment of the present disclosure. An emergency cooling water storage section 2611 is installed outside a containment 262, and a heat exchanger 2612 is installed within the containment 262. A connected line 2613 may include a first connected line 2613a through a fourth connected line 2613d. The first connected line 2613a is connected to the emergency cooling water storage section 2611 and the heat exchanger 2612 to form a flow path for supplying the cooling water of the emergency cooling water storage section 2611 to the heat exchanger 2612. In FIG. 22, the first connected line 2613a indicates a portion extended from a lower portion of the emergency cooling water storage section 2611 to a first header 2612a of the heat exchanger 2612. The first connected line 2613a allows the cooling water of the emergency cooling water storage section 2211 to flow into the heat exchanger 2612 to implement water-cooled type cooling. A second connected line 2613b is extended from the heat exchanger 2612 to an outside of the containment 262 to discharge the cooling water of the emergency cooling water storage section 2611 that has passed through the heat exchanger 2612 to an outside thereof. In FIG. 22, the second connected line 2613b indicates a portion extended from a second header 2612b of the heat exchanger 2612 to an outside of the containment 222 and continuously extended in a downward bending manner. A third connected line 2613c is branched from the second connected line 2613b to form a flow path for supplying atmosphere outside the containment 262 to the heat exchanger 2612. In FIG. 22, the third connected line 2613c indicates a portion branched from the second connected line 2613b and continuously extended to an outside of the containment 262. The third connected line 2613c allows the atmosphere of the containment 262 to flow into the heat exchanger 2612 so as to implement air-cooled type cooling when the cooling water of the emergency cooling water storage section 2611 is exhausted. A fourth connected line 2613d is branched from the first connected line 2613a to an outside of the containment 262 to discharge atmosphere heated while passing through the heat exchanger 2612 to an outside thereof. In FIG. 22, the fourth connected line 2613d indicates a portion branched from the first connected line 2613a and connected to a duct 2617. The duct 2617 is an air circulation system for exhausting atmosphere discharged from the heat exchanger 2612. At least one of isolation valves 2614a′. 614a″, 614b′, 614b″, 614c, 614d is installed at each connected line 2613a, 613b, 613c, 613d. Cooling medium flowing into the heat exchanger 2612 differs according to which one of the isolation valves 2614a′. 614a″, 614b′, 614b″, 614c, 614d is open. The cooling method of a cooling section 2610 may be switched to either one of water-cooled type cooling and air-cooled type cooling according to the switching of the isolation valves 2614a′. 614a″, 614b′, 614b″, 614c, 614d. The present embodiment illustrates a case of mixing the water-cooled type with the air-cooled type, but may be also configured with an air-cooled exclusive type by removing a water-cooled type related facility according to the characteristics of the nuclear power plant. FIGS. 23 and 24 are conceptual views illustrating when an event occurs on a passive safely system 2600 and a nuclear power plant 260 including the same illustrated in FIG. 22. The passive safely system 2600 undergoes a sequential operation process in FIGS. 23 and 24. First, referring to FIG. 23, the pure cooling water of a first cooling water storage section 2630a is supplied to a steam generator 261b as an accident occurs on the nuclear power plant. The first fluid discharged from the steam generator 261b is supplied to a circulation inducing jet device 2620 through a steam discharge line 2642. The circulation inducing jet device 2620 jets the first fluid and second fluid to an inside of the casing 2615. When an accident occurs, isolation valves 2614a′, 614b″ installed at a first connected line 2613a and a second connected line 2613b are open by an associated signal. Other isolation valves 2614″, 614′ installed at the first connected line 2613a and second connected line 2613b may be set to be open at the time point at which the maintenance is required, and to be closed when the isolation of the containment 262 is required due to the damage of the system or the like during an accident. Cooling water that has been stored in the emergency cooling water storage section 2611 is injected into the heat exchanger 2612 by a gravity water head. The cooling water is injected into the heat exchanger 2612 through the first connected line 2613a. The cooling water injected into the heat exchanger 2612 receives heat from the first fluid and second fluid in the heat exchanger 2612. The cooling water is discharged to an outside of the containment 262 through the second connected line 2613b. The first fluid and second fluid are cooled and condensed, and collected to a lower portion of the casing 2615. Noncondensible gas is filtered out by the filter facility 2670. Condensate is returned again to the first cooling water storage section through a return line 2660. In the aspect that cooling water that has been stored in the emergency cooling water storage section 2611 is injected into the heat exchanger 2612, the cooling section 2610 illustrated in FIGS. 22 through 24 may be divided into an injection type. The injection type may be divided again into a gravity injection type and a gas injection type. The gravity injection type injects cooling water based on a gravity head. The gas injection type pressurizes cooling water with gas filled in the sealed emergency cooling water storage section 2211 to inject cooling water. In the aspect that cooling water injection in FIGS. 22 through 24 is carried out by gravity, the cooling section 2610 is divided into a gravity injection type. The gravity injection type cooling may continue until the cooling water of the emergency cooling water storage section 2611 is exhausted. The operation of the passive safely system 2600 after the cooling water of the emergency cooling water storage section 2611 is exhausted will be described with reference to FIG. 24. Referring to FIG. 24, an isolation valve 2614b″ installed at the second connected line 2613b is closed, and isolation valves 2614c, 614d installed at the third connected line 2613c and fourth connected line 2613d are open. Accordingly, cooling carried out with a water-cooled type is switched to an air-cooled type. Atmosphere outside the containment 262 flows into the heat exchanger 2612 through the third connected line 2613c. The atmosphere that has received heat from the first fluid and second fluid in the heat exchanger 2612 flows again into the duct 2617 through the fourth connected line 2613d. The atmosphere is discharged to an outside thereof through the duct 2617. As illustrated in FIGS. 23 and 24, when cooling is carried out with a mixed type of both the water-cooled type and air-cooled type, the passive safely system 2600 may be more securely prepared for an accident. Cooling is carried out with the water-cooled type having a high cooling efficiency at an early stage of the accident with a large thermal load, and consistent cooling may be carried out with the air-cooled type in which the makeup of cooling water is not required at a later stage of the accident with a low thermal load. FIG. 25 is a conceptual view illustrating when a normal operation is carried out on a passive safely system 2700 and a nuclear power plant 270 including the same associated with still yet another embodiment of the present disclosure, FIG. 26 is an enlarged conceptual view illustrating part of the passive safely system 2700 illustrated in FIG. 25. A return line 2760 is connected to a lower portion of the casing 2715, and extended from the lower portion 2715b of the casing 2715 to a first cooling water storage section 2730a. The return line 2760 forms a flow path of condensate collected in the casing 2715. Condensate is returned to the first cooling water storage section 2730a from the casing 2715 through the return line 2760. A return line 2774 is connected to a lower portion of a filter facility 2770 in addition to the return line 2760 connected to the lower portion 2715b of the casing 2715. In order to distinguish two return lines 2760, 2774, the return line 2760 connected to the casing 2715 is referred to as a first return line 2760, and the return line 2774 connected to a lower portion of the filter facility 2770 is referred to as a second return line 2774. The second return line 2774 is connected to the first return line 2760. Condensate generated during the cooling process of the cooling section 2710 may be collected to a lower portion of the casing 2715, but part thereof may be flowed to a lower portion of the filter facility 2770. Condensate collected to the lower portion 2715b of the casing 2715 is returned to the first cooling water storage section 2730a through the first return line 2760. Condensate collected to the lower portion of the filter facility 2770 is returned to the first cooling water storage section 2730a through the second connected line 2774. Referring to FIG. 26, at least part of the first return line 2760′ forms a height difference from another part thereof. When the first return line 2760′ with the foregoing structure is used, it may be possible to block the discharge of air, thereby preventing noncondensible gas containing radioactive materials from being discharged through the first return line 2760. FIGS. 27 through 29 are conceptual views illustrating a circulation inducing jet device 1120 and a modified example thereof 3820, 3920. FIG. 27 illustrates the circulation inducing jet device 1120 illustrated in FIG. 1, and uses the principle of a jet pump. The description thereof will be substituted by the earlier description. FIGS. 28 and 29 as a modified example of a circulation inducing jet device 3820, 3920 different from FIG. 27 uses the principle of a turbine pump other than a jet pump. The circulation inducing jet device 3820, 3920 may include a first fluid jetting section 3821, 3921, a second fluid entraining section 3822b, 3922b, a circulating fluid jetting section 3822, 3922, a turbine blade 3823, 3923, and a pump impeller 3824, 3924. The turbine blade 3823, 3923 and pump impeller 3824, 3924 are installed at an outlet of the first fluid jetting section 3821, 3921, and when the rotational force thereof is used, it may be possible to entrain the second fluid within the containment to promote circulation flow within the containment. Referring to FIG. 28, a turbine may include a relatively small-sized turbine blade 3823 disposed at an outlet of the first fluid jetting section 3821 and a relatively large-sized pump impeller 3824 disposed at a position separated by a predetermined distance from the outlet of the first fluid jetting section 3821. Referring to FIG. 29, the position of the pump impeller 3924 is disposed closer to the turbine blade 3923 than that of the pump impeller 3824 of FIG. 28. In the circulation inducing jet device 3820, 3920 illustrated in FIGS. 28 and 29, the turbine blade 3823, 3923 induces an efficient jetting of the first fluid, and the pump impeller 3824, 3924 induces an efficient jetting of the second fluid. The present disclosure may promote circulation flow using a circulation inducing jet device without merely depending on natural circulation to increase an efficiency of cooling an inside of the containment. The first fluid discharged from the reactor coolant system may be directly supplied to the heat exchanger without discharging the first fluid to an inside of the containment as well as the second fluid within the containment may be induced to the heat exchanger at the same time, thereby solving the problems of size increase, cost increase and safety degradation in the heat exchanger for cooling the containment in a nuclear power plant. In addition, the present disclosure may filter out non-condensate using a filter facility without discharging the non-condensate as it is to an inside of the containment. Accordingly, it may be possible to reduce a concentration of radioactive materials within the containment at an early stage. As the concentration of radioactive materials is reduced, the present disclosure may decrease the exclusion area boundary. The configurations and methods according to the above-described embodiments will not be applicable in a limited way to the foregoing passive safety facility and a nuclear power plant including the same, and all or part of each embodiment may be selectively combined and configured to make various modifications thereto. The present disclosure may be used for safety enhancement in the nuclear power plant industry. |
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description | Field The present disclosure relates to a reactor pressure vessel assembly including a flow barrier structure and/or a method of manufacturing the reactor pressure vessel assembly. Description of Related Art FIG. 1 is a cross-sectional view of a conventional natural circulation reactor pressure vessel assembly. FIGS. 2-4 are a plan view, a cross-sectional view, and a perspective view, respectively, of a portion of the reactor pressure vessel assembly in FIG. 1. Referring to FIGS. 1-4, the reactor pressure vessel assembly 100 includes a housing H that surrounds a core inlet region 114, a shroud 104, a reactor core 112, a chimney assembly 108, and steam separators 118. The reactor core 112 is over the core inlet region 114. The chimney assembly 108 is between the steam separators 118 and the reactor core 112. The steam separators 118 are over the chimney assembly 108. The reactor core 112 may be defined by an inner surface of the shroud 104, a core plate 116 secured to a bottom of the shroud 104, and a top guide 120 secured to a top of the shroud 104. The shroud 104 may be a hollow cylindrical structure that separates the reactor core 112 from the downcomer annulus flow in the annulus A. The core plate 116 may support control rods and fuel assemblies that include a plurality of fuel rods in the reactor core 112. The top guide 120 may provide lateral support to the top of the fuel assemblies. The core plate 116 may support the control rods laterally. The control rods may be vertically supported by control rod guide housings that are welded to a bottom head in the reactor pressure vessel assembly. The chimney assembly 108 includes a chimney barrel B, chimney partitions C, a chimney head CH, and a plenum 106. An inner surface of the chimney barrel B defines a space between the reactor core 112 and the steam separators 118. The plenum 106 is a portion of the space defined by the inner surface of the chimney barrel B between a lower surface of the chimney head CH and an upper surface of the chimney partitions C. A height of the plenum 106 may be about 2 meters, but is not limited thereto. The chimney partitions C are located inside the chimney barrel B. The chimney partitions C divide the space defined by the inner surface of the chimney barrel B into smaller sections. The annulus A is defined by a space between an inner surface of the housing H and outer surfaces of the chimney assembly 108 (e.g., outer surfaces of the chimney barrel B) and reactor core 112 (e.g., outer surface of the shroud 104). Together, an inner surface of the chimney assembly 108 (e.g., inner surface of the chimney barrel B) and an inner surface of the reactor core 112 (e.g., an inner surface of the shroud 104) define a conduit for transporting a gas-liquid two phase flow stream from the reactor core 112 through the chimney assembly 108 to the steam separators 118. The upward arrows in FIG. 1 indicate a flow direction of the gas-liquid two phase flow stream through the reactor core 112, chimney assembly 108, and steam separators 118. The chimney partitions C act to channel the gas-liquid two phase flow exiting the reactor core 112 into the chimney assembly 108 in order to limit cross flow and/or reduce the potential for recirculating eddies. The steam separators 118 may separate a gas portion of the gas-liquid two phase flow that flows through the steam separators 118 out a top of the reactor pressure vessel assembly 100, as indicated by the arrows above the steam separators 118 in FIG. 1. A remaining portion of the gas-liquid two phase flow that corresponds to the downcomer fluid from the steam separators 118 and steam dryer (not shown), referred to as separator downcomer flow, flows down from the top of the reactor pressure vessel assembly 100. The separator downcomer flow may come from two sources: a steam dryer (not shown) and a return from the steam separators 118. A substantial portion of the separator downcomer flow (e.g., about 97%) may come from the return flow of the steam separators 118 and a comparatively smaller portion (e.g., about 3%) of the separator downcomer flow may come from the steam dryer (not shown). However, the relative contributions to the separator downcomer flow from the return flow of the steam separators 118 and the steam dryer (not shown) are not limited to about 97% and about 3%, respectively, and may be different depending on operation conditions and/or variations in design. FIG. 3 illustrates a fluid level L of the separator downcomer flow, but the fluid level L of the separator downcomer flow may vary from the fluid level L indicated in FIG. 3 depending on operation conditions. The reactor pressure vessel assembly 100 includes at least one feedwater sparger 126 in the housing H that is configured to deliver a sub-cooled feedwater into the annulus A. Each feedwater sparger 126 is connected to a corresponding feedwater opening defined by the housing H. The reactor pressure vessel assembly 100 may include a plurality of feedwater spargers 126 arranged in a circular pattern over the chimney assembly 108 and connected to a plurality of feedwater openings defined by the housing H. The housing defines a feedwater opening for each feedwater sparger 126. The annulus A is in fluid communication with the feedwater opening connected to the feedwater sparger 126 and the conduit for transporting of a gas-liquid two phase flow stream from the reactor core 112 through the chimney assembly 108 to the steam separators 118. As shown in FIG. 3, a support plate 128 may be arranged a distance H1 above the chimney head CH, but below a height of the feedwater spargers 126. The support plate 128 may be secured to the chimney head CH. For example, the support plate 128 may be welded to the steam separator stand pipes SP. Chimney head bolds (not shown) may fit inside the support plate 128 through slip fit holes. The support plate 128 may support the outer stand pipes, and may support the chimney head bolts, laterally. The support plate 128 may have a ring structure with a width W1. From a plan view, as shown in FIG. 2, the feedwater spargers 126 expose the width W1 of the support plate 128 below. The steam separators 118 are over an area surrounded by the support plate 128, but the steam separators 118 may be arranged so they are not directly over the support plate 128 in a vertical direction. In other words, as shown in FIG. 2, the support plate 128 may surround the steam separators 118 in a plan view. As shown in FIG. 4, even though some of the outer steam separators 118 may be on top of stand pipes SP that partially contact the support plate 128, the steam separators 118 are not directly over a portion of the support plate 128 in a vertical direction. As shown in FIG. 4, only part of the circumference of the outer stand pipes SP is in contact with support plate 128 where the outer stand pipes intersect the support plate 128. As indicated by the down arrows in the annulus A of FIG. 1 and the arrows in the core inlet region 114 of FIG. 1, the sub-cooled feedwater may flow down the annulus A through the core inlet region 114 into the reactor core 112. The arrows in FIG. 3 illustrate part of the separator downcomer flow may be redirected to flow around the feedwater sparger 126 and the support plate 128 into the annulus A. The mixture of the sub-cooled feedwater flowing in the annulus A with the portion of the separator downcomer flow that is redirected into the annulus A may be referred to as the annulus downcomer flow. In the reactor core 112, fuel rods may heat the annulus downcomer flow received from the core inlet region 114 and the portion of the separator downcomer flow received from the top of the reactor pressure vessel assembly 100 to provide the gas-liquid two phase flow stream that flows upward from the reactor core 112 through the chimney assembly 108 to the steam separators 118. Complete mixing (and/or a desired level of mixing) between the separator downcomer flow and the sub-cooled feedwater does not occur. All (or substantially all) separator downcomer flow may be directed into the annulus A; however, at least a portion of the separator downcomer flow may bypass the sub-cooled feedwater and avoid mixing or reduce a degree of mixing. In the conventional natural circulation reactor pressure vessel assembly 100, there is incomplete mixing of the separator downcomer flow and the sub-cooled feedwater before delivery to the reactor core 112. A temperature of the sub-cooled feedwater is generally less than a temperature of the separator downcomer flow. Consequently, the incomplete mixing between the separator downcomer flow and sub-cooled feedwater may cause temperature variations into the fuel rods and supports for the fuel rods in the reactor core 112. Accordingly, improved mixing between the separator downcomer flow and sub-cooled feedwater before entry into the reactor core 112 is desired. Some example embodiments relate to a reactor pressure vessel assembly including a fluid mixing plenum. Some example embodiments relate to a reactor pressure vessel assembly including a flow barrier structure. According to an example embodiment, a reactor pressure vessel assembly includes a reactor core; steam separators over the reactor core; a chimney between the reactor core and the steam separators; a housing surrounding the reactor core, the chimney, and the steam separators; at least one feedwater sparger in the housing; and a flow barrier structure in the housing. An inner surface of the chimney and an inner surface of the reactor core define a conduit for transporting a gas-liquid two phase flow stream from the reactor core through the chimney to the steam separators. The housing defines at least one feedwater opening. An inner surface of the housing, an outer surface of the chimney, and an outer surface of the reactor core define an annulus in fluid communication with the at least one feedwater opening and the conduit. Each feedwater sparger is connected to a corresponding one of the at least one feedwater opening. Each feedwater sparger is configured to deliver a sub-cooled feedwater into the annulus. The flow barrier structure is spaced apart in a vertical direction over the chimney and below the steam separators. The flow barrier structure is configured to force mixing between the sub-cooled feedwater and a downcomer fluid from the steam separators. At least one of the outer steam separators may be vertically over a portion of the flow barrier structure in a plan view. At least one of the outer steam separators may be on top of an outer stand pipe, where the outer stand pipe may include a cross-section with an entire perimeter in contact with the flow barrier structure. The flow barrier structure may have a ring shape. A height of the flow barrier structure in the housing may be about level with the at least one feedwater sparger. The flow barrier structure may include stainless steel. The at least one feedwater sparger may be a plurality of feedwater spargers arranged in a circular pattern over the chimney. The at least one feedwater opening may be a plurality of feedwater openings defined by the housing. The feedwater sparger may be connected to the feedwater openings. A dam plate may be in the housing between the chimney and the steam separators. A distance between an outer edge of the dam plate and the inner surface of the housing may be greater than or equal to a distance between an outer edge of the flow barrier structure and the inner surface of the housing. A vertical distance between the dam plate and a top of the chimney may be different than a vertical distance between the flow barrier structure and the top of the chimney. The vertical distance between the dam plate and the top of the chimney may be less than the vertical distance between the flow barrier structure and the top of the chimney. Alternatively, the vertical distance between the dam plate and the top of the chimney may be greater than the vertical distance between the flow barrier structure and the top of the chimney. A diagonal distance between the outer edge of the dam plate and an inner surface of the flow barrier structure may be equal to or greater than a width of the annulus. The reactor pressure vessel assembly may further include a backflow dam on the at least one feedwater sparger. The backflow dam may cover a top of the annulus. An edge of the backflow dam may be one of spaced apart from the inner surface of the housing and connected to the inner surface of the housing. The flow barrier structure may have a tub shape. The flow barrier structure may be configured to force the downcomer fluid past and over the at least one fed-fluid sparger. According to an example embodiment, a reactor pressure vessel assembly includes a reactor core; steam separators over the reactor core; a housing surrounding the reactor core and the steam separators; at least one feedwater sparger in the housing; and a flow barrier structure in the housing below the separators. An inner surface of the reactor core defines a conduit for transporting a gas-liquid two phase flow stream from the reactor core to the steam separators. The housing defines at least one feedwater opening. An inner surface of the housing and an outer surface of the reactor core define an annulus in fluid communication with the at least one feedwater opening and the conduit. Each feedwater sparger is connected to a corresponding one of the at least one feedwater opening. Each feedwater sparger is configured to deliver a sub-cooled feedwater into the annulus. The flow barrier structure is configured to force mixing between the sub-cooled feedwater and a downcomer fluid from the separators. A top of the flow barrier structure in housing is about level with the at least one sparger. At least one of the steam separators may be vertically over a portion of the flow barrier structure in a plan view. At least one of the outer steam separators may be vertically over a portion of the flow barrier structure in a plan view. At least one of the outer steam separators may be on top of an outer stand pipe, where the outer stand pipe may include a cross-section with an entire perimeter in contact with the flow barrier structure. The flow barrier structure may have one of a ring shape and a tube shape. The reactor pressure vessel assembly may further include a dam plate in the housing between the reactor core and the separators. A distance between an outer edge of the dam plate and the inner surface of the housing may be equal to or greater than a distance between an outer edge of the flow barrier structure and the inner surface of the housing. A vertical distance between the dam plate and a top of the reactor core may be different than a vertical distance between the flow barrier structure and the top of the reactor core. According to an example embodiment, a method of manufacturing a reactor pressure vessel assembly includes disposing steam separators over a reactor core; disposing a chimney between the reactor core and the steam separators; disposing a housing surrounding the reactor core, the chimney, and the steam separators; disposing at least one feedwater sparger in the housing; and disposing a flow barrier structure in the housing. An inner surface of the chimney and an inner surface of the reactor core define a conduit for transport of a gas-liquid two phase flow stream from the reactor core through the chimney to the steam separators. The housing defines at least one feedwater opening. An inner surface of the housing, an outer surface of the chimney, and an outer surface of the reactor core define an annulus in fluid communication with the at least one feedwater opening and the conduit. The flow barrier structure is spaced apart in a vertical direction over the chimney and below the separators. The flow barrier structure is configured to force mixing between the sub-cooled feedwater and a downcomer fluid from the steam separators. At least one of the steam separators may be vertically over a portion of the flow barrier structure in a plan view. At least one of the outer steam separators may be vertically over a portion of the flow barrier structure in a plan view. At least one of the outer steam separators may be on top of an outer stand pipe, where the outer stand pipe may include a cross-section with an entire perimeter in contact with the flow barrier structure. The method further includes connecting each feedwater sparger to a corresponding one of the at least one feedwater opening. Each feedwater sparger is configured to deliver a sub-cooled feedwater into the annulus. The flow barrier structure may have a ring shape. The disposing the flow barrier structure may include arranging the flow barrier structure at a height in the housing that is about level with the at least one feedwater sparger. The flow barrier structure may include stainless steel. The flow barrier structure may have a tub shape. The flow barrier structure may be configured to force the downcomer fluid past and over the at least one feedwater sparger. The method may further include placing a dam plate in the housing between the chimney and the separators. A distance between an outer edge of the dam plate and the inner surface of the housing may be equal to or greater than a distance between an outer edge of the flow barrier structure and the inner surface of the housing. A vertical distance between the dam plate and a top of the chimney may be different than a vertical distance between the flow barrier structure and the top of the chimney. The vertical distance between the dam plate and the top of the chimney may be greater than the vertical distance between the flow barrier structure and the top of the chimney. A diagonal distance between the outer edge of the dam plate and an inner surface of the flow barrier structure is greater than a width of the annulus. Example embodiments will now be described more fully with reference to the accompanying drawings, in which some example embodiments are shown. Example embodiments, may, however, be embodied in many different forms and should not be construed as being limited to the embodiments set forth herein; rather, these example embodiments are provided so that this disclosure will be thorough and complete, and will fully convey the scope of example embodiments to those of ordinary skill in the art. In the drawings, like reference numerals in the drawings denote like elements, and thus their description may be omitted. It should be understood that when an element or layer is referred to as being “on,” “connected to,” “coupled to,” or “covering” another element or layer, it may be directly on, connected to, coupled to, or covering the other element or layer or intervening elements or layers may be present. In contrast, when an element is referred to as being “directly on,” “directly connected to,” or “directly coupled to” another element or layer, there are no intervening elements or layers present. Like numbers refer to like elements throughout the specification. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It should be understood that, although the terms first, second, third, etc. may be used herein to describe various elements, components, regions, layers and/or sections, these elements, components, regions, layers, and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region, layer, or section from another region, layer, or section. Thus, a first element, component, region, layer, or section discussed below could be termed a second element, component, region, layer, or section without departing from the teachings of example embodiments. Spatially relative terms (e.g., “beneath,” “below,” “lower,” “above,” “upper,” and the like) may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It should be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the term “below” may encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. The terminology used herein is for the purpose of describing various embodiments only and is not intended to be limiting of example embodiments. As used herein, the singular forms “a,” “an,” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “includes,” “including,” “comprises,” and/or “comprising,” when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. Example embodiments are described herein with reference to cross-sectional illustrations that are schematic illustrations of idealized embodiments (and intermediate structures) of example embodiments. As such, variations from the shapes of the illustrations as a result, for example, of manufacturing techniques and/or tolerances, are to be expected. Thus, example embodiments should not be construed as limited to the shapes of regions illustrated herein but are to include deviations in shapes that result, for example, from manufacturing. Thus, the regions illustrated in the figures are schematic in nature and their shapes are not intended to illustrate the actual shape of a region of a device and are not intended to limit the scope of example embodiments. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which example embodiments belong. It will be further understood that terms, including those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and will not be interpreted in an idealized or overly formal sense unless expressly so defined herein. FIGS. 5-6 are a cross-sectional view and a perspective view, respectively, of a portion of a reactor pressure vessel assembly according to an example embodiment. The support plate 128 described previously with reference to FIGS. 1-4 may be omitted from the portion of the reactor pressure vessel assembly illustrated in FIGS. 5-6. Alternatively, although not illustrated, the portion of the reactor pressure vessel assembly in FIGS. 5-6 may further include the support plate 128 described previously in FIGS. 1-4. However, a non-limiting example is described below where the support plate 128 is omitted in order to avoid obscuring features of the portion of the reactor pressure vessel assembly in FIGS. 5-6. Referring to FIGS. 5-6, the portion of the reactor pressure vessel assembly includes a flow barrier plate 134 as a flow barrier structure in the housing H. The flow barrier plate 134 is spaced apart in a vertical direction over the chimney assembly 108 (e.g., chimney head CH) and below the steam separators 118. The flow barrier plate 134 may force mixing between the sub-cooled feedwater that enters the housing H through the feedwater sparger 126 and the separator downcomer flow from the steam separators 118. The flow barrier plate 134 has a width W2 that is greater than a width W1 of the support plate 128 described in FIGS. 1-4. An inner width W3 of the flow barrier plate 134 (e.g., inner diameter) is less than an inner width (e.g., inner diameter) of the support plate 128 described in FIGS. 1-4. As shown in FIGS. 5-6, some of the outer steam separators 118 may be directly over portions of the flow barrier plate 134 in the vertical direction. Consequently, some of the outer steam separators 118 may be connected to outer stand pipes SP that extend through portions of the flow barrier plate 134. For example, as shown in FIG. 6, some of the outer stand pipes SP may include a cross-section with an entire circumference in contact with the flow barrier plate 134. However, the inner stand pipes SP are not in contact with the flow barrier plate 134. The flow barrier plate 134 may be formed of steel such as type 304 or 316 stainless steel, but is not limited to these materials. The flow barrier plate 134 may have a ring shape, but may be other shapes (e.g., polygon) depending on the shape of the housing H and/or chimney assembly 108. A height H2 of the flow barrier plate 134 in the housing H above the chimney assembly 108 (e.g., chimney head CH) may be about level with one of the feedwater spargers 126. However, height H2 shown in FIG. 5 is a non-limiting example and the height H2 of the flow barrier plate 134 may be adjusted relative to the feedwater sparger 126. By positioning the flow barrier plate 134 adjacent to the feedwater sparger 126, the flow barrier plate 134 can create a flow barrier adjacent to the feedwater sparger 126, which forces the separator downcomer flow from the steam separators 118 and the sub-cooled feedwater entering through the feedwater sparger 126 to mix prior to moving through the annulus A. Additionally, because the flow barrier plate 134 has a width W2 that is greater than the width W1 of the support plate 128 in FIG. 1, a significant portion of the separator downcomer flow is redirected horizontally, which facilitates mixing before entering the annulus A. As shown by the arrows in FIG. 5, the flow barrier plate 134 redirects the normal flow path of the separator downcomer flow and enables the sub-cooled feedwater additional time to mix with the redirected separator downcomer flow in the annulus A prior to entry into the reactor core 112. Accordingly, temperature variations of the sub-cooled feedwater and the separator downcomer flow may be reduced prior to entry into the reactor core 112. A feedwater nozzle 122 (see FIG. 1) may be connected to each feedwater sparger 126 through the feedwater opening defined in the housing H. Although FIG. 1 illustrates the feedwater nozzle 122 is horizontally oriented, example embodiments are not limited thereto. For example, to help with the mixing process, the feedwater nozzle 122 may be directed upward at an angle of about 30 to 60 degrees. For example, the feedwater nozzles 122 may be tilted upward at an angle of about 45 degrees in order to create a mixing region which delays the sub-cooled feedwater from entering the normal flow path into the annulus A until the sub-cooled feedwater has had more time to mix with the separator downcomer flow. FIGS. 7-8 are a perspective view and a cross-sectional view, respectively, of a portion of a reactor pressure vessel assembly according to an example embodiment. Although not illustrated, the portion of the reactor pressure vessel assembly in FIGS. 7-8 may further include the support plate 128 described previously in FIGS. 1-4. However, a non-limiting example is described below where the support plate 128 is omitted in order to avoid obscuring features of the portion of the reactor pressure vessel assembly in FIGS. 7-8. Referring to FIGS. 7-8, the portion of the reactor vessel assembly in FIGS. 7-8 is the same as the portion of the reactor vessel assembly in FIG. 5-6, except the portion of the reactor vessel assembly in FIGS. 7-8 further includes a lower dam 136 positioned in the housing H between the steam separators 118 and the chimney assembly 108 (e.g. chimney head CH). The lower dam 136 may also be referred to as a lower dam plate. A vertical distance between the lower dam 136 and the top of the chimney assembly 108 is different than a vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. For example, a height of the lower dam 136 above the chimney assembly 108 (e.g., chimney head CH) may be less than a height H2 of the flow barrier plate 134 above the chimney assembly 108 (e.g., chimney head CH). A width W4 of the lower dam 136 may be less than a width W3 of an inner surface of the flow barrier plate 134. The width W4 of the lower dam 136 may correspond to the outer diameter of the lower dam 136 if the lower dam 136 has a circular shape. The width W3 of the inner surface of the flow barrier plate 134 may correspond to the inner diameter of the flow barrier plate 134 if the flow barrier plate 134 has a ring shape. The position of the lower dam 136 may be adjusted so the diagonal distance D1 between an edge of the lower dam 136 and an inner surface of the flow barrier plate 134 is tight enough to channel flow of the separator downcomer flow, but wide enough so the diagonal distance D1 is not too restrictive and/or does not create too much differential pressure as the separator downcomer flow is transported through the diagonal distance D1. For example, the position of the lower dam 136 may be adjusted so the diagonal distance D1 between the edge of the lower dam 136 and the inner surface of the flow barrier plate 134 may be greater than or equal to a width of the annulus A. The diagonal distance D1 may be greater than or equal to a flow area of the annulus A in a horizontal direction between an outer surface of the shroud 104 and the inner surface of the housing H. The diagonal distance D1 between the edge of the lower dam 136 and the inner surface of the flow barrier plate 134 may be about equal to a horizontal distance between an outer edge of the flow barrier plate 134 and an inner surface of the housing H. The diagonal distance D1 between the edge of the lower dam 136 and the inner surface of the flow barrier plate 134 may be greater than or equal to the horizontal distance D4 (see FIG. 11) between the outer surface of the chimney head CH and the inner surface of the housing H. In FIG. 8, the arrows illustrate how the flow barrier plate 134 and the lower dam 136 redirect the flow of the separator downcomer flow around the lower dam 136 and flow barrier plate 134 through the diagonal distance D1 between the edge of the lower dam 136 and the inner surface of the flow barrier plate 134 into the annulus A. FIGS. 9-10 are a perspective view and a cross-sectional view, respectively, of a portion of a reactor pressure vessel assembly according to an example embodiment. FIG. 11 is a cross-sectional view of a different portion of the reactor pressure vessel assembly in FIGS. 9-10. Although not illustrated, the portion of the reactor pressure vessel assembly in FIGS. 9-10 and/or 11 may further include the support plate 128 described previously in FIGS. 1-4. However, a non-limiting example is described below where the support plate 128 is omitted in order to avoid obscuring features of the portion of the reactor pressure vessel assembly in FIGS. 9-10 and/or 11. Referring to FIGS. 9-10, the portion of the reactor vessel assembly in FIGS. 9-10 is the same as the portion of the reactor vessel assembly in FIG. 5-6, except the portion of the reactor vessel assembly in FIGS. 9-10 further includes an upper dam 138 positioned in the housing H between the steam separators 118 and the chimney assembly 108 (e.g. chimney head CH). The upper dam 138 may also be referred to as an upper plate or upper dam plate. A vertical distance between the upper dam 138 and the top of the chimney assembly 108 is different than a vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. For example, a height of the upper dam 138 above the chimney assembly 108 (e.g., chimney head CH) may be greater than a height H2 of the flow barrier plate 134 above the chimney assembly 108 (e.g., chimney head CH). A width W5 of the upper dam 138 may be less than the width W3 of an inner surface of the flow barrier plate 134. The width W5 of the upper dam 138 may correspond to the outer diameter of the lower dam 138 if the upper dam 138 has a circular shape. The position of the upper dam 138 may be adjusted so the diagonal distance D2 between an edge of the upper dam 138 and an inner surface of the flow barrier plate 134 is tight enough to channel flow of the separator downcomer flow, but wide enough so the diagonal distance D2 is not too restrictive and/or does not create too much differential pressure as the separator downcomer flow is transported through the diagonal distance D2. For example, the position of the upper dam 138 may be adjusted so the diagonal distance D2 between the edge of the upper dam 138 and the inner surface of the flow barrier plate 134 may be greater than or equal to a width of the annulus A. The diagonal distance D2 may be greater than or equal to a flow area of the annulus A in the horizontal direction between the outer surface of the shroud 104 and the inner surface of the housing H. The diagonal distance D2 between the edge of the upper dam 138 and the inner surface of the flow barrier plate 134 may be about equal to the horizontal distance between an outer edge of the flow barrier plate 134 and an inner surface of the housing H. Referring to FIG. 11, the arrows in FIG. 11 illustrate how the upper dam 138 and flow barrier plate 134 redirect the flow of the separator downcomer flow around the upper dam 138 and flow barrier plate 134 through the diagonal distance D2 between the edge of the upper dam 138 and the inner surface of the flow barrier plate 134 into the annulus A. The diagonal distance D2 (see FIGS. 9-10) between the edge of the upper dam 138 and the inner surface of the flow barrier plate 134 may be greater than or equal to the horizontal distance D4 between the outer surface of the chimney head CH and the inner surface of the housing H. As discussed above with reference to FIGS. 7-8 and 9-11, a dam plate may be in the housing H between the chimney assembly 108 and the steam separators 118. A distance between the outer edge of the dam plate and the inner surface of housing may be greater than or equal to a distance between the outer edge of the flow barrier plate 134 and the inner surface of the housing. A vertical distance between the dam plate and the top of the chimney assembly 108 may be different than the vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. For example, in the case of the upper dam 138 illustrated in FIGS. 9-11, the vertical distance between the upper dam 138 and the top of the chimney of the chimney assembly 108 is greater than the vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. Alternatively, in the case of the lower dam 136 illustrated in FIGS. 7-8, the vertical distance between the lower dam 136 and the top of the chimney assembly 108 is less than the vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. The lower dam 136 and/or the upper dam 138 may be formed of steel such as type 304 or 316 stainless steel, but example embodiments are not limited to these materials. FIGS. 12-13 are a cross-sectional view and a perspective view of a portion of a reactor pressure vessel assembly according to an example embodiment. FIG. 14 is a cross-sectional view of a different portion of the reactor pressure vessel assembly in FIGS. 12-13. The support plate 128 described previously with reference to FIGS. 1-4 may be omitted in the portion of the reactor pressure vessel assembly in FIGS. 12-13 and 14. Alternatively, although not illustrated, the support plate 128 may be present in the portion of the reactor pressure vessel assembly in FIGS. 12-13 and/or 14. However, a non-limiting example is described below where the support plate 128 is omitted in order to avoid obscuring features of the portion of the reactor pressure vessel assembly in FIGS. 12-13 and/or 14. Referring to FIGS. 12-14, in an example embodiment, the portion of the reactor vessel assembly may include a flow barrier structure 146 having a tub shape. As shown by the arrows in FIGS. 12 and 14, the flow barrier structure 146 may redirect the normal flow of the separator downcomer flow up and over an outer surface (e.g., outer diameter) of the flow barrier structure 146 into the annulus A. Once in the annulus A, the redirected separator downcomer flow may be mixed with the sub-cooled feedwater that enters the annulus A through the feedwater sparger 126. As shown in FIG. 14, the feedwater sparger 126 may be just above a top of the flow barrier structure 146; therefore, a significant amount of mixing may be expected to take place within the flow barrier structure 146 before the downcomer flow and/or sub-cooled feedwater may be redirected into the annulus A, where even more mixing may be expected. As shown by the arrows in FIG. 5, the flow barrier plate 134 redirects the normal flow path of the separator downcomer flow and enables the sub-cooled feedwater additional time to mix with the redirected separator downcomer flow in the annulus A prior to entry into the reactor core 112. Accordingly, temperature variations of the sub-cooled feedwater and redirected separator downcomer flow in the annulus may be reduced prior to entry into the reactor core 112. A height H3 of the flow barrier structure 146 above the channel assembly may be greater than a height H1 of the support plate 128 described previously in FIGS. 1-4. The flow barrier structure 146 may be configured to redirect the separator downcomer flow past and over the at least one feedwater sparger. An upper surface of the flow barrier structure 146 may be adjacent to at least one feedwater sparger 126. Steam separator stand pipes SP (labeled in FIGS. 4 and 6) may extend from the flow barrier structure 146 to the steam separators 118. A support plate 144 may be connected to the flow barrier structure 146 and extend from the outer surface of the flow barrier structure 144 towards the inner surface of the housing H. The support plate 144 may have a shape that is the same as or similar to the support plate 128 described previously in FIGS. 1-4. The support plate 144 may be arranged at the height H1 above the channel assembly, but the vertical position of the support plate 144 is not limited thereto and may vary. The support plate 144 may support the stand pipes SP (see FIG. 4). Although not illustrated, chimney head bolts may be used to secure the chimney head CH to the chimney barrel B. The support plate 144 may provide support to chimney head bolts (not shown). A height of the support plate 144 may be about the same height as a top of the chimney head bolts and the chimney head bolts may extend from the support plate 144 to contact the chimney head CH and chimney barrel B. The support plate 144 and the flow barrier structure 146 may be integrally formed or formed of separate structures that are connected to each other. The flow barrier structure 146 and/or the support plate 144 may be formed of steel such as type 304 or 316 stainless steel, but example embodiments are not limited to these materials. The diagonal distance between the upper surface of the flow barrier structure 146 and the feedwater sparger 126 may be less than or about equal to a horizontal distance between the outer surface of the support plate 144 (e.g., outer diameter of the support plate 144 if the support plate 144 has a ring shape) and the inner surface of the housing H. FIG. 15 is a cross-sectional view of a portion of a reactor pressure vessel assembly according to an example embodiment. Referring to FIG. 15, in an example embodiment, the reactor pressure vessel assembly may include a backflow dam 140 on the feedwater sparger 126. The back flow dam 140 may be formed of steel such as type 304 or 316 stainless steel, but example embodiments are not limited to these materials. The backflow dam 140 may be connected to the feedwater sparger 126. For example, the back flow dam 140 may be welded to the feedwater sparger 126. The backflow dam 140 may cover a top of the annulus A. An edge of the back flow dam may be spaced apart from the inner surface of the housing H. A clearance between the backflow dam 140 and the inner surface of the housing H, as illustrated by the distance D3 in FIG. 15, may about ¼ inch to ½ inch, but example embodiments are not limited thereto. FIG. 16 is a cross-sectional view of a portion of a reactor pressure vessel assembly according to an example embodiment. Referring to FIG. 16, in an example embodiment, the reactor pressure vessel assembly may be the same as the reactor pressure vessel assembly in FIG. 15, except the backflow dam 140 may be connected to the inner surface of the housing H. When the backflow dam 140 is connected to the inner surface of the housing H, the flow barrier structure 146 redirects the separator downcomer flow through a space between the flow barrier structure 146 and feedwater sparger 126. FIG. 17 is a cross-sectional view of a portion of a reactor pressure vessel assembly according to an example embodiment. Although not illustrated, the support plate 128 described previously in FIGS. 1-4 may be present in the portion of the reactor pressure vessel assembly in FIG. 17. However, a non-limiting example is described below where the support plate 128 is omitted in order to avoid obscuring features of the portion of the reactor pressure vessel assembly in FIG. 17. Referring to FIG. 17, in an example embodiment, the reactor pressure vessel assembly may be the same as the reactor pressure vessel assembly described with reference to FIGS. 12-14, except the reactor pressure vessel assembly in FIG. 17 may further include at least one backflow dam 140 on a corresponding one of the feedwater spargers 126. The back flow dam 140 may cover a top of the annulus A. The back flow dam 140 may be spaced apart from the inner surface of the housing H or connected to the inner surface of the housing H. Alternatively, the reactor pressure vessel assembly in FIG. 17 may include a plurality of backflow dams 140 on a plurality of feedwater spargers 126, respectively. All of the backflow dams 140 may be spaced apart from the inner surface of the housing H or connected to the inner surface of the housing H. Alternatively, some of the backflow dams 140 may be spaced apart from the inner surface of the housing H and other backflow dams 140 may be connected to the inner surface of the housing H. The arrows in FIG. 17 illustrate how the flow barrier structure 146 may redirect the normal path of the downcomer separator flow into the annulus A. Although FIGS. 16 and 17 have been described in non-limiting examples where the reactor pressure vessel assembly includes the flow barrier structure 146, example embodiments are not limited thereto. According to some example embodiments, the reactor pressure vessel assemblies described previously with reference to FIGS. 5-6 and/or 7-8 and 9-11 may be modified to include the backflow dam 140 on some or all of the feedwater spargers 126. Additionally, the back flow dam 140 may be spaced apart from the inner surface of the housing H or connected to the inner surface of the housing H. In an example embodiment, referring to FIG. 1 and FIGS. 5-6, a method of manufacturing a reactor pressure vessel assembly includes disposing the steam separators 118 over the reactor core 112, disposing a chimney assembly 108 between the reactor core 112 and the steam separators 118, disposing a housing H surrounding the reactor core 112, the chimney assembly 108, the steam separators 118, disposing at least one feedwater sparger 126 in the housing H, and disposing a flow barrier structure in the housing H. An inner surface of the chimney assembly 108 and an inner surface of the reactor core 112 define a conduit for transporting a gas-liquid two phase flow stream from the reactor core 112 through the chimney assembly 108 to the steam separators 118. The housing defines at least one feedwater opening. An inner surface of the housing H, an outer surface of the chimney assembly 108 (e.g., chimney barrel B), and an outer surface of the reactor core 112 (e.g., shroud 104) defined an annulus A in fluid communication with the at least one feedwater opening. The method may further include connecting each feedwater sparger 126 to a corresponding feedwater opening. Each feedwater sparger 126 is configured to deliver a sub-cooled feedwater into the annulus A. The flow barrier structure may be the flow barrier plate 134 in FIGS. 5-6 or the flow barrier structure 146 described in FIG. 14. The flow barrier structure may be spaced apart in a vertical direction over the chimney assembly 108 and below the steam separators 118. The flow barrier structure may be configured to force mixing between the sub-cooled feedwater and a downcomer fluid from the steam separators such as the separator downcomer flow. If the flow barrier structure is the flow barrier plate 134, the method may further include placing a dam plate in the housing H between the chimney assembly 108 and the steam separators 118. A distance between an outer edge of the dam plate and the inner surface of the housing H may be equal to or greater than a distance between an outer edge of the flow barrier plate 134 and the inner surface of the housing H. A vertical distance between the dam plate and a top of the chimney assembly 108 may be different than a vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. For example, if the lower dam 136 described in FIGS. 7-8 is used as the dam plate, a vertical distance between the lower dam 136 and the top of the chimney assembly 108 may be less than a vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. Alternatively, if the upper dam 138 described in FIGS. 9-10 is used as the dam plate, a vertical distance between the upper dam 138 and the top of the chimney assembly 108 may be greater than a vertical distance between the flow barrier plate 134 and the top of the chimney assembly 108. Descriptions and/or features in each of the above-described reactor pressure vessel assemblies according to example embodiments and/or portions thereof should be considered as available in other reactor pressure vessel assemblies according to example embodiments and/or portions thereof. While a number of example embodiments have been disclosed herein, it should be understood that other variations may be possible. Such variations are not to be regarded as a departure from the spirit and scope of the present disclosure, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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062884018 | claims | 1. A source to emit a beam of charged particles, said source comprising: a charged particle emitter; an extraction electrode spaced-apart from said emitter and having an aperture, with said aperture being in superimposition with said emitter 1 and having a center, said extraction electrode defining an optical axis passing through said center; and a centering electrode disposed between said extraction electrode and said emitter, said centering electrode having an orifice lying in said optical axis. directing said charged particle beam over a path defined by an extraction field produced between said emitter and said extraction electrode; aligning said path with said optical axis by providing, between said emitter and said extraction electrode, an electrostatic deflection field. applying a first voltage to said emitter; and applying a second voltage to said extraction electrode; wherein said extraction field is the difference between said first voltage and said second voltage. applying a first centering voltage on one of said plurality of centering electrode elements disposed on a first side of said optical axis; and applying a second centering voltage on another of said plurality of centering electrode elements disposed on a second side of said optical axis, with said second side being opposite said first side. applying a first centering voltage on said first electrode element; applying a second centering voltage on said second centering electrode element; applying a third centering voltage on said third centering electrode element; and applying a fourth centering voltage on said fourth centering electrode element, with said first, second, third and fourth electrode elements being complanar. providing an additional electrostatic deflection field near said optical axis, with said electrostatic deflection field deflecting said charged particle beam along a first direction and said additional electrostatic deflection field deflecting said charged particle beam along a second direction, opposite to said first direction. a charged particle emitter; and an electrode system spaced-apart from said emitter to extract said beam of charged particles from said emitter to travel toward said electrode system, said electrode system defining an aperture in superimposition with said emitter and having a plurality of electrode segments to produce a deflection field between said electrode system and said emitter to guide said beam of charged particles through said aperture. 2. The source of claim 1 wherein said orifice has a cross-sectional area associated therewith that is substantially greater than a cross-sectional area of said aperture and superimposes a portion of said extraction electrode, with said portion defining an extraction region, with said orifice being radially and symmetrically disposed about a point, with said point lying is said optical axis. 3. The source of claim 1, wherein said centering electrode includes a plurality of spaced-apart electrode elements disposed adjacent to said extraction electrode with an insulating layer being disposed between said extraction electrode and said electrode elements. 4. The source of claim 1, wherein said emitter includes a filament having a tip, with said filament being coupled to a first source of voltage and said extraction electrode being connected to a second source of voltage to produce an electric field proximate to said tip and direct said charged particles along a path toward said extraction electrode and said centering electrode is connected to third and fourth voltage sources to align said path, proximate to said extraction electrode, with said optical axis. 5. The source of claim 1, wherein said emitter includes a filament having a tip, with said filament coupled to a first source of voltage and said centering electrode includes a pair of spaced apart electrode elements, with said extraction electrode being connected to a second source of voltage to produce an electric field proximate to said tip to direct said charged particles along a path toward said extraction electrode and each electrode element of said pair being connected to a voltage supply having a polarity associated therewith that is opposite to the polarity of the voltage supply associated with the remaining electrode element of said pair. 6. The source of claim 1, wherein said emitter is selected from the group consisting essentially of a Schottky emitter and a cold-field emitter. 7. The source of claim 1, wherein said centering electrode includes a first set of four electrode elements disposed between said extraction electrode and said emitter, and a second set of four centering electrode elements disposed between said first set of four electrode elements and said emitter, with a first insulating layer being disposed between said first set of four electrode elements and said extraction electrode and a second insulating layer being disposed between said first and second sets of four electrode elements. 8. The source of claim 7, wherein said emitter includes a filament having a tip, with said filament being coupled to a first source of voltage and said extraction electrode being connected to a second source of voltage to produce an electric field proximate to said tip and direct said charged particles along a path toward said aperture, with each of said electrode elements of said first and second sets of four electrode elements being connected to a differing voltage supply to align said path with said optical axis. 9. A method of aligning a charged particle beam, produced by an emitter, with an optical axis defined by an extraction electrode, said method comprising: 10. The method of claim 9, wherein directing said charged particle beam further includes: 11. The method of claim 9 further including providing a centering electrode having a plurality of centering electrode elements defining an orifice disposed in said optical axis, wherein directing said charged particle beam further comprises: 12. The method of claim 9 further including providing a centering electrode having a first, second, third and fourth centering electrode elements, defining an orifice disposed in said optical axis, with said first electrode element being disposed opposite to said second electrode element and said third electrode element being disposed opposite to said fourth electrode element, wherein directing said charged particle beam further comprises: 13. The method of claim 9 wherein said extraction electrode includes an aperture having a center and aligning said path with said optical axis includes providing a deflection field having a magnitude associated therewith that is asymmetrically distributed about said center to direct said charged particle beam through said aperture. 14. The method of claim 9 further including providing a second electrostatic deflection field near said optical axis, wherein said second electrostatic deflection field deflects said charged particle beam to travel approximately parallel to said optical axis. 15. The method of claim 9, further comprising: 16. A source to emit a beam of charged particles, said source comprising: 17. The source as recited in claim 16 wherein said plurality of electrode elements includes an extraction electrode defining said aperture and a centering electrode having a plurality of electrode segments, said centering electrode being disposed between said extraction electrode and said emitter and having an orifice centered about said optical axis, with a cross-sectional area of said orifice being substantially greater than a cross-sectional area of said aperture and superimposing a portion of said extraction electrode. 18. The source as recited in claim 16 wherein said electrode assembly includes an extraction electrode defining said aperture and a centering electrode defined by said plurality of electrode elements, with said plurality of electrode elements being disposed adjacent to said extraction electrode with an insulating layer being disposed between said extraction electrode and said electrode elements. 19. The source as recited in claim 17 wherein said emitter includes a filament having a tip, with said filament being coupled to a first source of voltage and said centering electrode is defined by a plurality of spaced-apart electrode segments, with said extraction electrode being connected to a second source of voltage to produce an electric field proximate to said tip and direct said charged particles along a path toward said extraction electrode, with said plurality of spaced-apart electrode segments being connected to differing voltage supplies to provide said deflection field, with said deflection field having a magnitude associated therewith that is asymmetrically distributed about said optical axis. 20. The source as recited in claim 19 wherein said plurality of electrode segments are arranged in first and second sets of four electrode segments, with said first set being disposed between said extraction electrode and said emitter and said second set being disposed between said first set said emitter, with a first insulating layer being disposed between said first set of four electrode segments and said extraction electrode and a second insulating layer being disposed between said first and second sets of four electrode segments. |
description | The invention relates to security scans of intermodal containers. Items to be transported are often placed in intermodal containers, also referred to as cargo containers, which are included within the term “container” as used herein. Each intermodal container is a standardized packing case, typically having a length of 20 or 40 feet, although other standardized lengths are known. The 20-foot container is often used as a unit of measure, and containers are rated according to TEU, Twenty-foot Equivalent Units, for storage and transportation calculations. Container ships large enough to carry 6,000 TEU or more are known. There is a risk of unauthorized, potentially dangerous material being hidden in a container. While scanning equipment exists, it is challenging to adequately scan a large number of containers, especially at transfer facilities where space is at a premium and other vessels, vehicles or other transportation equipment are waiting for loading and unloading. It is therefore desirable to have an improved system and procedure for scanning many containers while minimizing impact on commerce. An object of the present invention is to enable thorough security scanning of multiple containers. Another object of the present invention is to utilize storage and/or transit periods for security scanning. This invention features a system of scanning a number of objects such as intermodal containers in a storage area having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one object. The racks are capable of being moved to establish successive vacant multi-level aisles among the racks. At least one scanner is movable substantially vertically within the successive vacant aisles to scan successive objects substantially adjacent to each vacant aisle to detect at least one pre-determined characteristic. In some embodiments, each rack includes a chassis movable along at least one track. Each rack is movable in at least one side direction. In certain embodiments, the scanner is suspended from a crane capable of lifting at least one intermodal container. In one embodiment, the system generates an alert when the scanner detects at least one pre-determined characteristic. This invention may also be expressed as a method of scanning a number of objects in a multi-level environment, including selecting a storage area having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one object. A scanner is selected, and a first vacant multi-level aisle is established among the racks. The method further includes moving the scanner substantially vertically within the first vacant aisle to scan successive objects substantially adjacent to the first vacant aisle to detect at least one pre-determined characteristic. At least a second vacant multi-level aisle is then established among the racks, and the scanner is moved substantially vertically within at least the second vacant aisle to scan additional objects substantially adjacent to the second vacant aisle. In some embodiments, each rack is configured to hold at least one intermodal container. In certain embodiments, the scanner is suspended from a crane capable of lifting at least one intermodal container. In one embodiment, the method further includes generating an alert when the scanner detects the at least one pre-determined characteristic. This invention may be accomplished by a system and method of scanning a number of objects such as crates or intermodal containers in a storage area having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one object. The racks are capable of being moved to establish successive vacant multi-level aisles among the racks. At least one scanner is movable substantially vertically within the successive vacant aisles to scan successive objects substantially adjacent to each vacant aisle to detect at least one pre-determined characteristic such as radiation from an unauthorized radioactive item within the object. The system and method according to the present invention are particularly useful with prior art cargo container storage and retrieval systems such as disclosed by the present inventor in U.S. Pat. Nos. 5,860,783 and 6,077,019, which are incorporated herein by reference in their entireties, and as represented by FIGS. 1 and 2 herein. In the illustrated embodiment, prior art storage and retrieval system 10, FIG. 1, includes a mobile gantry crane 12 that slides fore (forward) and aft (backwards) on crane tracks 13 to lower and raise containers 14. Gantry crane 12 includes a hoist 120 installed within housing 122 and having cables 124, 126, 128 and 130 shown removably attached to an intermodal container 132 with known implements. Hoist housing 122 is carried by a trolley 134 that travels along a beam 136 which is supported on legs 138 and 140. One prior art stacking plan for containers is illustrated schematically in FIG. 2 within a cargo hold 16, defined by borders 46, 48, 50 and 52, of an ocean transport vessel or other transportation equipment. A plurality of containers 14a are arranged in a number of levels vertically spaced one above another within vessel hold 16. Each level may also be referred to as a tier. A plurality of racks are arranged along each level. In this construction, each rack has a single chassis seat in one of tiers 18, 20, 22, 24 and 26. Each chassis seat can be positioned in one of columns 28, 30, 32, 34, 36, 38, 40, 42 and 44 within each tier. One chassis seat position 59, 60, 62, 64 and 66 remains vacant in each tier so that, after appropriate movement of containers as described in U.S. Pat. Nos. 5,860,783 and 6,077,019, a selected container within any tier can be retrieved as desired. As shown in FIG. 2, the vacant positions 59, 60, 62 and 64 have been established above container 133 so that it can be raised through hatch 81, which is one of a number of access hatches 80 defined in main deck 78. Scanning system 200 according to the present invention is illustrated in FIG. 3 as part of a schematic partial cross-sectional view looking aft through a hull H of a loaded cargo vessel V. In this view, a scanner 202 is shown in a raised condition suspended from a shipboard gantry crane 203 having a hoist 204 with cables 206 and 208. Hoist 204 is installed within a housing 210 that is carried by a trolley (not visible) that travels along a beam 212 in the side directions indicated by arrows 214 and 216. Gantry crane 203 includes legs 220 and 222 which are fixedly secured to main deck 224 in some constructions and, in other constructions, are movable fore and aft along rails or other features (not shown). In this construction, some of containers 227 are positioned in racks on levels 230, 232, 234 and 236 as an upper, above-deck group 228 and the remaining containers 227 are positioned on racks in levels 240, 242, 244, 246, 248, 250 and 252 as a lower, below-deck group 239 within hold HD of vessel V. The containers 227 are also positioned in columns 260, 262, 264, 266, 268, 270, 272, 274, 276, 278, 280, 282 and 284. Column position 282 has been cleared within each level to establish a vacant aisle VA in FIGS. 3 and 6, as described in more detail below. In general, the present invention includes a method of scanning a number of objects in a multi-level environment, including selecting a storage area having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one object. A scanner is selected, and a first vacant multi-level aisle is established among the racks. The method further includes moving the scanner substantially vertically within the first vacant aisle to scan successive objects substantially adjacent to the first vacant aisle to detect at least one pre-determined characteristic such as radiation from an unauthorized radioactive item within the object. At least a second vacant multi-level aisle is then established among the racks, and the scanner is moved substantially vertically within at least the second vacant aisle to scan additional objects substantially adjacent to the second vacant aisle. One benefit of the present invention is that the scanning of objects can be accomplished during normally inactive periods for the objects such as during ocean transport of intermodal containers between ports. Preferably, all containers or other selected objects are sufficiently scanned to locate unauthorized radioactive materials, explosives, drugs or other contraband prior to off-loading of the objects. FIG. 4 is a flow chart depicting sequential scanning according to the present invention of objects such as intermodal containers, beginning with step 300 during which a column to be scanned is selected. The method proceeds to decision step 302 in which the selected column to be scanned is analyzed to confirm that a vacant aisle has been established substantially adjacent to the selected column. If not, the logic proceeds to step 304 in which containers are shifted to establish the vacant aisle next to the selected column and, after step 302 is satisfied, a scanner is positioned, step 306, within the vacant aisle substantially adjacent to a container to be scanned. The scanner is activated, step 308, until the scan is complete as assessed in step 310. If a positive result is detected, step 312, an alert can be generated, step 314. Types of scanners suitable for the present invention include Geiger counters and other detectors of radioactivity, chemical detectors for explosives or illicit drugs, x-ray scanners, and other detectors that will occur to those of ordinary skill in the art after reviewing this disclosure. Types of scanners include both active emitters and passive detectors and sensors. A video camera or other observation equipment is included in some constructions. Possible actions to be taken after an alert is generated include sounding an alarm to inform an operator, marking the suspicious container with high-visibility spray paint or other identifying substance for later quarantine and/or more detailed inspection. For containers found to carry highly dangerous materials, those targeted containers can be individually removed from the storage area and handled appropriately, including safe deactivation, containment and/or disposal. If a positive result is not detected for that container, scanning continues along the vacant aisle until the entire selected column has been completed as assessed in step 316. Arrow 318 represents a return to positioning the scanner, step 306, at the next container to be scanned within the selected column. After every container has been scanned adjacent to the vacant aisle, the logic returns to step 300 in which another column is selected to be scanned. FIGS. 5-9 are schematic cross-sectional views similar to FIG. 3, with corresponding reference numerals, illustrating one procedure for successive scanning according to the present invention. FIG. 5 shows all containers 227 in both groups 228 and 239 shifted to the starboard side of vessel V, in the direction of arrow 340, to establish column 284 as a vacant aisle. Hoist 204 has been moved in direction 216 to the port side adjacent crane leg 222 to position scanner 202 over the vacant aisle. Scanner 202 is lowered by cables 206 and 208 from hoist 204 to scan containers in column 282 in a pre-selected sequence, typically top-to-bottom, beginning with container 337, or bottom-to-top, beginning with container 339. Wheels, rollers, fenders, skids or other protective items (not shown) can be added to the sides of the scanner 202 to cushion it during travel down the vacant aisle, direction 322, or up, direction 324. Scanner 202 is shown in FIGS. 5-9 with starboard-side detectors 328 and 330 and port-side detectors 332 and 334 in this construction. When a vacant aisle is established at the outer sides of the containers 227, such as shown in FIGS. 5 and 9, then the outer-most detectors can be deactivated or can be utilized as a “control” or reference. FIG. 6 shows a vacant aisle VA in the same position as illustrated in FIG. 3, that is, in column 282, to scan containers in columns 280 and 284 after the containers in column 282, FIG. 5, have been shifted in the direction of arrow 342. FIG. 7 shows containers 227 arranged to establish a vacant aisle in column 280 to scan containers in columns 278 and 282. In this construction, racks establishing columns of containers are successively shifted in the direction of arrow 342 until, as shown in FIG. 8, scanner 202 scans containers in columns 260 and 264 along a vacant aisle in column 262. FIG. 9 shows containers in column 262 being scanned adjacent to a vacant aisle in column 260. The system and method according to the present invention enable scanning of at least two sides of each selected container, which increases detection accuracy. When scanning a single side of each selected container is sufficient, then vacant aisles can be established every other column such as columns 262 (FIG. 8), 266, 270, 274, 278 and 282 (FIG. 6) while skipping columns 260 (FIG. 9), 264, 268, 272, 276, 280 (FIG. 7) and 284 (FIG. 5). While a system and method according to the present invention has been described above in connection with intermodal containers on a vessel, that is not a limitation of the invention. Containers, crates and other objects can be scanned according to the present invention in other transportation equipment in some constructions. In other constructions, objects are scanned according to the present invention in a stationary warehouse or other storage facility such as disclosed by the present inventor in U.S. Patent Publication 2012-0251277, which is incorporated by reference in its entirety. Although specific features of the present invention are shown in some drawings and not in others, this is for convenience only, as each feature may be combined with any or all of the other features in accordance with the invention. While there have been shown, described, and pointed out fundamental novel features of the invention as applied to one or more preferred embodiments thereof, it will be understood that various omissions, substitutions, and changes in the form and details of the devices illustrated, and in their operation, may be made by those skilled in the art without departing from the spirit and scope of the invention. For example, it is expressly intended that all combinations of those elements and/or steps that perform substantially the same function, in substantially the same way, to achieve the same results be within the scope of the invention. Substitutions of elements from one described embodiment to another are also fully intended and contemplated. It is also to be understood that the drawings are not necessarily drawn to scale, but that they are merely conceptual in nature. It is the intention, therefore, to be limited only as indicated by the scope of the claims appended hereto. Other embodiments will occur to those skilled in the art and are within the following claims. Every issued patent, pending patent application, publication, journal article, book or any other reference cited herein is each incorporated by reference in their entirety. |
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043137938 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows a nuclear reactor installation 10 including a nuclear reactor 12 which contains a plurality of nuclear fuel assemblies (not shown). A plurality of in-core instrument guide tubes 14 extend from the seal table 16, through guide pipes 17 permanently mounted in brackets 19, and into the reactor 12. A typical reactor 12 may contain fifty or more individual tubes 14. Each tube 14 contains a detector cable (not shown) which, together with the tube, are referred to as instruments (ICI) or assemblies. Each tube 14 provides a guide into an individual fuel assembly in the reactor 12. The detector cable within each tube is driven by a drive train shown generally at 18. At the seal table 16 the drive train is detachably connected to each guide tube 14 and detector cable contained therein whereby the detector may be moved axially through the reactor 12. It should be appreciated that the ICI guide tubes 14 are typically very stiff, being made of stainless steel and having an outer diameter of approximately 0.5 inches and a thickness of about 0.15 inch. The detector cable itself which moves within the tube 14, is more flexible and, for the purpose of the present invention, does not affect the characteristics of the tube 14. The tube 14 is typically about sixty feet in length. During each refueling, a fraction of the total number of ICI guide tubes 14 and the detectors contained therein must be removed from service and permanently disposed of. Early in the refueling procedure the refueling pool 20 is filled with water. The pool walls 22 typically support a refueling deck 24 which typically includes tracks for equipment to move in the horizontal plane. The overhead crane (not shown) is located high above the reactor installation 10 and as described above, was previously used to remove and dispose of the ICI tubes 14. The present invention is shown removing tube 14a from the reactor 12. The machine 26 is vertically located by a pulley system 30 and the winding action for reel 32 is provided by motor 28. The pulley 30 and motor 28 are located for horizontal movement on the track on the deck 24. It should be understood that other arrangements of ICI reactor penetrations have been used, such as the seal plate 16 being mounted directly above the reactor 12 for penetration through the top thereof. However, the depth of water in the refueling pool 20 is too shallow to permit the radioactive portion 14' of the ICI 14 to be removed from the the reactor and vertically suspended in the pool 20, without violating the radiation limits above the pool 20. The present invention permits complete removal of the ICI 14 from the reactor 12 and transfer for disposal while keeping all of the ICI sufficiently below the surface of the pool 20 to avoid exceeding radiation limits. In reactor installations where the ICI's are supported within the reactor upper guide structure (UGS) (not shown), the UGS may be removed from the reactor and placed on the fuel pool floor 37, thereby allowing access to the reactor for handling the fuel assemblies. The invention can then be used to remove each ICI from the UGS while fuel assemblies are being shuffled in the reactor. Referring now to FIGS. 2 and 3, the reel 32 is carried by a frame 33. In the illustrated embodiment, the reel 32 has a plurality of spokes 34 connected to a hub 36. A split tapered bushing 38 connects the hub 36 to the drive shaft 40, which is operated in a conventional manner by the drive motor 28 (see FIG. 1). A slot 42 is provided in the reel for capturing one end of the ICI tube 14. The detector cable 43 may be secured on a spoke 34. The outer surface of the reel 32 has helical grooves 44 formed in a substantially continuous path around the reel. The groove diameter is preferably equal to the outer diameter of the tube 14. In the illustrated embodiment the end of the ICI tube 14 is manually placed through the slot 42 and located in one of the grooves 44. It is noted that the captured portion of the ICI tube 14 (i.e., the portion closest to the seal table 16 shown in FIG. 1) is not highly radioactive and therefore the leading end can be manually inserted into the slot 42. A variety of means may be provided for capturing the tube 14 on the reel 32. To facilitate capture, the reel may be provided with means, such as the split taper bushing 38, the thrust bearing 54, and radial bearing 56, for selectively permitting the movement of the reel 32 away from the frame 33 and out of alignment with the cam rollers 52. This feature may or may not be necessary depending on the particular means chosen for capturing the tube 14. After the tube 14 is captured in the reel 32, the sixty foot length of stiff tubes 14 is wound onto the reel 32. This can be difficult because of the tendency of the tube to retain its original shape and therefore to come off the reel 32. According to the present invention, a plurality of cam rollers 52 are connected to the frame 33 and spaced about the reel 32 so that the tube 14 is held closely within the grooved path 44. As the reel 32 is turned counterclockwise in FIG. 2, the tube 14 winds into the groove in a smooth, uninterrupted manner. Since the rollers 52 force the tube 14 to enter the groove 44, there is a unique relationship between the number of revolutions of the reel 32 and the length of tube 14 that has been wound thereon. This may be used to advantage when the length of unwound tube must be known precisely. The frame 33 is conveniently chosen to be a square about five feet on each side and the reel 32 is approximately four feet in diameter. Twelve rubber or other non-metallic cam rollers 52 having a four inch outer diameter are sufficient to provide the required frictional force on the tube 14. The rollers 52 are mounted on the frame 33 for relatively free rotation but for maintaining a fixed relationship relative to the moving reel 32. The proper distance between the rollers 52 and the grooves 44 is dependent on the diameter of the tube 14, but need not be precisely determined. It has been found, for example, that for the illustrated embodiment the maximum perpendicular distance between the base of the groove 44 and the cam roller 52 should not exceed the nominal outer diameter of the tube 14 by more than 25 percent. To facilitate the feeding and, as will be shoen, the unwinding of the tube 14, a plurality of straightening rollers 58 are provided on the frame 33 for straightening the tube 14 as it moves onto or off of the reel 32. Referring also to FIG. 1, the alternate position of the machine is shown in phantom. After the tube 14 has been fully wound onto the reel 32, the frame 33 is moved along the track 24 to a cutting station 60 where the tube 14 is unwound from the reel 32 and fed into a cutting device 62. It can be appreciated that, but for the retaining force provided by the cam rollers 52, it would be very difficult to unwind the stiff tube 14 from the reel 32. The rollers 52 prevent the tube 14 from radially coming out of the groove 44 so that the circumferential driving of the reel 32 in the clockwise direction unwinds the tube 14 through the straightening rollers 58 and into the cutting device 62. In one known cutting arrangement, the tube 14 must be fed into a receptable 64 before the cutting operation takes place. As the number of segments in the receptacle increases, however, it becomes exceedingly difficult to feed the next portion of tube 14 into the receptacle. The present invention provides sufficient force to overcome this difficulty without the need for additional feeding equipment. In the preferred embodiment the reel 32 includes an outer portion 46 in the form of a cartridge detachably connected to the rim portion 60. Such connection is schematically represented at 48. With this embodiment of the invention, the entire ICI 14 is wound onto the reel 32, but only the radioactive portion 14' (see FIG. 1) is unwound into the cutting machine. After the radioactive portion has been cut into segments and placed in the receptacle, the remaining nonradioactive portion (approximately two-thirds of the ICI) is still on the reel. The operator then simply slides the cartridge portion 46, with the tube 14 intact, from the reel 32 and then snaps a cover (not shown) into place over the cartridge. The cartridge with cover thus forms a container that can be safely and conveniently removed from the reactor containment and stored in an area for low level radiation waste. The preferred embodiment has been described, wherein the ICI consists of a stiff tube containing a flexible detector. It should be understood that the invention also has similar advantages when used with other kinds of ICI, such as the Penflex sheath which is not as stiff as the tube type. The Penflex ICI are more easily coiled so that a smaller reel may be employed on a smaller frame. It should be appreciated that the present invention not only eliminates the need to use the overhead crane for ICI tube removal and disposal purposes, but also reduces the time, and the radiation exposure to the operators, relative to conventional procedures. The winding operation can be made at a safe depth of water because the ICI tube 14 assumes a compact coil shape and is not strung out and dangled from the crane. The frame 33 may be adapted to include mounts for placement of lead shields around the reel 32 so that the radiation exposure to operating personnel can be further reduced. Since the operation of the inventive machine is under much tighter control than when the overhead crane is used, it will be possible to remove and dispose of the ICI tubes 14 at a faster rate. |
claims | 1. Apparatus comprising:a source adapted to produce a target of a material in a liquid state;a laser adapted to irradiate said target to change a state of the material from said liquid state to a plasma state to produce EUV light in an irradiation region;an optical system adapted to convey said EUV light from said irradiation region to a workpiece;said source comprisinga target generator anda target generator steering system mechanically coupled to the target generator, the target generator steering system including a first member adapted to be fixed relative to said irradiation region, and a second member adapted to receive the target generator and adapted to be movable with respect to said irradiation region; anda coupling system mechanically coupling the first member to the second member, wherein the coupling system comprises at least one flexure, said first member being substantially plate-shaped, and wherein said second member is substantially plate shaped and substantially parallel to said first member in a neutral position. 2. Apparatus as claimed in claim 1 wherein the coupling system comprises a first coupling subsystem configured to constrain parallel movement of the second member with respect to the first member while permitting tilting movement of the second member with respect to the first member, and a second coupling subsystem adapted to control the tilting of the second member with respect to the first member. 3. Apparatus as claimed in claim 2 wherein the first coupling subsystem comprises a plurality of first coupling subsystem elements mechanically coupling said first member to said second member. 4. Apparatus as claimed in claim 2 wherein the second coupling subsystem comprises a plurality of second coupling subsystem elements mechanically coupling said first member to said second member. 5. Apparatus as claimed in claim 4 wherein each of said second coupling subsystem elements comprises at least one first flexure. 6. Apparatus as claimed in claim 5 wherein said at least one first flexure is a cartwheel flexure. 7. Apparatus as claimed in claim 5 wherein each of said second coupling subsystem elements comprises at least one second flexure coupled to said at least one first flexure. 8. Apparatus as claimed in claim 7 wherein said at least one second flexure is a parallelogram flexure. 9. Apparatus as claimed in claim 5 wherein each of second coupling subsystem elements comprises at least one linear motor coupled to said first member and to said first flexure. 10. Apparatus as claimed in claim 9 wherein each of said second coupling subsystem elements comprises at least one second flexure coupled to said at least one first flexure. 11. Apparatus as claimed in claim 10 wherein each of second coupling subsystem elements comprises at least piezoelectric element coupled to said second member and to said at least one second flexure. 12. Apparatus as claimed in claim 3 wherein the second coupling subsystem comprises a plurality of second coupling subsystem elements mechanically coupling said first member to said second member. 13. Apparatus as claimed in claim 12 wherein each of said second coupling subsystem elements comprises at least one first flexure. 14. Apparatus as claimed in claim 13 wherein said at least one first flexure is a cartwheel flexure. 15. Apparatus as claimed in claim 13 wherein each of said second coupling subsystem elements comprises at least one second flexure coupled to said at least one first flexure. 16. Apparatus as claimed in claim 13 wherein each of second coupling subsystem elements comprises at least one linear motor coupled to said first member and to said first flexure. 17. Apparatus as claimed in claim 13 wherein each of said second coupling subsystem elements comprises at least one second flexure coupled to said at least one first flexure. 18. Apparatus as claimed in claim 12 wherein each of second coupling subsystem elements comprises at least piezoelectric element coupled to said second member and to said at least one second flexure. 19. Apparatus comprising:a source adapted to produce a target of a material in a liquid state;a laser adapted to irradiate said target to change a state of the material from said liquid state to a plasma state to produce EUV light in an irradiation region;an optical system adapted to convey said EUV light from said irradiation region to a workpiece;said source comprisinga target generator anda target generator steering system mechanically coupled to the target generator, the target generator steering system including a first member adapted to be fixed relative to said irradiation region, and a second member adapted to receive the target generator and adapted to be movable with respect to said irradiation region; anda coupling system mechanically coupling the first member to the second member, wherein the coupling system comprises at least one flexure, said first member being substantially plate-shaped, and wherein said second member is substantially plate shaped and substantially parallel to said first member in a neutral position, the second coupling subsystem comprising a plurality of second coupling subsystem elements mechanically coupling said first member to said second member, each of said second coupling subsystem elements comprising at least one first flexure. 20. Apparatus comprising:a source adapted to produce a target of a material in a liquid state;a laser adapted to irradiate said target to change a state of the material from said liquid state to a plasma state to produce EUV light in an irradiation region;an optical system adapted to convey said EUV light from said irradiation region to a workpiece;said source comprisinga target generator anda target generator steering system mechanically coupled to the target generator, the target generator steering system including a first member adapted to be fixed relative to said irradiation region, and a second member adapted to receive the target generator and adapted to be movable with respect to said irradiation region; anda coupling system mechanically coupling the first member to the second member, wherein the coupling system comprises at least one flexure, said first member being substantially plate-shaped, and wherein said second member is substantially plate shaped and substantially parallel to said first member in a neutral position, the second coupling subsystem comprises a plurality of second coupling subsystem elements mechanically coupling said first member to said second member and each of said second coupling subsystem elements comprising at least one first flexure. |
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claims | 1. An ion implantation method of implanting ions into a substrate using both a ribbon-like ion beam in which, with or without performing an X direction sweep, a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, and a mechanical scan of the substrate in a direction intersecting with a principal face of the ion beam, said method comprising:calculating a scan number of the substrate as an integer value in which digits after a decimal point are truncated by using a beam current of the ion beam, a dose amount to the substrate, and a reference scan speed which is used as a reference for calculating a scan number of the substrate;determining whether the calculated scan number is 2 or more or not; if the scan number is smaller than 2, aborting a process of obtaining a practical scan number and a practical scan speed; if the scan number is equal to or larger than 2, determining whether the calculated scan number is even or odd; if the scan number is even, setting the current scan number as the practical scan number; and, if the scan number is odd, obtaining an even scan number which is smaller by 1 than the odd scan number, and setting the obtained even scan number as the practical scan number;calculating the practical scan speed of the substrate by using the practical scan number, the beam current, and the dose amount; andperforming ion implantation on the substrate in accordance with the practical scan number and the practical scan speed. 2. An ion implantation apparatus for implanting ions into a substrate using both a ribbon-like ion beam in which, with or without performing an X direction sweep, a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, and a mechanical scan of the substrate in a direction intersecting with a principal face of the ion beam, said apparatus comprising:a controlling device having functions of: (a) calculating a scan number of the substrate, as an integer value in which digits after a decimal point are truncated by using a beam current of the ion beam, a dose amount to the substrate, and a reference scan speed which is used as a reference for calculating a scan number of the substrate; (b) determining whether the calculated scan number is 2 or more or not; if the scan number is smaller than 2, aborting a process of obtaining a practical scan number and a practical scan speed; if the scan number is equal to or larger than 2, determining whether the calculated scan number is even or odd; if the scan number is even, setting the current scan number as the practical scan number; and, if the scan number is odd, obtaining an even scan number which is smaller by 1 than the odd scan number, and setting the obtained even scan number as the practical scan number; (c) calculating the practical scan speed of the substrate by using the practical scan number, the beam current, and the dose amount; and (d) performing ion implantation on the substrate in accordance with the practical scan number and the practical scan speed. 3. An ion implantation method of implanting ions into a substrate using both a ribbon-like ion beam in which, with or without performing an X direction sweep, a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, a mechanical scan of the substrate in a direction intersecting with a principal face of the ion beam, and performance of ion implantation while, during a period when the ion beam does not impinge on the substrate, rotating the substrate by a step of 360/m deg. about a center portion of the substrate, and dividing one rotation of the substrate into a plurality m of implanting steps, said method comprising:calculating a scan number per implanting step of the substrate as an integer value in which digits after a decimal point are truncated by using a beam current of the ion beam, a dose amount to the substrate, a implanting step number, and a reference scan speed which is used as a reference for calculating a scan number per implanting step of the substrate;determining whether the calculated scan number per implanting step is 1 or more or not; if the scan number is smaller than 1, aborting a process of obtaining a practical scan number per implanting step and a practical scan speed; if the scan number is equal to or larger than 1, determining whether the calculated scan number per implanting step is even or odd; if the scan number is even, setting the current scan number as the practical scan number per implanting step; if the scan number is odd, determining whether the scan number is 1 or not; if the scan number is 1, setting the current scan number as a practical scan number per implanting step; and, if the scan number is not 1, obtaining an even scan number which is smaller by 1 than the odd scan number, and setting the obtained even scan number is set as the practical scan number per implanting step;calculating the practical scan speed of the substrate by using the practical scan number per implanting step, the beam current, the dose amount, and the implanting step number; andperforming ion implantation on the substrate in accordance with the practical scan number per implanting step and the practical scan speed. 4. An ion implantation apparatus for implanting ions into a substrate using all of a ribbon-like ion beam in which a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, a mechanical scan of the substrate in a direction intersecting with a principal face of the ion beam, and performance of ion implantation while, during a period when the ion beam does not impinge on the substrate, rotating the substrate by a step of 360/m deg. about a center portion of the substrate, and dividing one rotation of the substrate into a plurality m of implanting steps, said apparatus comprising:a controlling device having functions of: (a) calculating a scan number per implanting step of the substrate, as an integer value in which digits after a decimal point are truncated by using a beam current of the ion beam, a dose amount to the substrate, a number of the implanting steps, and a reference scan speed which is used as a reference for calculating a scan number per implanting step of the substrate; (b) determining whether the calculated scan number per implanting step is 1 or more or not; if the scan number is smaller than 1, aborting a process of obtaining a practical scan number per implanting step and a practical scan speed; if the scan number is equal to or larger than 1, determining whether the calculated scan number per implanting step is even or odd; if the scan number is even, setting the current scan number, as the practical scan number per implanting step; if the scan number is odd, determining whether the scan number is 1 or not; if the scan number is 1, setting the current scan number, as the practical scan number per implanting step; and, if the scan number is not 1, obtaining an even scan number which is smaller by 1 than the odd scan number, and setting the obtained even scan number as the practical scan number per implanting step; (c) calculating a practical scan speed of the substrate by using the practical scan number per implanting step, the beam current, the dose amount, and the implanting step number; and (d) performing ion implantation on the substrate in accordance with the practical scan number per implanting step and the practical scan speed. |
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abstract | Embodiments of methods of modifying surface features on a workpiece with a gas cluster ion beam are generally described herein. Other embodiments may be described and claimed. |
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044255080 | abstract | In the electron beam lithographic system disclosed herein, a semiconductor wafer to be exposed is carried on an air bearing puck which is, on both sides, supported or located by balanced annular regions of air pressure. These annular supporting regions surround central evacuated regions which are also balanced so that the puck is not subject to large bending forces. Accordingly, the puck can be constructed to light-weight materials facilitating rapid and precise positioning of the semiconducter wafer with respect to an E-beam generating column. |
claims | 1. A storage module for vertically-stacked storage of nuclear waste canisters comprising:an elongated outer shell defining a vertical axis and an internal cavity;a removable top lid mounted on a top end of the outer shell covering the cavity;a bottom plate mounted on a bottom end of the outer shell;a first canister positioned in the cavity in a lower position, the first canister including a top, a bottom, and a sidewall extending between the top and bottom;a second canister vertically stacked above the first canister in an upper position and including a top, a bottom, and a sidewall extending between the top and bottom, the first and second canisters being concentrically aligned with the vertical axis;a middle ring-lug assembly interspersed between and engaging the first and second canisters;the middle ring-lug assembly comprising a ring and plurality of vertically elongated centering lugs spaced circumferentially apart around a perimeter of the ring;wherein the lugs of the middle ring-lug assembly are each configured to laterally support and restrain the first and second canisters against the module;wherein the lugs of middle ring-lug assembly each protrude laterally outwards beyond the ring and the sidewalls of the first and second canisters to engage an interior surface of the module; andwherein lugs of the middle ring-lug assembly each extend upwards above the ring and downwardly below the ring. 2. The module of claim 1, wherein the middle ring-lug assembly is configured and operable to transfer weight of the second canister to the first canister. 3. The module of claim 1, wherein the ring of the middle ring-lug assembly includes an upward facing upper castellated surface engaging the bottom of the first canister. 4. The module of claim 3, wherein the upper castellated surface includes a plurality of alternating arcuate raised segments and arcuate recessed segments. 5. The module of claim 4, wherein the raised segments engage the bottom of the first canister and the recessed segments are spaced apart from the bottom of the first canister. 6. The module of claim 4, wherein the ring of the middle ring-lug assembly includes a downward facing lower castellated surface engaging the bottom of the first canister. 7. The module of claim 6, wherein the lower castellated surface includes a plurality of alternating arcuate raised segments and arcuate recessed segments. 8. The module of claim 6, wherein the ring has a composite structure formed of an upper ring and a lower ring attached together in back-to-back relationship with flat bottoms of the upper and lower rings in mutual abutting contact. 9. The module of claim 8, wherein the upper ring is welded to the lower ring. 10. The module of claim 1, wherein the lugs of the middle ring-lug assembly each have a flattened plate-like body which is vertically oriented on the ring. 11. The module of claim 10, wherein the lugs of the middle ring-lug assembly each include an angled inner edge above the ring arranged to engage and center the bottom of the first canister when lowered into the storage module on top of the second canister. 12. The module of claim 11, wherein each lug further includes a vertically oriented axial inner edge adjoining the angle inner edge. 13. The module of claim 12, wherein each lug further includes an angled inner edge below the ring arranged to engage and center the ring on the top of the second canister when the ring is lowered into the storage module on top of the second canister. 14. The module according to claim 1, wherein each lug of the middle ring-lug assembly extends vertically upwards above the bottom of the first canister to engage and restrain its sidewall against lateral movement, and vertically downward below the top of the second canister to engage and restrain its sidewall against lateral movement. 15. The module of claim 1, further comprising a top ring-lug assembly engaging the top of the first canister beneath the lid of the module, the top ring-lug assembly comprising a ring and plurality of vertically elongated centering lugs spaced circumferentially apart around a perimeter of the ring and protruding radially outwards from the ring, wherein the lugs of the top ring-lug assembly are each configured to laterally support and restrain the first canister against the module. 16. The module of claim 15, wherein the lugs of the top ring-lug assembly extend downwards below the ring to engage a sidewall of the first canister. 17. The module of claim 15, further comprising a bottom ring-lug assembly engaging the bottom of the second canister seated on the bottom plate of the module, the bottom ring-lug assembly comprising a ring and plurality of vertically elongated centering lugs spaced circumferentially apart around a perimeter of the ring and protruding radially outwards from the ring, wherein the lugs of the top ring-lug assembly are each configured to laterally support and restrain the second canister against the module. 18. The module of claim 17, wherein the lugs of the bottom ring-lug assembly extend upwards above the ring to engage a sidewall of the second canister. 19. The system of claim 15, wherein the rings of the top and bottom ring-lug assemblies each have a castellated surface. 20. A storage module for vertically-stacked storage of nuclear waste canisters comprising:an elongated outer shell defining a vertical axis and an internal cavity;an elongated inner shell disposed in the cavity and concentrically aligned with the outer shell;a first canister positioned in the cavity in a lower position, the first canister including a top, a bottom, and a sidewall extending between the top and bottom;a second canister vertically stacked above the first canister in an upper position and including a top, a bottom, and a sidewall extending between the top and bottom, the first and second canisters being concentrically aligned with the vertical axis;a middle ring-lug assembly interspersed between and engaging the bottom of the first and the top of the second canister; athe middle ring-lug assembly comprising a ring and plurality of vertically elongated centering lugs spaced circumferentially apart around a perimeter of the ring, each lug protruding radially outwards from ring and configured to engage the inner shell to laterally restrain and support the first and second canisters;wherein an annular space is maintained by the middle ring-lug assembly between the first and second canisters and the inner shell;wherein the ring of the middle ring-lug assembly includes an upward facing upper castellated surface engaging the bottom of the first canister, and a downward facing lower castellated surface engaging the top of the second canister. 21. The module according to claim 20, wherein each lug of the middle ring-lug assembly has a flat plate-like body including a first portion which extends vertically upwards above the bottom of the first canister to engage and restrain its sidewall against lateral movement, and a second portion which extends vertically downward below the top of the second canister to engage and restrain its sidewall against lateral movement. |
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claims | 1. An apparatus comprising:a nuclear reactor core comprising fissile material;a core basket; anda core former surrounding the nuclear reactor core, the core former comprising a stack of two or more single-piece annular rings wherein each single-piece annular ring comprises neutron-reflecting material,wherein adjacent single-piece annular rings of the stack of single-piece annular rings include a key feature disposed on a first one of the adjacent single-piece annular rings and a keyway feature disposed on a second one of the adjacent single-piece annular rings, wherein the key feature is received in the keyway feature to prevent relative rotation of the first one and the second one of the adjacent single-piece annular rings, and wherein the stack of single-piece annular rings does not include welds or fasteners that axially restrain adjacent single-piece annular rings together or that secure the rings to the core basket. 2. The apparatus of claim 1, wherein the single-piece annular rings are forged or cast annular rings. 3. The apparatus of claim 2, wherein the single-piece annular rings are stainless steel annular rings. 4. The apparatus of claim 1, wherein the single-piece annular rings are stainless steel annular rings. 5. The apparatus of claim 1, wherein adjacent single-piece annular rings of the stack have mating surfaces shaped to define an annular joint between the adjacent single-piece annular rings that provides a tortuous path for fluid flow into or out of the core former via the joints. 6. The apparatus of claim 1, wherein the core basket contains the nuclear reactor core and the core former. 7. The apparatus of claim 6, wherein an annular gap is defined between the core former and the core basket. 8. The apparatus of claim 6, wherein the outer surface of the core former includes axially extending channels. 9. The apparatus as set forth in claim 6, further comprising:a pressure vessel containing the nuclear reactor core, the core former, and the core basket. 10. The apparatus as set forth in claim 9, wherein the core basket is suspended from a mid-flange of the pressure vessel. 11. The apparatus of claim 1, wherein each single-piece annular ring is a monolithic element without joints or seams. 12. The apparatus of claim 1, wherein the core former does not include welds and does not include fasteners. 13. An apparatus comprising:a core basket; anda core former sized for a nuclear reactor core, the core former comprising a stack of two or more single-piece annular rings wherein each single-piece annular ring comprises neutron-reflecting material and has an inner surface conforming with a periphery of the nuclear reactor core,wherein a first one of the single-piece annular rings includes a circumferential recess in one of its top surface and its bottom surface, and a second one of the single-piece annular rings includes a circumferential protrusion in one of its top surface and its bottom surface, and the circumferential protrusion is received within the circumferential recess,wherein the stack of single-piece annular rings does not include welds or fasteners axially restraining adjacent single-piece annular rings together or that secure the rings to the core basket, andwherein the first one of the single-piece annular rings includes a key feature disposed on one of its top surface and its bottom surface and the second one of the adjacent single-piece annular rings includes a keyway feature disposed on one of its top surface and its bottom surface, and the key feature is received in the keyway feature to prevent relative rotation of the first one and the second one of the adjacent single-piece annular rings. 14. The apparatus of claim 13, wherein the core basket contains the nuclear reactor core and the core former. 15. The apparatus of claim 13, further comprising:a pressure vessel containing the nuclear reactor core, the core former, and the core basket. 16. The apparatus of claim 13, wherein the outer surface of the core former includes axially extending channels. |
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046408134 | description | DETAILED DESCRIPTION OF THE INVENTI0N In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also, in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. In General Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a fuel assembly, represented in vertically foreshortened form and being generally designated by the numeral 10. The fuel assembly 10 is the type used in a pressurized water reactor (PWR) and basically includes a lower end structure or bottom nozzle 12 for supporting the assembly on the lower core plate (not shown) in the core region of a reactor (not shown), and a number of longitudinally extending guide tubes or thimbles 14 which project upwardly from the bottom nozzle 12. The assembly 10 further includes a plurality of transverse grids 16 axially spaced along the guide thimbles 14 and an organized array of elongated fuel rods 18 transversely spaced and supported by the grids 16. Also, the assembly 10 has an instrumentation tube 20 located in the center thereof and an upper end structure or top nozzle 22 attached to the upper ends of the guide thimbles 14. With such an arrangement of parts, the fuel assembly 10 forms an integral unit capable of being conveniently handled without damaging the assembly parts. As mentioned above, the fuel rods 18 in the array thereof in the assembly 10 are held in spaced relationship with one another by the grids 16 spaced along the fuel assembly length. Each fuel rod 18 includes nuclear fuel pellets 24 and the opposite ends of the rod are closed by upper and lower end plugs 26,28. The fuel pellets 24 composed of fissile material are responsible for creating the reactive power of the PWR. A liquid moderator/coolant such as water, or water containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract some of the heat generated therein for the production of useful work. Improved Burnable Absorber Rod In the operation of a PWR it is desirable to prolong the life of the reactor core as long as feasible to better utilize the uranium fuel and thereby reduce fuel costs. To attain this objective, it is common practice to provide an excess of reactivity initially in the reactor core and, at the same time, provide means to maintain the reactivity relatively constant over its lifetime. The present invention relates to such means in the form of an improved soluble burnable poison or neutron absorber rod 30, as seen in FIG. 1, inserted into one of the guide thimbles 14. At least one, and preferably several of the rods 30 are stationarily supported by a spider 32 in the guide thimbles 14 of some of the fuel assemblies 10 to assist the movable control rods in the guide thimbles of other fuel assembles (not shown) in maintaining a substantially constant level of neutron flux or reactivity in the core throughout its operating cycle. Referring to FIG. 2, the improved burnable absorber rod 30 basically includes an elongated hollow tubular member 34 having upper and lower opposite ends 36,38 and a hermetically sealed chamber 40 defined within the tubular member between its ends. A neutron absorber material 42 in liquid form is contained in the sealed chamber 40 of the tubular member 34. At the upper end 36 of the tubular member 34 is means 44 providing a hydrogen getter and being disposed in communication with the sealed chamber 40 for a purpose to be described later below. At the other, lower end 38 of the tubular member 34 is means 46 providing a hydride sink and also being disposed in communication with the sealed chamber 40 for a purpose also to be described hereafter. The tubular member 34 is formed by a tubular body 48 and a pair of upper and lower end plugs 50,52, which are composed of any suitable material, but preferably a zirconium-based alloy such as Zircaloy-4. The plugs 50,52 are rigidly attached, such as by girth welds 54,56, to the opposite ends of the tubular body 48 so as to hermetically seal the same. The liquid neutron absorber material 42 contained in the chamber 40 is preferably composed of boron dissolved in water (boric acid) with the boron enriched with B-10 over and above that proportion naturally present in boron. The use of enriched boron material in the improved rod 30 addresses and resolves one of the concerns with the use of boron: its solubility limit in water at room temperature. Its solubility limit is approximately 15,000 ppm at 100 degrees F. (at 325 degrees F. the limit is 125,000 ppm). To achieve its desired function, the rod 30 would require substantially more boron loading than 15,000 ppm. For the prior art fixed type absorber rod, the loading would be about 79,000 ppm; however, the improved absorber rod 30 would require only about 64,000 ppm of natural boron (or 12,800 ppm of B-10). But this is still higher than the low temperature solubility limit. Therefore, the boron would need to be enriched in B-10 as compared to natural boron which contains about 20% B-10 (the isotope of interest). A boron concentration using 85% enriched boron would be about 15,000 ppm natural boron (12,800 ppm B-10) which would stay in solution at the lower temperature. In normal operation of the core, temperature gradients are formed along the length of the rod 30, with temperatures being cooler at the lower portion of the rod and hotter at the upper portion of the rod. The liquid absorber material 42 therefore spontaneously circulates within the sealed chamber 40 due to the presence of these gradients. Such circulation tends to self-equalize the rate of depletion of the neutron absorber material throughout the length of the chamber 40. The tubular body 48 of the rod 30 has one or more reinforcing convolutions formed therein which enhance the structural rigidity and integrity of the rod so as to enable it to better withstand both high internal and external pressures acting on it. The reinforcing convolutions can take on the shape of a recess or groove 58 formed in the body 48 so as to extend along a spiraling path between the ends of the tubular member 34. Or, alternatively, the convolutions can be a series of ring-like circular grooves 60 formed in the body so as to extend circumferentially about and be spaced axially along the tubular member 34 between the ends thereof. Still further, some combination of both can be used. The use of spiral and/or circular convolutions 58,60 formed in the body 48 addresses and resolves another concern with the use of boron in the rod 30: how to handle the release of helium gas within the chamber 40 which results from the boron interaction with a neutron. Parenthetically, it should be mentioned that the liquid boron solution does not fill the entire chamber 40 during non-operating periods, but only up to the level indicated by the numeral 62. This is so because the specific volume of the solution will change with temperature. In going from 100 degrees F. to 600 degrees F, water increases in volume by 45%. Since the rod 30 is designed to have absorber material at full core height during operation, the absorber column in the rod must be reduced during non-operating periods (refueling, etc.). Thus, there is a vapor space 64 present in the chamber 40. Returning now to the concern with the release of helium gas and its solution by the use of convolutions, it should be noted that the helium released increases the end of life (EOL) pressure in the rod 30. The operating vapor pressure within the chamber 40 due to the water being at 600 degrees F. is about 1550 psi. Since the EOL internal pressure must be limited to approximately the reactor coolant pressure (2250 psi) to avoid outward creep of the tubular body (or cladding) 48, the final pressure of the released helium and backfill gas should be about 700 psi to equalize internal with external pressure. Thus, the beginning of life (BOL) backfill pressure must be reduced to around 300 psi. The reduced backfill pressure results in a fairly high pressure acting on the body 48 at BOL which can cause creep collapse of the rod 30. This dilemna is resolved by providing the above-mentioned convolutions, spiral groove 58 and/or circular grooves 60, which increase the collapse strength of the tubular body 48 due to BOL external pressure. If the circular grooves 60 are used, they would preferably be spaced about every one inch increment on the rod 30. Calculations show that a 0.039 inch deep convolution will double the collapse pressure of a 0.45 inch diameter x 0.024 inch wall tube. A 0.059 inch deep convolution will triple the collapse pressure. In summary, therefore, the convolutions 58 and 60 add sufficient strength to the thin wall of the tubular body 48 to withstand the external pressure at BOL. To reduce corrosion of the inside of the tubular member 34, its material is beta quenched. This also reduces the hydride pickup in the member due to free hydrogen from the oxidation process. The solid lower end plug 52 also serves as the means 46 providing the hydride sink. Since this end of the member 34, being the lower one, is the cooler, the hydrogen will tend to migrate toward it. The upper end plug 50 is an attachment fitting for connecting the rod 30 to the spider 32, and adjacent to it is positioned the hydrogen getter means 44. The means 44 takes the form of a Zircaloy sponge 66 adapted to remove hydrogen from the vapor space 64 in the chamber 40 of the member 34. The sponge 66 is retained adjacent the upper end plug 50 by an annular disc 68 which has a central opening 70 for allowing passage of the hydrogen gas to the sponge 66. The disc 68 is held against the sponge 66 by a circumferential bulge 72 formed in the body 48 of the tubular member 34. The lower end plug 52 of the tubular member 34 has a reduced diameter end portion 74 which adapts it to fit within a dashpot (not shown) at the bottomn of the guide thimble 14. The dashpot functions as a shock absorber when the rods are inserted during a scram operation. The lower end plug 52 also has a fill passage 76 extending axially through it which is used to prepressurize the rod 30 with helium and then it is closed by weld 78. It is thought that the improved soluble burnable absorber rod of the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
description | This application claims the benefit of U.S. Provisional Application No. 62/097,235, filed Dec. 29, 2014, which application is hereby incorporated by reference in its entirety. Irradiation of chemical compounds and separating materials from the irradiated compound have a range of technical applications which includes the production of radioisotopes, nuclear fuel processing, and fundamental scientific research. For example, the following is a table that lists some valuable radioisotopes including those with medical applications. Some of the radioisotopes are generated directly from precursor fissionable material while others are decay products of other radioisotopes. TABLE 1Radioisotopes and UsesRadioisotopeSymbolHalf LifeUse(s)Actinium-227227Ac21.8yAs a parent of 223Ra, used to create a 223Ragenerator (such an isotope generator may alsosometimes be referred to as a 227Ac “cow” thatis occasionally “milked” to obtain the 223Raisotope).Bismuth-213213Bi46minUsed for targeted alpha therapy (TAT),especially cancers, as it has a high energy (8.4MeV).Cesium - variousxxCsUsed in brachytherapy, particularly 133Cs andisotopes131Cs.Carbon-1111C20.3mPositron emitter used in positron emissiontomography (PET) for studying brainphysiology and pathology, in particular forlocalizing epileptic focus, and in dementia,psychiatry and neuropharmacology studies.Also has a role in cardiology.Chromium-5151Cr27.7dUsed to label red blood cells and quantifygastro-intestinal protein loss.Cobalt-5757Co271.8dUsed as a marker to estimate organ size andfor in-vitro diagnostic kits.Cobalt-6060Co5.271yFormerly used for external beam radiotherapy.Copper-6464Cu12.7hUsed to study genetic diseases affectingcopper metabolism, such as Wilson's andMenke's diseases, and for PET imaging oftumors, and therapy.Copper-6767Cu2.6dBeta emitter, used in therapy.Dysprosium-165165Dy2.33hUsed as an aggregated hydroxide forsynovectomy treatment of arthritis.Erbium-169169Er9.4dUsed for relieving arthritis pain in synovialjoints.Fluorine-1818F1.83hPositron emitter used in PET for studyingbrain physiology and pathology, in particularfor localizing epileptic focus, and in dementia,psychiatry and neuropharmacology studies.Also has a role in cardiology. 18F has becomevery important in detection of cancers and themonitoring of progress in their treatment,using PET.Gallium-6767Ga78hUsed for tumor imaging and localization ofinflammatory lesions (infections).Gallium-6868Ga68minPositron emitter used in PET and PET-CTunits. A daughter of 68Ge typically obtainedfrom a 68Ge generator.Germanium-6868Ge271dUsed as the ‘parent’ in a generator to producethe daughter isotope 68Ga.Gold, variousxxAuUsed in brachytherapy. 198Au in particular isisotopesused for treatment of the prostate.Holmium-166166Ho26hBeing developed for diagnosis and treatmentof liver tumors.Indium-111111In2.8dUsed for specialized diagnostic studies, e.g.brain studies, infection and colon transitstudies.Iodine-123123I13.2hIncreasingly used for diagnosis of thyroidfunction, it is a gamma emitter from electroncapture. This isotope does not have the betadecay that occurs in 131I.Iodine-124124I4.18dUsed as a tracer.Iodine-125125I59.4dUsed in cancer brachytherapy (prostate andbrain), also diagnostically to evaluate thefiltration rate of kidneys and to diagnose deepvein thrombosis in the leg. It is also widelyused in radioimmuno-assays to show thepresence of hormones in tiny quantities.Iodine-131131I8.02dWidely used in treating thyroid cancer and inimaging the thyroid; also in diagnosis ofabnormal liver function, renal (kidney) bloodflow and urinary tract obstruction. A stronggamma emitter, but used for beta therapy.Iridium-192192Ir74dSupplied in wire form for use as an internalradiotherapy source for cancer treatment (usedthen removed). Beta emitter.Iron-5959Fe46dUsed in studies of iron metabolism in thespleen.Krypton-81m81mKr13.1sec81mKr gas can yield functional images ofpulmonary ventilation, e.g. in asthmaticpatients, and for the early diagnosis of lungdiseases and function.Lead-212212Pb10.6hUsed in TAT for cancers or alpharadioimmunotherapy, with decay products212Bi and 212Po delivering the alpha particles.Lutetium-177177Lu6.7dIncreasingly important as it emits just enoughgamma for imaging while the beta radiationdoes the therapy on small (e.g. endocrine)tumors. Its half-life is long enough to allowsophisticated preparation for use.Molybdenum-9999Mo66hUsed as the ‘parent’ in a generator (e.g., a 99Mocow) to produce 99mTc.Nitrogen-1313N9.97mPositron emitter used in PET for studyingbrain physiology and pathology, in particularfor localizing epileptic focus, and in dementia,psychiatry and neuropharmacology studies.Also has a role in cardiology.Neptunium-238238N2.11dCan be obtained from neutron bombardment of237N, a parent of 238Pu.Oxygen-1515O122.2sPositron emitter used in PET for studyingbrain physiology and pathology, in particularfor localizing epileptic focus, and in dementia,psychiatry and neuropharmacology studies.Also has a role in cardiology.Palladium-103103Pd17dUsed to make brachytherapy permanentimplant seeds for early stage prostate cancer.Phosphorus-3232P14dUsed in the treatment of polycythemia vera(excess red blood cells) and laboratoryexperiments. Beta emitter.Plutonium-238238Pu87.7yUsed as a source in radioisotopethermoelectric generators.Potassium-4242K12hUsed for the determination of exchangeablepotassium in coronary blood flow.Radium-227227Ra42mParent of 223Ra.Radium-223223Ra11.4dUsed for treating pain associated withmultifocal bone metastases. Decay product of227Ra via 227Ac and 227Th.Rhenium-186186Re3.8dUsed for pain relief in bone cancer. Betaemitter with weak gamma for imaging.Rhenium-188188Re17hUsed to beta irradiate coronary arteries froman angioplasty balloon.Rubidium-8181Rb4.6hParent of 81mKr.Rubidium-8282Rb1.26minConvenient PET agent in myocardial perfusionimaging.Ruthenium -xxRuUsed in brachytherapy.various isotopesSamarium-153153Sm47hVery effective in relieving the pain ofsecondary cancers lodged in the bone, sold asQuadramet ™. Also very effective for prostateand breast cancer. Beta emitter.Selenium-7575Se120dUsed in the form of seleno-methionine to studythe production of digestive enzymes.Sodium-2424Na15hFor studies of electrolytes within the body.Strontium-8282Sr25dUsed as the ‘parent’ in a generator to produce82Rb.Strontium-8989Sr50dVery effective in reducing the pain of prostateand bone cancer. Beta emitter.Technetium-99m99mTc6.0hUsed to image the skeleton and heart muscle inparticular, but also for brain, thyroid, lungs(perfusion and ventilation), liver, spleen,kidney (structure and filtration rate), gallbladder, bone marrow, salivary and lacrimalglands, heart blood pool, infection andnumerous specialized medical studies.Thallium-201201Tl73hUsed for diagnosis of coronary artery diseaseother heart conditions such as heart muscledeath and for location of low-gradelymphomas. It is the most commonly usedsubstitute for 99Tc in cardiac-stress tests.Thorium-227227Th18.7dDecay product of 227Ac and parent of 223Ra.Xenon-133133Xe5.24dUsed for pulmonary (lung) ventilation studies.Ytterbium-169169Yb32dUsed for cerebrospinal fluid studies in thebrain.Ytterbium-169169Yb32dUsed for cerebrospinal fluid studies in thebrain.Yttrium-9090Y64hUsed for cancer brachytherapy and as silicatecolloid for relieving the pain of arthritis inlarger synovial joints. Pure beta emitter and ofgrowing significance in therapy, especiallyliver cancer. Current techniques for the production of radioisotopes involve the irradiation of a precursor material in which some of the precursor is converted into one or more radioisotopes distributed throughout the material. This is followed by dissolution of the material and subsequent separation of the unreacted precursor material from the produced radioisotopes. In currently known techniques for producing radioactive isotopes, target materials are often sealed in capsules and placed into irradiation locations. The irradiations can be performed with reactors or accelerators using a variety of particles and targets. After irradiation, the capsules are placed in shielded containers and transported to chemistry laboratories capable of handling the high activity of the material for batch dissolution and recovery of the radioactive isotope product or products. Using currently known techniques, the vast majority of the initial starting material in irradiations is unreacted and often must be dissolved to allow the irradiation products to be chemically separated. The separation of the minuscule amount of the irradiation product from a large excess of unreacted starting material often requires multiple purification routes after dissolution. If additional product is to be formed from nuclear reactions with the remaining starting material, the remaining starting material must be reformed after dissolution into a form suitable for subsequent irradiations. This requires additional processing steps, often with losses in potential product and additional waste generation. However, if the dissolved remaining starting material is not reformed for additional production, that remaining material must be disposed of, representing both a loss of potentially usable (and valuable) material and an additional disposal cost. For isotopically rare or enriched materials, this can be a large expense. The same issue can also arise in the context of nuclear fuel reprocessing. In this context, nuclear fuel, rather than precursor starting material, can be processed after irradiation in a power generating nuclear reactor, nuclear fission test reactor, research reactor, or teaching reactor, to remove byproducts of the nuclear fission reaction and to reprocess the unreacted nuclear fuel for recycling and/or reuse. Similar to radioisotope production, current recycling of used nuclear fuel from extraction, ion exchange, or electrochemical methods first requires dissolution of the fuel into a solution. After the fuel has been dissolved, the remaining steps are done to remove the unwanted byproducts and to reprocess the unreacted nuclear fuel back into a suitable form for reuse as fuel. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a starting material and further describes systems and methods that allow the same starting material to undergo multiple irradiations and separation operations without extensive, if any, damage to its original form. In one aspect, targetry coupled separation refers to the selection of a starting material (including selection of the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined irradiation product. The disclosure further describes how with targetry coupled separations, by removing the newly-created product in a way that allows the same starting material to undergo one or more subsequent irradiations (that is, without having to dissolve or otherwise destroy the material between irradiations), significant cost savings can be achieved using repeated irradiation and separation operations on the same starting material. The systems and methods described achieve this with a minimum waste of the starting material during the irradiation and product isotope removal and separation operations. In contrast with the prior art that requires complete or substantial dissolution or destruction of the starting material before recovery of any irradiation products, the repeated reuse of the starting material allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art. One aspect of this disclosure is a system for generating radioisotopes that includes: one or more containers, including a first container, the first container containing source material that includes at least one target material; a radiation generator; a radiation bombardment chamber that receives radiation from the radiation generator in which the radiation bombardment chamber is adapted to hold and expose the one or more containers to the received radiation, thereby creating at least some first radioisotopes that are a direct product of the exposure of the target material to the radiation; a insertion component adapted to transfer an extraction material into the first container, thereby contacting the source material within the first container with the extraction material in which the extraction material is selected to dissolve, without dissolving the target material, one or more of a first radioisotope, a second radioisotope that is a daughter product of a first radioisotope, or both a first radioisotope and a second radioisotope; and an extraction component adapted to remove extraction material with the dissolved radioisotope from the first container without removing the target material from the first container. In the system the target material may be a fissionable material and the radiation generator is a neutron generator. The source material may be a porous form with a pore wall width based on a recoil distance of a direct radioisotope product of the fissionable target material. The system may also include a recovery component adapted to receive the extraction material with the dissolved radioisotope from the first container and recover some of the at least one species of radioisotope from the extraction material. The system may also include a conveyance system adapted to physically move a container from the radiation bombardment chamber to a second location for interaction with one or both of the insertion component or the extraction component. The conveyance system may be further adapted to repeatedly move a container from the radiation bombardment chamber to the second location and from the second location to the radiation bombardment chamber. The conveyance system may also be adapted to physically move a plurality of containers from the radiation bombardment chamber to the second location. The fissionable target material may include grains containing uranium oxide or uranium metal having an average particle size of less than an average recoil distance of 99Mo as a fission product of uranium. In the system, the extraction material may be selected from a supercritical fluid and an aqueous fluid that preferentially dissolves the one or more of a first radioisotope, a second radioisotope that is a daughter product of a first radioisotope, or both a first radioisotope and a second radioisotope. The extraction material may be supercritical carbon dioxide containing a ligand that dissolves the one or more of a first radioisotope, a second radioisotope that is a daughter product of a first radioisotope, or both a first radioisotope and a second radioisotope. The system may automatically perform a radioisotope generation cycle in which the system exposes the first container to radiation, transfers the extraction material into the first container, and removes extraction material with the dissolved radioisotope from the first container. The system may also automatically repeat the radioisotope generation cycle on the first container. The system may process a plurality of containers, including the first container, such that each of the plurality of containers is exposed to radiation. The target material may include one or more of uranium oxide or uranium metal in the form of powder, salt, cloth, foam or a colloidal suspension in liquid. The source material may include radium or radium electroplated on beryllium. The at least one species of radioisotope generated by the system include one or more of 227Ac, 213Bi, 131Cs, 133Cs, 11C, 51Cr, 57Co, 60Co, 64Cu, 67Cu, 165Dy, 169Er, 18F, 67Ga, 68Ga, 68Ge, 198Au, 166Ho, 111In, 123I, 124I, 125I, 131I, 192Ir, 59Fe, 81mKr, 212Pb, 177Lu, 99Mo, 13N, 15O, 103Pd, 32P, 238Pu, 42K, 227Ra, 223Ra, 186Re, 188Re, 81Rb, 82Rb, 101Ru, 103Ru, 153Sm, 75Se, 24Na, 82Sr, 89Sr, 99mTc, and 201Tl. The radiation generator used by the system may be selected from one or more of Pu—Be sources, 252Cf sources, sealed tube radiation generators, dense plasma focus devices, pinch devices, inertial electrostatic confinement devices, sub-critical source driven assemblies, fission reactors, and accelerator spallation devices. Another aspect of this disclosure is a method for generating 99Mo. The method includes: providing a source containing a first mass of uranium in which the source is in a form in which a majority of uranium atoms are within a selected distance from an available surface of the source; exposing the source to neutrons, thereby reducing the first mass of uranium in the source to a second mass of uranium less than the first mass and creating at least some atoms of the 99Mo radioisotope and thereby also causing at least some of the newly created atoms of the 99Mo radioisotope to move toward an available surface of the source; and after exposing the source to neutrons, removing at least some of the atoms of the 99Mo radioisotope from the source without substantially removing uranium from the second mass of uranium in the source. In the method, the removing operation may remove less than 0.1% of the uranium or even less than 0.01% of the uranium from the second mass of uranium in the source. The providing operation may include providing a source made at least partially from particles containing uranium oxide or uranium metal in which the particles have a particle size based on a recoil distance of 99Mo in the source. The method may include enclosing the source in a neutronically-translucent container. The method may further include exposing the container enclosing the source to neutrons and removing at least some of the atoms of the 99Mo radioisotope from the container. The method may include selecting an extraction material that dissolves atoms of the 99Mo radioisotope without changing the phase of the uranium in the source or selecting an extraction material in which atoms of the 99Mo radioisotope are more soluble than atoms of uranium. The method may include determining the form of the source based on the selected extraction material. The method may also include selecting a combination of a source and an extraction material, wherein the combination allows 99Mo radioisotope to be removed from the source after exposure to neutrons without substantially affecting the source. The removing operation of the method may include passing an extraction material selected to dissolve the 99Mo radioisotope through the container, thereby contacting the available surface of the source with the extraction material. In the method, the extraction material may be selected from a supercritical fluid and an aqueous fluid. If the extraction material is supercritical carbon dioxide, it may contain a ligand that dissolves the 99Mo radioisotope. The ligand may be selected from 8-hydroxyquinoline, α-benzoinoxime, disodium 4,5-dihydroxy-1,3-benzenedisulfonate, phosphate compounds, and diketone compounds. In an alternative embodiment, the ligand may have one or more functional groups selected from hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate, isothiocyanate, phosphate, and phosphono groups. The method may include repeatedly performing the exposing operation and the removing operations on the container without removing the uranium from the container. The method may include removing the 99Mo radioisotope from the extraction material and, afterwards, repeating the exposing operation on the same source. The method may further include removing, in addition to the 99Mo radioisotope, an amount of one or more other fission products created during the exposing operation. Yet another aspect of this disclosure is a method for selectively manufacturing a radioisotope. The method includes: selecting the radioisotope; identifying a target material from which the selected radioisotope can be created as a fission product; determining a recoil distance of the selected radioisotope in the target material; creating a plurality of grains of target material having a grain size based on the recoil distance of the selected radioisotope; exposing the grains of target material to neutrons, thereby causing at least some atoms of the target material to undergo nuclear fission to create atoms of the selected radioisotope and also causing at least some of the newly created atoms of the selected radioisotope to move the recoil distance relative to the target material; and extracting atoms of the selected radioisotope from the target material. In the method, the exposing and extracting operations may be performed without changing the phase of the target material. The method may further include repeating the exposing operation and extracting operation on the plurality of grains of target material. The plurality of grains of target material may be contained within a neutronically-translucent container and the method may further include repeatedly performing the exposing operation and extracting operation on the same plurality of grains of target material without removing the plurality of grains of target material from the container. The grains may be particles of uranium oxide or uranium metal having a grain size of less than 20 microns. In an embodiment, the grains are particles of uranium oxide or uranium metal having a grain size from about 0.001 to 10 micrometers. In the method, at least some of the plurality of grains may have a characteristic length along at least one dimension smaller than or equal to the recoil distance. The extraction operation of the method may include exposing at least some of the grains of target material to a solvent that preferentially extracts the selected radioisotope from the target material. The method may further include processing the grains of target material into a solid, porous source prior to the exposing operation. The processing may include processing the grains of target material into an open-cell foam, an open lattice, an open framework, a ceramic, a cloth, a thin film, a monolayer, a sponge, a nanocage, or a nanocrystal. The processing may include processing the grains of target material into a solid, porous source having a surface area greater than 10 m2/g as measured by Brunauer, Emmett and Teller (BET) analysis. The processing may include one or more of sintering, milling, sieving, 3D printing, crystallizing, precipitating, or heating the grains of target material. Another aspect of this disclosure is a method for selectively manufacturing a radioisotope. The method includes: receiving a source having solid fissionable material in a neutronically-translucent container in which the source has a porous form with pore walls having a width substantially similar to a recoil distance of a radioisotope product of the solid fissionable material; exposing the source to neutrons, thereby converting at least some atoms of the solid fissionable material via nuclear fission into atoms of the radioisotope so that the source contains radioisotope and unconverted solid fissionable material; selecting an extraction material that preferentially dissolves the radioisotope relative to the fissionable material; injecting the extraction material into the container, thereby contacting the source material with the extraction material; removing extraction material from the container after a residence time, thereby removing at least some dissolved radioisotope from the container while leaving substantially all of the unconverted solid fissionable material in the container; and, after removing the extraction material, re-exposing the source material to neutrons, thereby converting at least some atoms of the unconverted solid fissionable material via nuclear fission into atoms of the radioisotope. The method may further include repeating the injecting and removing operations on the target after re-exposing the target to neutrons. The method may also include separating dissolved radioisotope from the extraction material; and incorporating the dissolved radioisotope into a daughter isotope generator. The method may further include periodically milking the daughter isotope generator for the daughter isotope. The method may include waiting until at least a first predetermined period of time has elapsed after removing the extraction material from the container in which the predetermined period of time being based on a half-life of the radioisotope; and separating the radioisotope from the extraction material. Another aspect of this disclosure is a method for manufacturing a radioisotope-generating target. The method includes: providing a dissolved salt of a fissionable material in a solution in which the fissionable material capable of generating a first designated radioisotope when exposed to neutrons and the first designated radioisotope having a recoil distance associated with the fissionable material, precipitating an oxide of the fissionable material from the solution; and selectively forming the precipitated oxide into grains, i.e., individual particles, having a grain size based on the recoil distance of the first designated radioisotope. The method may further include mixing a precipitant into the solution and/or selecting the fissionable material based on the first designated radioisotope. The method may include determining the recoil distance of the first designated radioisotope based on the selected fissionable material and/or forming grains having a grain size equal to or less than 10 micrometers. The method may include forming grains having a grain size equal to or less than 1 micrometer. The method may include forming grains having a grain size equal to or less than 100 nanometers and the method may include forming grains having a grain size equal to or less than 10 nanometers. In the method, the forming operation may include one or more of milling, drying, filtering, washing, calcining, or sintering the precipitated oxide. The method may include packaging the grains of the precipitated oxide in a container. The container may have a first valve adapted to allow the injection of a solvent into the container and a second valve adapted to allow the extraction of a solvent from the container. The container may be neutronically-translucent. Packaging the grains may further include placing the grains in a cavity defined by the container; and sealing the container, thereby trapping the grains in the cavity. The method may include synthesizing a ceramic from the precipitated oxide grains. In the method, the providing operation may further include providing the dissolved salt of the fissionable material in a solution selected from one or more of an acidic solution, a basic solution, an aqueous solution, and an alcohol solution. In yet another aspect of this disclosure, a radioisotope-generating target is described. The target includes a target material capable of generating the radioisotope upon prolonged exposure to neutrons and the radioisotope associated with a recoil distance; and the target material having a characteristic distance selected based on the recoil distance of the radioisotope. In the target, the radioisotope may be a direct fission product of nuclear fission of the target material. A neutronically-translucent container may be used to contain the plurality of grains. The container may have an input valve and an output valve allowing the injection and extraction of a fluid. The target's container may include a body portion and at least one removable lid portion that, when engaged, encloses the target material within the container. The container may be made of one or more of aluminum, aluminum alloy, zirconium, zirconium alloy, molybdenum, molybdenum alloy, and stainless steel. The target material may include a plurality of grains of target material loosely packed in the container and the characteristic distance is a grain size selected based on the recoil distance of the radioisotope. The target material may include a plurality of grains of target material formed into a ceramic. The target material may include a plurality of grains of target material formed into or attached to a metal-organic framework. The target material may include one or more of uranium oxide or uranium metal. The target material may include a plurality of grains of target material formed into a loose powder, a cloth, a foam or a colloidal suspension in liquid. The target material may include radium or radium electroplated on beryllium. The target material may include grains of an actinide monolayer and the actinide monolayer may be a monolayer of uranium. The target material may include grains of a high surface area, uranium metal which may be created using the Kroll process. Yet another aspect of this disclosure is a supercritical carbon dioxide separation method. The method extracts a first radioisotope from irradiated fissionable material containing a plurality of radioisotopes including the first radioisotope. The method includes: selecting a ligand that is soluble in supercritical carbon dioxide (sCO2), forms a chelate with the first radioisotope, and does not form a chelate with the fissionable material; dissolving the identified ligand into sCO2 to form an sCO2-ligand solution; contacting the irradiated material with the sCO2-ligand solution for a contact time, thereby creating an sCO2-radioisotope complex solution; separating the sCO2-radioisotope complex solution from the irradiated material; and after separating the sCO2-radioisotope complex solution from the irradiated material, removing the radioisotope from the sCO2-radioisotope complex solution. In the method, removing the radioisotope from the sCO2-radioisotope complex solution may include removing the radioisotope complex from the sCO2-radioisotope complex solution. The removing operation may generate the sCO2-ligand solution suitable for reuse without decompressing and repressurizing the sCO2 ligand solution. This may be achieved by contacting the sCO2-radioisotope complex solution with an acidic solution, thereby generating an acid-radioisotope solution and a regenerated sCO2-ligand solution. In the method, the irradiated material may be enclosed in a container and the exposing operation may further include passing the sCO2-ligand solution through the container without removing substantially any of the fissionable material from the container. In this case, the container may be operated as a packed bed reactor. In the method, the irradiated material may be in the form of loose grains and the exposing operation further includes passing the sCO2-ligand solution through the container at a flow rate sufficient to fluidize the plurality of grains within the container. In the method, the irradiated material may also be a liquid. In the method, the radioisotope may be 99Mo, the fissionable material may be 235U and the ligand may have one or more functional groups selected from hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate, isothiocyanate, phosphate, and phosphono groups. The ligand may be selected from a fluorinated β-diketone and a trialkyl phosphate, or a fluorinated β-diketone and a trialkylphosphine oxide. The ligand may be selected from dithiocarbamates, thiocarbazones, β-diketones and crown ethers. The ligand may have one or more functional groups selected from hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate, isothiocyanate, phosphate, and phosphono groups. Radioisotopes that may be created by this method include one or more of 227Ac, 213Bi, 131Cs, 133Cs, 11C, 51Cr, 57Co, 60Co, 64Cu, 67Cu, 165Dy, 169Er, 18F, 67Ga, 68Ga, 68Ge, 198Au, 166Ho, 111In, 123I, 124I, 125I, 131I, 192Ir, 59Fe, 81mKr, 212Pb, 177Lu, 99Mo, 13N, 15O, 103Pd, 32P, 238Pu, 42K, 227Ra, 223Ra, 186Re, 188Re, 81Rb, 82Rb, 101Ru, 103Ru, 153Sm, 75Se, 24Na, 82Sr, 89Sr, 99mTc, and 201Tl. Another aspect of this disclosure is a method of obtaining a radioisotope from a bulk material, in which the bulk material includes at least the radioisotope and a fissionable material. The method includes: selecting an extraction material that removes the radioisotope from the bulk material without substantially dissolving the fissionable material; contacting the bulk material with the extraction material for a residence time, thereby creating an extraction material and radioisotope mixture; after the residence time, removing the extraction material and radioisotope mixture; and separating the radioisotope from the extraction material. In the method, the contacting operation may further include one or more of: agitating one or both of the bulk material and the extraction material during at least a portion of the residence time; changing a temperature of one or both of the bulk material and the extraction material during at least a portion of the residence time; and changing a pressure of one or both of the bulk material and the extraction material during at least a portion of the residence time. In embodiments of the method in which the bulk material is solid, contacting the bulk material may include contacting the bulk material with a liquid extraction material for a residence time, thereby creating an extraction material and radioisotope liquid mixture. In embodiments of the method in which the bulk material is a liquid, contacting the bulk material may include contacting the bulk material with a liquid extraction material for a residence time, thereby creating an extraction material and radioisotope liquid mixture immiscible in the bulk material. In an embodiment, the bulk material may be in the form of solid grains stored in a container and the contacting operation may include inserting an amount of the extraction material into the container, and retaining the extraction material in the container for the residence time. In the method, the extraction material may include an extractant and a solvent. The extractant may be a ligand soluble in the solvent under temperature and pressure conditions of the contacting operation. The solvent may be sCO2. The ligand may form a carbon dioxide soluble chelate with the radioisotope. The ligand may be selected from a fluorinated β-diketone and a trialkyl phosphate, or a fluorinated β-diketone and a trialkylphosphine oxide or selected from dithiocarbamates, thiocarbazones, β-diketones and crown ethers. The ligand may have one or more functional groups selected from hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate, isothiocyanate, phosphate, and phosphono groups. These and various other features as well as advantages which characterize the systems and methods described herein will be apparent from a reading of the following detailed description and a review of the associated drawings. Additional features are set forth in the description which follows, and in part will be apparent from the description, or may be learned by practice of the technology. The benefits and features of the technology will be realized and attained by the structure particularly pointed out in the written description and claims hereof as well as the appended drawings. It is to be understood that both the foregoing general description and the following detailed description are explanatory and are intended to provide further explanation of the invention as claimed. FIG. 1 illustrates, at a high level, an embodiment of a targetry coupled separation method that repeatedly generates irradiation products, such as radioisotopes, from some amount of target material. In the method 10 as illustrated, some amount of target material is irradiated in an irradiation operation 12 which creates a desired irradiation product. The irradiation operation 12 is followed by a separation operation 14 in which the desired product is removed from the target material without substantially reducing the amount of post-irradiation target material. (As discussed in greater detail below, the word ‘substantially’ shall be used at times when referring to the amount of target material that remains in the irradiated object after a separation operation 14 to remind the reader that no separation technique is perfect and a small or de minimis amount of the target material may, in fact, be removed during the separation operation 14.) The desired product, after removal from the target material, may then be subjected to subsequent processing and use. For example, in an embodiment the target material is incorporated into a porous, solid object and the product is removed using a liquid solvent that dissolves the product but does not substantially, if at all, dissolve or remove the remaining post-irradiation target material from the object. The irradiation and separation operations 12, 14 are then repeated on the remaining target material. The method 10 may be repeated any number of times and, in an ideal system, could be repeated until all of the target material is completely consumed. Realistically, however, it is presumed that after some number of repetitions it will become more economical to dispose of the remaining target material rather than reuse it for another cycle. While many more detailed embodiments, some of which are discussed below, are possible, FIG. 1 is presented as a simplified embodiment in order to provide a convenient reference point for further discussion and to introduce the concepts and terminology that will be discussed in greater detail below. As discussed above, in targetry coupled separations a target material is subjected to one or more irradiation operations. The “target material,” as that term will be used herein, refers to a material that, upon exposure to the particular radiation used in an irradiation operation, results in the creation of one or more irradiation products. Depending on the embodiment, the radiation used may include one or more of alpha particles, beta particles, gamma rays, x-rays, neutrons, electrons, protons, and other particles capable of producing nuclear reaction products. In any particular irradiation operation, some amount of the target material will be converted into the irradiation product(s), resulting in a mass decrease of target material and a newly created mass of the irradiation product(s). In some embodiments, targetry coupled separations may be tailored to enhance the recovery of one or more predetermined, desired products from the irradiated target material. A desired product refers to either a direct or an indirect irradiation product that the operator wants to remove from the target material after irradiation in the separation operation 14. Depending on the combination of radiation and target material used in an embodiment, undesired reaction products may also be created by the irradiation, which may not be removed from the target material in the separation operation 14. For example, if the radiation is in the form of neutrons and the target material includes uranium-235 (235U), one of the fission products will be the molybdenum isotope, 99Mo. After irradiation, atoms of 99Mo will be dispersed within the target material and each 99Mo atom will be from one uranium atom that existed prior to irradiation. However, due to the nature of neutron irradiation, many other fission products will also exist in the target after irradiation, each also representing atoms produced from fissioned uranium atoms. In an embodiment, the 99Mo is a desired product and subsequently removed in the separation operation 14 while the other fission products are not removed and remain with the target material during subsequent irradiations. Target material may be incorporated into a larger mass of source material. The source material may be formed into a single object or discrete mass, occasionally referred to herein simply as the “source”, that can be exposed to radiation in an irradiation operation to convert at least some of the target material (either directly or indirectly, as discussed in greater detail below) into the desired product (or its parent, as will be discussed below). In addition to the target material, source material also may optionally include material that does not react to irradiation to produce the desired product. Such material may be completely unreactive to the radiation or may form something other than the desired product. Where appropriate, the term ‘ancillary material’ may be used to refer to any component of the source material that does not form the desired product when irradiated. Ancillary materials could include, for example, trace contaminants, materials present in the source material to provide a physical structure for the target, or unharvested products of previous irradiations of the target. FIG. 11 illustrates the material conversion cycle showing the changes in a source over two passes of targetry coupled separation. The material conversion cycle 1100 starts with some amount of source material 1102 that, in the embodiment shown, includes some amount of target 1120 and another amount of ancillary material 1122. Although the target and ancillary material of the source material are shown as separate boxes, it is to be appreciated that the FIG. 11 is for illustrative purposes of the cycle of target and product, and it should be appreciated that the elements of FIG. 11 are not representative of actual amount or ratio of target and ancillary material, intermixing of target and ancillary material, and/or structure of the source. An irradiation operation (such as irradiation operation 12 of FIG. 1) changes the source material 1102 into an irradiated source material 1104 in which some amount of target material 1120 has been changed into irradiation products 1124. While exaggerated for illustration purposes, in FIG. 11 approximately half of the target material 1120 has been changed into some amount of irradiation product 1124. Because products 1124 are considered an ancillary material, FIG. 11 also illustrates a relative increase in the mass of ancillary material in the source 1104 and a commensurate decease in the amount of target material 1120. The reader will be reminded that, especially in fission reactions, the immediate result of irradiation will be a spectrum of direct irradiation products, some which over time may subsequently decay into indirect products which may, themselves, further decay into other indirect products. Thus, the exact makeup of irradiation products 1124 may change over time as various direct and indirect products decay. However, for the purposes of this discussion, FIG. 11 does not distinguish between direct irradiation products and indirect irradiation products or attempt to track how the makeup of irradiation products changes over time. The cycle 1100 further shows the effects of a first separation operation (such as separation operation 14 of FIG. 1) on the irradiated source material 1104. The separation results in a certain amount of desired product 1126 being removed from the irradiated source material 1104. Again, exaggerated for illustration purposes, FIG. 11 shows the post-separation source 1106 having had some of the product removed, so that the post-separation source material 1106 has relatively less ancillary material, but the amount of target remains the same as in the irradiated source material 1104. This graphically illustrates that the separation operation has no effect or substantially no effect on the mass of target material 1120 in the source. FIG. 11 also illustrates that some irradiation products 1124 may remain as ancillary material 1122 in the source 1106 after the separation. This may be the case either because the separation is not 100% efficient, because not all of the irradiation products 1124 are desired products and the separation operation intentionally does not remove those products, or both. FIG. 11 further illustrates a second set of irradiation and separation operations on the source material. FIG. 11 shows a second-irradiated source 1108 that again illustrates that some amount of target material 1122 of the precursor source material 1106 is converted into product 1124 by the second irradiation operation. The second separation operation then reduces the overall mass of the second-irradiated source material 1108 by removing some of the desired product 1126, but without changing the mass of the target material 1122 in the source material 1108. The resulting post-second separation source material 1110 is then ready for subsequent irradiation and separation operations as illustrated by the arrow at the bottom of the illustration. As mentioned above, FIG. 11 is exaggerated for illustration purposes. However, it clearly shows certain aspects of targetry coupled separation. Specifically, it illustrates that target material 1122 is converted into product 1124 by the irradiation operation and some amount of product 1126 is removed, without removing substantially any target material 1122 from the source, in the separation operation. Thus, by subjecting the same source to repeated irradiation and separation operations, the target material 1122 in the source can be consumed until such time as it is completely converted into product 1124 or it is no longer economical to repeat the process. FIG. 11 further illustrates that not all of the product 1124 may be removed by the separation operation. This may occur for different reasons. While it is preferable to remove as much of the desired product 1126 as possible with each separation, not all of the irradiation products 1124 may be desired products 1126 and/or removal of all of the desired product 1126 may not be technologically practical or possible. Thus, products 1124 from prior irradiation operations (such as undesired products) may remain in the source material by design (e.g., by appropriate selection of the extraction material to avoid or reduce removal of the undesired product). It is also possible that the separation operation is not 100% efficient at removing all the product, thus leaving some desired product in the source material. As discussed above, target material can include any one or more isotopes or elements that, directly or indirectly, can form the desired product upon irradiation. The term ‘directly or indirectly’ is used here to point out to the reader that, while some desired isotopes may be the direct irradiation product of a target material, other desired products may be created by the natural decay of a direct irradiation product. For example, 99Mo is one of many direct fission products of 235U. That is, in the thermal neutron fission of a mass of 235U, some of the atoms (6.1% to be precise) of 235U will be converted directly into atoms with a mass of 99, including 99Mo. Other atoms of 235U will be converted into other products such as 135I and 157Gd. However, many direct fission products are unstable and will, after some period of time based on their half-lives, naturally decay into indirect products. Using 99Mo again as the example, 99Mo has a half-life of 65.94 hours, primarily decaying into 99mTc. The isotope 99mTc, with a 6.01 hour half-life, decays into 99Tc. Thus, 99Mo is a direct product of the fission of 235U while 99Tc is an indirect product. It should be noted that 99Tc is also a direct product of fission, but with different independent fission yield than 99Mo. It should be noted that a desired product may be both a direct product from irradiation of a target and an indirect product that is created by decay of a different direct product of the same irradiation of the target. Target material can include the elemental form of a material, metals, alloys, intermetallic compounds, hydrides, oxides, hydroxide, halides, chalcogenide, nitrides, phosphides, carbides, silicides, carbonates, nitrates, sulfates, thiosulfate, sulfites, perchlorates, borides, arsenates, arsenites, phosphates, nitrite, iodate, chlorate, bromate, chlorite, chromate, cyanides, thiocyanates, amides, peroxides, organic complexes, mixed species, ternary compounds, quaternary compounds or greater, or a combination of any of these compounds. The source material can be in a variety of structures, forms or morphologies that permit the separation of the desired product from the target without significant alteration of the physical form of the source material (other than the removal of some or all of the desired products), thus allowing previously irradiated source material to undergo a subsequent irradiation without substantial reprocessing. Morphologies, forms, and shapes can include sheets, monoliths, sol-gels, ceramics, polymers, metallic phases, particles, spheres, layers, aggregates, crystalline phases, metal-organic frameworks, fibers, precipitates, tubes, micelles, sponges, cages, powders, granules, suspensions, slurries, emulsions, porous particles, and colloids. Furthermore, as will be described in greater detail below with reference to FIGS. 4 and 5, a source material's physical form or morphology may be selected or altered in order to improve the performance or efficiency of the separation operation 14, e.g., by tailoring the form of the target material in the source to suit the selected extraction material or process. For example, a particularly high surface area form of source material may be used to improve the contact between a solid target material and a liquid or gaseous extraction material, such as a supercritical carbon dioxide and ligand mixture. Alternatively, a form of source material may be selected to take advantage of the effect of irradiation. For instance, some uranium fuels (e.g., ceramic and metal fuels) can become porous after irradiation in a reactor, which can prepare the target for separation of the product from the target and subsequent re-irradiation without the need to dissolve or destroy most or all, if any, of the remaining source material as part of the separation. Although, in an embodiment, the source may be a solid piece or structure including the target material, in many embodiments discussed herein the source material may be contained in a container that, at least partially, encases the source material. For example, in an embodiment the source material may be in a particulate or pelletized form and a container may be provided to hold the source material during some or all of the operations of FIG. 1. Depending on the physical form of the source material (e.g., aggregate, powder, liquid, etc.), a container may be used to provide a physical constraint and may also be used to provide contact points for ease of handling. In addition, the container may be adapted to simplify the separation operation 14. Suitable container embodiments are discussed in greater detail with reference to FIG. 4. The form of the porous source material can be manufactured or selected, such as through 3D printing, foam, molds, particulate, sintering particulate, etc. as will be discussed further below. Returning to FIG. 1, in an embodiment of the irradiation operation 12, one or more sources are exposed to radiation that causes at least some of the target material to be converted into desired product. Radiation generators can include reactors, particle accelerators, electron accelerators, plasma focus devices, pinch devices, and/or sealed tube neutron generators. The accelerators can supply reaction particles directly or can be used to produce particles from reactions. In an embodiment, the irradiation operation 12 may include placing one or more sources containing target material in a controlled environment where the source(s) may be safely exposed to the radiation. For example, in an embodiment in which the radiation includes neutrons, exposure is achieved by placing the source material in, or passing a source through, a neutron bombardment chamber that receives neutrons from a neutron generator. In the separation operation 14, the exposed source material is treated to remove the desired product without substantially dissolving or removing the remaining target material in the source. In an embodiment, this may include contacting an available surface of the source material with an extraction material, such as a fluid, that preferentially dissolves the desired product but for which the target material and ancillary material, if any, is either insoluble or has a substantially reduced solubility relative to the desired product. In an alternative embodiment, some other separation technique may be used that preferentially removes the desired product from the source material. The target material in a source is left in a form suitable for subsequent irradiation to generate additional desired product. The reader will appreciate that no separation system is perfect and that some trace amount of target material may be unintentionally entrained, dissolved and/or otherwise removed with the extraction material during the separation operation 14. As mentioned above, the word ‘substantially’ shall be used at times when referring to the amount of target that remains in the source material after a separation operation 14 to remind the reader that some small or de minimis amount (less than 0.1% although less than 0.01% is anticipated) of the target material by mass may, in fact, be removed from the source during the separation operation 14. Although the techniques introduced above and discussed in detail below may be implemented for a variety of desired products such as radioisotopes or other fission products, this disclosure will primarily discuss targetry coupled separation systems and methods in the context of systems and methods that repeatedly generate and remove one or more fission products from source material containing fissionable material as the target. More particularly, this disclosure will primarily discuss targetry coupled separation in the context of systems and methods that repeatedly generate and remove 99Mo as the desired product from grains of a source material that includes 235U as the target. Upon fission of the uranium, 99Mo product is one of the many isotopes produced as fission products. The 99Mo product can be separated from the uranium by the formation of a molybdenum-specific species that can be easily removed from the uranium target without the need to remove the target material from the source or alter the form of the target material to facilitate separation. An example of suitable molybdenum specie that facilitate separation from the source includes MoO42−, which can be removed by dissolution, or Mo(CO)6, which can be removed by volatilization. The reader will understand that the technology described in the context of 99Mo could be adapted for use in generating any nuclear reaction product, such as those listed in Table 1, either directly by irradiating an appropriate target material with neutrons, or indirectly by irradiating the appropriate target material with neutrons to form a radioisotope parent of the desired product and allowing the parent to decay. More generally, the targetry coupled separation methods and systems described herein may be adapted to generate any desired product that could be obtained through irradiation of a target material using any type of radiation, not just neutron irradiation. FIG. 2 illustrates, again at a high level, an embodiment of a targetry coupled separation system. The system illustrated is adapted for the continuous or semi-continuous production of products such as the 99Mo radioisotope from a target containing 235U. While embodiments of the system 200 may include manual operations, the system 200 is particularly suited for automation and the entire process may be implemented as an automated system that continuously or semi-continuously generates 99Mo product until such time as the target is consumed or otherwise fouled with unwanted byproducts to the extent that further generation of 99Mo from the targets is uneconomical. For example, some fission products of 235U are neutron poisons (such as 135Xe, 149Sm and 151Sm) and, if these products are allowed to buildup in the source material over successive re-irradiations, the subsequent yield of desired product from each irradiation will be reduced. Even then, in an embodiment, old source material may be automatically stored and new source material placed into the system until all available or a desired amount of target material is consumed. The system 200 includes: a neutron generator 202; a neutron bombardment chamber 204; a conveyance system 206 (illustrated as a conveyor-type system 206); a separation system 208, which in this embodiment includes two components: an insertion component 210 and an extraction component 212; an optional treatment system 228; a product storage system 224; and a supply or source of extraction material 226. A plurality of uranium-containing sources 214 is illustrated undergoing various operations by the separation system 200 and traveling in the direction of the conveyor-type system 206 as indicated by arrows 220 and 222. In the embodiment shown, the neutron generator 202 can be any appropriate generator of neutrons. Examples include Pu—Be sources, 252Cf sources, sealed tube neutron generators, dense plasma focus device, pinch devices, inertial electrostatic confinement device, fission reactors, or accelerator spallation devices. The neutron bombardment chamber 204 receives neutrons from the neutron generator 202 and exposes any sources 214 within the chamber 204 to neutron bombardment. The chamber 204 may include multiple components designed to allow the sources to enter and exit. The chamber 204 may be constructed to reduce the release of stray neutrons to the outside environment of chamber 204 itself or outside of the containment of system 200. The chamber 204 may include an irradiation zone within which the sources are exposed to neutrons. The irradiation zone may be sized to irradiate any desired number of sources at the same time. In the embodiment shown, the conveyor 206 causes sources to pass through the irradiation zone. Because the rate of source transport into and through the irradiation zone determines, in part, the total exposure of the source to neutrons, the transport rate may be selected to achieve the desired amount of irradiation of the target based on the neutron flux of the neutron generator 202. Transportation of sources may be continuous (wherein the sources are continuously in motion), non-continuous (in which the conveyance system 206 starts and stops to achieve the desired rate), or a combination of the two (e.g., continuous movement through the bombardment chamber 204 but sources are held in the separation system 208 until a desired amount of separation has been obtained alter which transportation is resumed). The rate may be constant, decreasing, intermittent or varied based on monitoring of neutron flux or any other parameter that can be used to identify the exposure of the target material to radiation. It should be appreciated that the exposure level of irradiation in the irradiation zone need not be constant for a particular source, or from one source to the next source introduced to the irradiation zone. That is, the neutron generator 202 may not have a constant neutron flux over time. In this situation, the neutron flux may be monitored and the transport rate may be varied as necessary to achieve the desired irradiation results. When a new source enters the irradiation zone, there is little or no product or an undesired amount of product in the source material. Given time in the irradiation zone, the nuclear reaction product concentration increases in the source as some of the atoms of uranium undergo a nuclear reaction due to the uranium atoms' interaction with the neutrons. The conveyor speed, stopping points and times and/or neutron flux in the irradiation zone may be tuned so that sources are exposed for a desired irradiation time or dosage, thus generating a designed amount of fission products including 99Mo in each source. The source is then removed from the irradiation zone by further movement of the conveyor 206 and passed to the separation system 208. The separation system 208 refers to those components that together pass extraction material through the source to remove at least some of the atoms of the 99Mo radioisotope product from the source without substantially reducing the post-irradiation uranium content of the source. The separation system 208 obtains extraction material from the extraction material supply 226, contacts the irradiated source material with the extraction material, and then removes the extraction material (along with at least some of the 99Mo product) for the source material. As illustrated graphically in FIG. 11, the separation system 208 does not substantially, if at all, reduce the mass of the target material in the source, but rather exclusively or primarily removes only the desired product or products. Generally, an extraction material may be introduced to the irradiated source material (including the product material) to dissolve the 99Mo radioisotope product into the extraction material and substantially retain the full, post-irradiation mass of target material in the source material, separate from the extraction material. The chemistry of the extraction material used to perform the separation may be tailored to the target material and the desired product, and, depending on the embodiment, may use aqueous solutions, organic phases, ionic liquids, supercritical fluids, fluidized beds, reactive gases, thermal treatments, or their combinations. However, in this embodiment the separation system 208 uses an extraction material that is placed in contact with the irradiated target material, now containing an amount of 99Mo product. In an embodiment, the extraction material preferentially dissolves the 99Mo product without dissolving the uranium target or, preferably, any of the ancillary material including any other byproducts such as other fission products. In an alternative embodiment, the desired product is, in fact, multiple fission products and the extraction material preferentially dissolves all of the desired products simultaneously, without substantially affecting the remaining target material in the source material. In yet another embodiment, multiple different extraction materials are used sequentially in separate contacting operations to remove the various products. In yet another embodiment, multiple different extraction materials are used in a single contacting operation to remove the various desired products. The removed product(s) are then recovered from the extraction material(s) and processed as necessary into a usable form and the sources are returned to the conveyance system 206 for further irradiation. In the embodiment shown, the product is output into a product storage system 224. The recovery of the product(s) may be performed by the extraction system 208 so that a final, usable form of product(s) is stored by the storage system 224. In an alternative embodiment, the product and extraction material mixture may be stored in the storage system 224 for future processing by a separate recovery system (not shown), which may be local to or remote from the system 200. In the embodiment illustrated in FIG. 2, the first stage of the separation system 208 is the insertion component 210. In the embodiment shown, the targets are contained in containers 214 encasing some amount of target material and the insertion component 210 refers to that equipment that transfers the extraction material into the containers. The insertion component is adapted to transfer an extraction material into containers, thereby contacting the source material within the first container with the extraction material. In an embodiment, the extraction material is selected to dissolve, without dissolving the target material, the desired product or products. Transferring the extraction material into the containers may include one or more of injecting the extraction material under pressure into the container, applying a vacuum to a container open to a reservoir of extraction material, allowing the extraction material to flow under gravity into the container, submerging an open container into a pool of extraction material, or any other technique whereby the extraction material is transferred into the container. The insertion component 210 may include automated or manually-operated equipment that accesses the container and delivers the extraction material into the container, such as through one or more valves or other access points provided on the container. As is known in the art, there are many different ways of inserting fluids into a container and any suitable method may be used. In an embodiment, the extraction material is maintained in the container for an appropriate residence time. During some or all of the residence time, the container may be subjected to additional actions such as heating, cooling, pressurization, depressurization, agitation, circulation of extraction material, and/or secondary irradiation as desired to improve the removal of the product from the source material. For example, in an embodiment the source material is a loose particulate or powder and the extraction material is repeatedly flowed (circulated) under pressure through the container (e.g., flowed into a valve at one end of the container and removed from a valve at the other end of the container) such that the container temporarily becomes a packed bed reactor or, if the flow rate through the container is sufficient, a fluidized bed reactor. In these embodiments, the contacting of the extraction material with the source material is performed substantially without removing target material from its container, and in some cases without removing any source material other than the desired product from the container. After the appropriate residence time, the extraction component 212 of the separation system 208 removes the extraction material from the container and passes the extraction material including the removed product to a treatment system 228. As with the insertion component 210, any suitable technique for removing the extraction material and product mixture including those described above for inserting the extraction material into the container may be used. The extraction solution can admix or carry the product. Alternatively, the extraction solution (including the extraction method and parameters of operation) can be selected to dissolve the product. The treatment system may separate the dissolved 99Mo product from the extraction material. The treatment system 228 and/or a post-processing system (not shown) that is considered a part of the separation system 208 for the purposes of this discussion may purify the removed product into a usable 99Mo or further decay product, which is then stored in the product storage system 224. For example, in an embodiment the 99Mo product may be incorporated into an isotope generator by the separation system 208 as a final processing step. The extraction material may be further regenerated for reuse, such as by removal of any unwanted byproducts or trace source material picked up by the extraction material. However, regeneration is optional and the separation system 208 may or may not regenerate the extraction material as part of the recovery of the 99Mo product. The extraction material may be returned to the extraction material supply 226 for reuse by the insertion component 210. Alternatively, the extraction material may be processed for waste and/or removal from the system 200. For example, in an embodiment that uses sCO2 as part of the extraction material, the treatment system 228 may maintain the sCO2 in the supercritical state during the separation and returned recycled sCO2 to the separation system 208. The purification of 99Mo product may include removal of a trace amount of target elements or isotopes, removal of other products from the nuclear reaction, and/or removal of separation chemical(s) used in the separation of the product from the source material. The methods for purification are based on existing techniques and can include any one or more appropriate techniques including column chromatography, gravity separation, distillation, evaporation, centrifugation, precipitation, ion exchange, sorption, filtration, and solvent extraction. These methods can be performed with an automated chemistry system. Additionally, the extraction component 212 may also perform one or more regeneration operations to prepare the source material for further irradiation. Such regeneration operations can include washing the remaining source material with a volatile, acidic or basic solution, heating, treatment under vacuum, sparging with gas, flushing with a solution, or any other appropriate process or a combination of any of these processes. The regeneration operations can occur at the same location as the extraction operation and may use the same equipment, as shown in FIG. 2. For example, the extraction component 212 may perform the source regeneration and, in that capacity, may also be considered a source regeneration component. In an alternative embodiment (not shown), the source regeneration operations can occur at a different location and/or use separate equipment, such as an independent source regeneration component (not shown). Some embodiments of targetry coupled separation may have advantages compared to the existing methods of isotope production. The reuse of the source containing the target is an attribute in this regard. Because targets may be composed of enriched or rare isotopes, embodiments may provide a ready route to re-irradiate the target with reduced preparation and/or regeneration expenses. Furthermore, in various embodiments separating the produced isotope product from the source does not substantially reduce the amount of target material in the source (after conversion of some amount of target material into product through irradiation) nor even require the target material be removed from the source material or even the container. As will be appreciated, target dissolution can result in waste formation, which can represent a significant expense with radioactive material. While the target material may be recovered after dissolution and reformed into a new source, losses of target through imperfect reformation and/or costs of reformation may have an impact on fabrication costs and waste formation. And that is not to mention the extra cost associated with reformation of the target material into a new source. Embodiments of targetry coupled separation can be incorporated into existing reactors or accelerator centers, thereby utilizing current infrastructure which often are included in and/or accompany these facilities. This utilization can help to decrease potential production start-up and/or change costs for existing irradiation facilities and can help to result in a broader distribution of isotope production centers. In addition, automated or manual embodiments of the system 200) may easily be installed into existing equipment or installations. For example, embodiments may be incorporated into existing irradiation facilities which may include any component of or combination of equipment to perform and/or support a reactor, accelerator center, target/product chemical processing equipment, etc. Embodiments can be combined with particle accelerators or reactors to produce desired isotopes. Accelerators and reactors produce different isotopes for a range of diagnostic and therapeutic medical applications as well as industrial usage. Through adjustments of the target material, its morphology, and the separation chemistry, embodiments can be tuned to produce a range of product isotopes for medical applications in the same facility or similar facilities to those existing. Embodiments can incorporate existing chemical automation tools. These automation tools can be applied to the separation of the produced radionuclide product from the target, purification of the separated radionuclide, and any preparation and/or regeneration of the source prior to re-irradiation. The final radioisotope product of the system 200 can be integrated into existing generators. These generators can be distributed to medical facilities to provide radionuclides for medical applications. The processes used by the separation system 208 can be selected to regulate the chemistry and solution conditions of the produced isotope to meet desired conditions for generator use. Because radioisotope products are time-sensitive due in part to half-life limitations of the produced species, producing product near its preparation or ultimate use location can increase their availability for medical or other applications. Additionally and/or alternatively, the ability to automate separations and isotope production can increase production rates and help to decrease potential worker close. The high activity of targets and/or waste may result in radiation closes to workers involved in their handling. Thus, embodiments of the system 200 and method 10 can couple shorter irradiation time with automation for separation and/or reformation and reduce waste processing with target re-use, thereby helping to decrease the potential worker close due to material handling. In the embodiment illustrated in FIG. 2, the conveyor 206 is the conveyance system that physically moves the sources from the neutron bombardment chamber to a different location for interaction with one or both of the insertion component 210 or the extraction component 212. Conveyance system 206 may be open to the environment of other components of the system 200 or alternatively enclosed and possibly shielded to reduce radiation emissions around conveyance system 206. For example, as illustrated in FIG. 2 a portion of the conveyance system 206 such as a conveyor belt may physically move through some or all of the other components and systems. Alternatively, the conveyance system 206 may simply transfer containers 214 between the various components and systems, each of which is provided with its own container handling mechanisms for receiving the containers from, and returning them, to the conveyance systems. In an alternative embodiment, other conveyance systems than conveyors may be used such as robotics, or any other suitable container handling or transfer system including without limitation belts, chutes, diverter gates, bucket elevators, pneumatic conveyances, screw conveyors, etc. The conveyor 206 may be operated in a semi-continuous fashion (e.g., periodically pausing while containers are being acted on by a system or component) or a continuous fashion. In an alternative embodiment, the system 200 may produce the product through a batch irradiation followed by a batch separation, for example, irradiation of multiple sources and/or containers as a batch. The irradiated containers may be processed for product extraction either serially or in one or more batches or sets of containers. Although FIG. 2 shows a substantially continuous irradiation followed by a batch separation of the individual containers, any combination of batch or substantially continuous irradiation and batch or substantially continuous separation may be used as appropriate. Many different configurations of the targetry coupled separation system are possible and all are considered within the scope of this disclosure. For example, in an embodiment, the conveyor 206 may be eliminated in favor of a manual transfer operation. In this embodiment, operators manually or by remote control move the sources between the various components of the system 200. In yet another embodiment, the various components of the system 200 are designed so that the source is not moved, but rather the different components interact with a stationary source at different times during the process. In yet another embodiment, one or more sources are fixed inside a mobile neutron bombardment chamber 204 and the chamber is moved between a neutron generator and a separation system 208. FIG. 3 illustrates an embodiment of a method for selectively generating a desired radioisotope using targetry coupled separation. The method 300 begins with the selection of the radioisotope to be created. This is illustrated by a selection operation 302. In the selection operation 302 any radioisotope may be selected, for example from Table 1 above, such as 99Mo, 238U, 131I, 51Cr, 227Ra, 223Ra, 227Ac, etc. that the operator ultimately wants to obtain. In an embodiment, more than one radioisotope may be selected. As already noted, some desirable radioisotopes may not be direct products of an irradiation operation. In those situations, the selection operation 302 may be equally considered a selection of the decay chain or a selection of any of the radioisotopes in the decay chain. For example, to obtain 223Ra one may wish to create 223Ra generator out of 227Ac, as is known in the art. However, for the purpose of this disclosure, the term ‘selected radioisotope’ refers to the radioisotope that is the direct product of the irradiation of the target in the irradiation operation, and the selected radioisotope can be processed (including providing a holding time for anticipated decay) as necessary to ultimately generate the desired product. For example, if one wished to use targetry coupled separation to ultimately create 223Ra for medical use, the selected radioisotope or direct product would be 227Ac, e.g., for subsequent incorporation into a 223Ra generator, the selected radioisotope would be 227Ac. Likewise, if one wished to use targetry coupled separation to generate 99Mo for subsequent incorporation into a 99mTc generator, the selected radioisotope would be 99Mo (because it is a direct product). However, in this instance, there are also many direct products with atomic number 99 that are parents of 99Mo, by one decay chain or another, and that relatively quickly decay into 99Mo. These direct product parents include: 99Nb which decays into 99Mo and has a half-life of 15 seconds; 99Yr which decays into 99Nb and has a half-life of 1.47, and 99Zr which has a half-life of 2.2 seconds and decays into 99Y, to mention just a few in one particular decay chain. Thus, to use targetry coupled separation to generate 99Mo for a 99mTc generator, in an embodiment the selected radioisotopes may include some or all those direct products with atomic number 99 that decay into the desired product of 99Mo. After the radioisotope or isotopes are selected, a target material is identified from which the selected radioisotope(s) can be created through irradiation. This is referred to as the target identification operation 304. The target identification operation may further include identifying the overall source material (target and the ancillary material) including the physical properties of the source. In this manner, the target material identification operation may be referred to as the source material identification operation. For example, if 99Mo is a desired product and 99Mo and its atomic number 99 direct product parents are the selected radioisotopes, then one suitable target material may be created from 235U, such as an oxide of 235U or pure 235U metal, from which 99Mo can be obtained directly and indirectly through neutron bombardment. Many radioisotopes may be obtained from different compounds. e.g., from 235U or 239Pu and a combination of compounds may be selected as the target material. The identified target material may include any fissionable material, or combination of fissionable materials, or other isotopes suitable for production of desired isotopes by nuclear reactions, and may be selected based on the type of radiation generator, bombardment chamber, spectrum of the reactor (thermal or fast), and other equipment available. For example, the target may incorporate any known material which can be fissioned with a neutron to create the direct, selected radioisotope product and/or absorb a neutron to create the selected radioisotope product. The target material may include, but is not limited to, a uranium-based material, a plutonium-based material, or a thorium-based material. For instance, a target material may contain 235U. In another instance, the target material may contain 239Pu. Further, it should be recognized that the target material need not be fissile directly upon fabrication, but rather could be or include a fertile material that could be converted into a fissile material through neutron absorption. For example, the target may include any known nuclear fertile material which can be bred up through neutron absorption to the selected product and/or bred up and then fissioned to create the selected radioisotopic product. Fissionable material includes any nuclide capable of undergoing fission when exposed to low-energy thermal neutrons or high-energy neutrons. Furthermore, for the purposes of this disclosure, fissionable material includes any fissile material, any fertile material or combination of fissile and fertile materials. The identified target material may not be fertile or fissile. For example, 232Th may be used as a target material, which may be exposed to neutrons to yield the isotopes 225Ac or 227Ac. The isotope 226Ra is another example, which when exposed to protons may also generate 225Ac. Yet another example is using 153Eu as a target material, which when exposed to fast neutron (i.e., kinetic energy above 1 keV) radiation yields 153Sm. A further example includes using 14NH3 as a target material, which when exposed to gamma rays may undergo a photonuclear reaction to generate 13NH3. The target material may include one or more metallic target materials, such as, but not limited to, a substantially pure metal target material, a metal alloy target material, or an intermetallic target material. For example, a pure metal target material may include, but is not limited to, 233U, 235U, 239Pu, and/or 232Th. In another example, a metal alloy target material may include, but is not limited to, uranium-zirconium, uranium-plutonium-zirconium, uranium-zirconium-hydride, thorium-aluminum, or uranium-aluminum. By way of a further example, an intermetallic target material may include, but is not limited to, UFe2 or UNi2. It should be recognized that the above list of suitable metallic target materials for inclusion in a target is not exhaustive and should not be interpreted as a limitation but rather merely as examples. In another embodiment, the target material of a source may include one or more ceramic target materials, such as, but not limited to, an oxide target material, a nitride target material, or a carbide target material. For example, an oxide-based nuclear material may include, but is not limited to, uranium dioxide (UO2), plutonium dioxide (PuO2), or thorium dioxide (ThO2). Moreover, an oxide-based target material may include a mixed oxide target material, such as, but not limited to, a mixture of PuO2 and depleted or natural UO2. In another example, a nitride-based target material may include, but is not limited to, uranium-nitride or plutonium-nitride. By way of a further example, a carbide-based target material may include, but is not limited to, uranium carbide. It should be recognized that the above list of suitable ceramic target material materials for inclusion in the target material should not be interpreted as a limitation but rather merely as an illustration. In an embodiment, the target material identification operation 304 includes the determination of the complete compound or combination of compounds for the source material. It should be recognized that, in addition to the fissionable materials described above, the source material may also include ancillary material, which in some cases may include portions of non-fissionable material, such as, but not limited to, radiation-inert material, neutron moderating material or neutron reflective material. Such non-fissionable material may be provided to add strength, form, structure, or other properties to the target that could not be easily achieved using fissionable material alone. It should also be noted that, in an alternative embodiment of the targetry coupled separation method (not shown), the target identification operation 304 may precede the radioisotope selection. This embodiment may occur in situations where the target material is provided and not substitutable. In this embodiment, the owner of the target material may wish to use targetry coupled separation on the provided target material in order to extract some valuable radioisotopes from the target material in lieu of or prior to simply disposing of the target material. For any given solid selected target material, a recoil distance of the selected radioisotope(s) may be determined in a recoil distance determination operation 306. When a nuclear reaction occurs that coverts an atom of fissile material into a radioisotope atom, kinetic energy is imparted to the radioisotope atom. The amount of kinetic energy imparted varies based on the initial kinetic energy of the neutron, the atomic mass of the fissile atom, and the atomic mass of the direct product radioisotope, among other things. This kinetic energy causes the selected radioisotope(s) to recoil, i.e., move relative to the initial position of the fissile atom undergoing the nuclear reaction in the source material. The term recoil distance refers to the average distance or range of distances which a specific radioisotope is expected to move based on the imparted kinetic energy. Because many nuclear reactions have been well characterized, the kinetic energy and/or recoil distance can often be calculated or has been determined empirically for many given combinations of nuclear chemistry and neutron generator. For example, the recoil distance of fission products in uranium dioxide is generally described in S. G. Prussin et al., “Release of fission products (Xe, I, Te, Cs, Mo, and Tc) from polycrystalline UO2.” Journal of Nuclear Materials, Vol. 154, Issue 1 pp. 25-37 (1988), the recoil of fission products in thorium metal is generally described in C. H. Fox Jr. et al., “The diffusion of fission products in thorium metal,” Journal of Nuclear Materials, Vol. 62, Issue 1 pp. 17-25 (1976) and the migration of gaseous and solid fission products in a uranium-plutonium mixed oxide fuel is generally described in L. C. Michels et al., “In-pile migration of fission product inclusions in mixed-oxide fuels,” Journal of Applied Physics, Vol. 44, Issue 3 pp. 1003-1008 (1973). Such references allow one of skill in the art to estimate the recoil of selected radioisotopes for a particular system. The recoil distance determination operation 306 refers to calculating, estimating or otherwise identifying the expected recoil distance for the selected radioisotope within the selected target material. In an embodiment, the recoil distance determination operation 306 takes into account the density of the target material, the particulars of the neutron generator and other aspects of the system design. The recoil distance may be determined empirically from prior experiments or may be estimated using known characteristics of the materials and atoms involved, such as the atom number of the direct irradiation products. The range of any particle in material can be found with the stopping power, which is the relationship between a particle's kinetic energy and the range in material. For the production of radioisotope products, the energy can be due to the recoil from the decay route, as in fission or alpha decay, or the nuclear reaction, as in fast neutron or accelerated particle bombardment. The product isotope's energy will need to be determined based on its production route. A number of routes are known and data are available to assess the distance an energetic particle can travel through material. The Bethe-Bloch formula provides the energy loss of particles traveling though material in units of energy distance squared per unit mass, an example is MeV cm2 g−1. Stopping power and range tables are available from numerous references, e.g., from the International Atomic energy Agency and the National Institute of Standards and Technology, that can provide data to assess the recoil range, including continuous-slowing-down approximation, of produced isotopes. The units for ranges and stopping power can be the same as the Bethe-Bloch formula, or as a range in mass per area, such as g cm−2. Programs are also available that provide ranges and stopping powers for ions in materials (see, e.g., the SRIM software package available from Dr. James F. Ziegler). Once ranges or stopping powers are obtained, the distance a particle will travel in material can be estimated using the material density and the particle energy. If the data for a specific product or nuclide cannot be found, relationships between energy loss, velocity, and charge can be used. The recoil distance is then used in a source manufacture operation 308 in order to design and create a source that, for the particular combination of selected target and source material, preferentially results in radioisotopes distributed within the source material after the reaction so that the radioisotopes are more readily available to the extraction material than would occur in a bulk solid, or non-porous source. Specifically, the solid portions, i.e., the pore walls, of a porous source material (such as foams, particles, and the like) may be sized to be substantially similar to the recoil distance of the selected radioisotope product. In this manner, the anticipated recoil of the selected product can be used to improve placement of the product near an available surface of the source material to improve extraction of the product from the source (e.g., dissolution and extraction of the product without dissolution of the target). The term ‘available surface’ is used to describe a location, on or near a surface of a solid source material, from which the extraction material can obtain the product. In cases such as a source material formed as a foam or other porous structure (e.g., manufactured pores), the structure of the source material forming the pores (e.g., the walls of the pores) may be selected and formed to have a thickness substantially similar to the recoil distance of the selected radioactive product. In cases such as particles, one-half of the particle size or the particle radius may be sized to be substantially similar to the recoil distance of the selected radioisotope. In another example, in an embodiment in which the extraction material is a liquid, the available surface of the source material is a surface that the liquid can access during the separation process without having to alter the physical properties of the target material. In some cases, the available surface may include locations that are not physically on a surface of the source material, but that are close enough to an accessible surface that the extraction material can still obtain the product atoms, such as through diffusion. Thus, targetry coupled separation exploits the recoil from the nuclear reaction used to produce the selected radioisotope to simultaneously make that radioisotope more easily recoverable in the separation operation. In the source manufacture operation 308, the selected source material is formed into a source based on the recoil distance of the selected radioisotope. For example, in an embodiment the source material is formed into solid grains and the size of the grains is selected based on the recoil distance of the desired radioisotope. As a further example, if 99Mo is the selected radioisotope product (with an anticipated decay to 99mTc, the desired product) and the selected target is an oxide of 235U, then in an embodiment a source material including grains having an average particle size (such as diameter or average width) of equal to or less than two times (2×) the recoil distance of the 99Mo product but greater than 10% of the recoil distance. In another embodiment, the average particle size may be selected to be within ±50% of the recoil distance (0.5-1.5×) of the 99Mo product and, in yet another embodiment, the average particle size may be selected to be ±50% of half (0.25-0.75×) the recoil distance of the 99Mo product. In another embodiment, the average particle size may be selected to be within ±50% of the twice the recoil distance (1-3× the recoil distance of the selected radioisotope). In situations where there are more than one selected radioisotope each with a different recoil distance, the recoil distance used for sizing may be selected from that of any one of the selected radioisotopes, an average of the recoil distances of some or all of the selected radioisotopes, or a weighted average based on the expected yield of the selected radioisotopes. In an alternative embodiment, grain size of less than 20 micrometers may be used. In yet another embodiment, a grain size between about 0.1 to 10 micrometers may be used. Generally fission products have a recoil range of around 10 microns in UO2. For solid source embodiments, the processing of the grains of source material into a solid, porous solid may include any suitable processing technique including one or more of sintering, milling, sieving, 3D printing, crystallizing, precipitating, or heating the grains of target material. The solid source may take any high surface area form such as an open-cell foam, an open lattice, an open framework, a ceramic, a cloth, a thin film, a monolayer, a sponge, a nanocage, or a nanocrystal. The nuclear reactions can also induce chemical changes that can be used for selective separation. Such induced chemical changes are called hot atom chemistry and described in the literature. In hot atom chemistry, the nuclear reaction changes the chemical form of the reaction product compared to that of the target. The difference in chemistry between the target and reaction product, and the morphology of target, permit a separation of the reaction product without target destruction. As an example, a target could be a compound in a high oxidation state. Upon reaction with a neutron, the new isotope undergoes reduction and has different chemical properties than the target even though it is the same element as the target. The target morphology permits a separation of the product with the lower oxidation state without the need to dissolve the target. Additional detail regarding embodiments of targets, source materials, and the source manufacture operation 308 are discussed with reference to FIG. 5, below. The source manufacture operation 308 may include further selecting, creating and/or providing a suitable container for the source material. For example, in an embodiment in which neutrons are the form of radiation used, the container may be made of a neutronically-translucent material, so that neutrons are capable of passing through the container. A container may be in any suitable shape and form and may be provided with one or more valves for allowing the easy introduction and/or removal of the extraction material. In the embodiment shown in FIG. 3, after the source or sources have been created, the sources are exposed to neutrons for some irradiation period in an irradiation operation 310. This operation 310 may include transporting the source(s) to the irradiation facility/equipment for safe irradiation, for example by conveyor belt as described above. In the irradiation operation 310, source material is exposed to neutrons, thereby causing at least some atoms of the source material to undergo nuclear fission or neutron capture to create atoms of the selected radioisotope. This results in an irradiated source material that contains some amount of the selected radioisotope product within a reduced amount of unreacted target as discussed with reference to FIG. 11. In addition, because of the recoil from the fission reaction, at least some of the newly created atoms of the selected radioisotope move the recoil distance relative to the remaining, unreacted target within the source material. As described above, the recoil of the selected radioisotope product may make that radioisotope more available to the extraction material such as by making the radioisotope product closer to an available surface of the source material, which may then improve extraction by the extraction material. After the irradiation period, a separation operation 312 is performed, extracting atoms of the desired product or products from the source material. As mentioned above, the desired product may be the selected radioisotope, a decay daughter of the selected radioisotope, or, as is the case with 99Mo, both. This operation 312 may include transporting the source(s) to a separation facility/equipment, for example by conveyor system as described above. The operation 312 may also include introducing a storage or holding period before the separation to allow time for decay to occur. In an embodiment of the separation operation 312, the target in the source material is exposed to an extraction material such as a solvent that preferentially extracts the desired product from the source without substantially dissolving the remaining target in the source material. Embodiments of separation techniques are further discussed elsewhere in this disclosure, particularly with reference to FIGS. 1 and 7. In an embodiment, the remaining, unreacted source material is not chemically reactive with or affected by the extraction solvent. Specifically, it is not necessary to dissolve the target to recover some of desired product from the target. Thus, the target is not substantially dissolved nor is its physical phase altered by the separation operation 312. For example, in one embodiment the target is in a solid phase and remains in the solid phase throughout the irradiation and separation operations. In embodiments in which sources include a container, the source material may or may not be removed from the container during the separation operation 312. For example, in an embodiment, a source may comprise loose or packed individual loose grains of source material in a neutronically-translucent container, in which the grain size is based on a recoil distance of the selected radioisotope to be produced as discussed above. The source material grains may be repeatedly subjected to successive irradiation and extraction operations without removing the grains from the container. In this embodiment, a gaseous or liquid solvent may be flowed through the container or the container may be filled or partially filled with solvent and left in the container for some contact period of time after which the extraction material, now containing at least some of the selected radioisotope, is removed. In an alternate embodiment, rather than individual grains meeting some size requirement based on the recoil distance, the sources may include a solid mass of target. As discussed in greater detail below, such a solid target may be made by sintering or otherwise bonding individual grains (which may be tailored similar to that described above with respect to recoil distance) together to form a larger source material mass. Such a larger mass may be porous to facilitate penetration of a solvent into the porous mass, thereby facilitating contact with the generated radioisotope. The separation operation 312 may further include regeneration of the target to prepare it for subsequent irradiation. This may involve one or more washing operations to remove extraction material from the source material prior to subsequent irradiation. After a separation operation 312, the same source may be re-irradiated to create more of the selected radioisotope allowing the irradiation and separation operations 310, 312 to be repeated multiple times without substantially dissolving, changing the phase of, or removing any of the remaining mass of target material in the source. As discussed above, this allows the fissionable material to be more efficiently converted into the desired product that would be possible with a single neutron exposure. The method 300 further includes a final processing operation 314 that converts the extracted radioisotope product into a final product or final form suitable for commercial use. The final processing operation 314 includes separating the radioisotope from the extraction material and may also include additional processes to purify the radioisotope. The radioisotope may then be further processed into a final form suitable for transport and use as an industrial reagent or feedstock. In an embodiment, the final processing operation 314 includes incorporating the radioisotope into a daughter isotope generator. For example, the method 300 may be used to manufacture 223Ra generators made from the radioisotope 227Ac, 68Ga generators made using 68Ge, 99mTc generators made from 99Mo, and 82Rb generators made from 82Sr, to name but a few. Daughter isotope generators and methods for manufacturing daughter isotope generators from a parent radioisotope are known in the art. Any suitable method may be used. For example, 99mTc generators may be created from 99Mo in the form of the molybdate, MoO42−. To create the generator, the 99Mo molybdate is adsorbed onto acid alumina (Al2O3) substrate and placed in a shielded column. When the 99Mo atoms decay, they form 99mTc pertechnetate, TcO4−, which, because of its single charge, is less tightly bound to the alumina. Pouring normal saline solution through the column of immobilized 99Mo elutes the soluble 99mTc, resulting in a saline solution containing the 99mTc pertechnetate, with sodium as the counterbalancing cation. In an embodiment, the final processing operation 314 may be an automated or semi-automated process. As described with reference to FIG. 2, in an embodiment a targetry coupled separation system incorporates equipment necessary to separate the radioisotope from the extraction fluid, modify the radioisotope into the generator material necessary for use in a daughter generator (such as 99Mo bound to a substrate suitable for column chromatography), and package the material into the generator body in an automated or semi-automated process. Container FIG. 4 illustrates an example of a suitable container. The container 400 includes a cylindrical body 402 defining an internal cavity, a top or lid portion 404 that, when engaged, seals the cavity, and a bottom 406 which defines an interior chamber 414 that contains the source material. One or both of the top 404 and the bottom 406 may be removably attached to the body 402 to allow the source material to be inserted into or removed from the container 400. This may be achieved by any known system, such as corresponding threaded portions, for example on the lid portion and in the cylindrical body (not shown). Alternatively, the container 400 may be of a unitary construction and the source material charged through a sealable access port (not shown) or during the construction of the container. In the embodiment shown, two fluid flow valves 408, 410 are provided, first valve 408 (which may be an output value in some examples) in the top 404 and a second valve 410 (which may be an input valve in some examples) in the bottom 406. In yet another embodiment, the container may not be completely sealed when the lid is engaged, for example, to allow gas to escape or to allow the container to be immersed in the extraction material rather than having extraction material injected into the container through a valve or access port. Although valves 408, 410 are shown at the top and bottom of the container 400 respectively, one of skill in the art will recognize that the valves 408, 410 can be located in any appropriate location and/or orientation and do not necessarily have to be placed on opposing sides of the container. Alternatively, one valve 408 or additional valves (not shown) may be used for any of input, output, redundancy, and/or safety measures of the extraction material and/or container. A container may be of any shape, both externally and internally in the source material chamber. Any number, type, and configuration of access ports, valves, shackles, connectors, contact points, or other ancillary components may be used as desired. For example, in the embodiment shown a diffuser 412 is provided so that the container may be easily used as a fluidized bed or packed bed reactor. In the embodiment, the diffuser is in the form of a perforated plate with perforations sized such that the source material (such as the particulate matter) is prevented or reduced from passing through it. Solvent introduced from the bottom valve 410, however, passes easily through the diffuser 412 allowing contact with the source material. This is but one example of ancillary components that could be provided on the container. For example, many different fluidized bed reactor designs could be incorporated into a container having additional ancillary components such as additional diffusers, manifolds, baffles for distributing solvent flow evenly, non-cylindrical internal shape of the source material chamber/cavity 414, baffles for directing flow, etc. In an embodiment in which neutrons are the radiation used in the targetry coupled separation, the container may be neutronically-translucent as discussed above. Examples of suitable neutronically-translucent container materials include aluminum, zirconium, and molybdenum and alloys thereof as well as stainless steel alloys. Some or all of a container may be made from one or more of these neutronically-translucent materials. Containers may be made with an opening to facilitate the insertion and removal of the physical form of source material to be used. For example, when one or more large masses of source material are used as discussed above, a container may be provided with a relatively large opening that allow for the insertion and removal of the masses. This would allow containers to be reused after the source material is spent. Alternatively, a container may be constructed around the source material with the intention that the source material be disposed of with the container and no provision is made for removing the source material from the container once the target is sufficiently spent which may reduce waste and/or waste processing. Source Manufacture, Recoil Movement, Surface Treatment Improvement FIG. 5 illustrates an embodiment of a method of manufacturing a radioisotope-generating source in greater detail. As such, the method 500 represents an embodiment of the source manufacture operation 308 discussed above with reference to FIG. 3. In the embodiment shown in FIG. 5, the target material includes an oxide of a fissile or fertile material, such as thorium, uranium, or plutonium oxide. In various embodiments, targets in targetry coupled separation may include oxides which may be manufactured using any appropriate method, although many possible examples are provided below. Uranium oxides and plutonium oxides with suitable target properties have been prepared and characterized. Actinide salts may be dissolved in solution and precipitated to form solids. For embodiments using solid sources, any morphology may be used, although higher surface area morphologies will have a better recovery of product. Suitable high surface area morphologies include porous sources of: loose or sintered particles or powders; open-cell foams; 3D printed, milled, or crystallized open lattices or open frameworks; cloths; thin films and monolayers; sponges; ceramics; nanocages; and nanocrystals. Preferably, a solid source will have a surface area greater than 10 m2/g as measured by Brunauer, Emmett and Teller (BET) analysis. For embodiments that use a liquid source, the target material may be solid, such as solid particles, suspended in a liquid, such as in a colloid suspension. In the embodiment shown in FIG. 5, the method 500 begins with a dissolved salt of a fissionable material in a solution as a starting material, in providing operation 502. In an embodiment, the starting material may be created by dissolving and mixing chloride or nitrate salts of the appropriate fissionable material in purified water. In embodiments, the provided dissolved salt of the fissionable material may be in a solution such as an acidic solution, a basic solution, an aqueous solution, and an alcohol solution Next, a precipitant, such as sodium hydroxide, ammonium hydroxide, and/or oxalic acid, is mixed into the solution in a precipitant addition operation 504. The solution is maintained at the proper conditions for the precipitation to occur and the precipitate, an oxide of the fissionable material, is collected in a collection operation 506. Variations in the precipitation can include addition of ammonium hydroxide, peroxide, carbonate, or oxalate. Precipitation has been used to produce thorium, uranium, and plutonium containing oxides and is appropriate for other metal oxide formation. Any suitable method for precipitating a fissionable material oxide, now known or later developed, may be used. The precipitated oxide is then formed into grains in a grain forming operation 508. This may include milling, calcining, or sintering the precipitated oxide to form powders and/or pellets and/or any other suitable form of the target. For example, in an embodiment of the grain forming operation 508, the precipitate may be washed with acetone and purified water after collection, milled, and dried at 90° C. The dried precipitate can be milled again and redried. It can be again milled and/or then calcined up to 750° for 1 hour. The calcined powder can be milled and additionally or alternatively then cold pressed into pellets (of any appropriate size as determined based on desired properties of the source material such as the recoil distance of the selected radioisotope) for an appropriate time (which in some cases may be approximately 2 minutes) before being sintered. In an embodiment, sintering may be under a mixture of argon and 4% hydrogen for four hours at 1500° C. In an embodiment, the grain forming operation 508 may include a sizing operation to ensure either a particle size distribution of the grains and/or that the grains have a particle size less than some threshold size, such as a recoil distance. Sizing of grains to obtain a desired result is known in the art and any suitable method of sizing grains may be used, such as mechanical screening, filtration, and classification, electrical methods such as electrophoresis and electrostatic precipitation, and flotation. For example, suitable equipment for grain sizing, depending on the embodiment, may include sieves, gas or liquid elutriation columns; stationary screens; grizzlies; gyrating screens; vibrating screens; centrifugal sifters; cake filters; clarifying filters; classifiers; and crossflow filters. In some embodiments, after the precipitates are formed and sized, calcination of the product yields compounds suitable for sintering. Sintering time, temperature, atmosphere, and oxide preparation can be varied to produce suitable target properties as is known in the art. In embodiments, in addition to the sizes enumerated above with reference to FIG. 3, grains of target may be sized to have a maximum grain size of equal to or less than 10,000 nm (10 micrometers), or, alternatively, less than 1,000 nm, less than 100 nm, less than 50 nm, less than 10 nm, less than 5 nm or less than 2 nm. Furthermore, grains having a grain size ranging between 1 nm and 10 mm are anticipated to be particularly useful based upon recoil from nuclear reactions, fission, alpha decay, or beta decay. In an embodiment, the method 500 may be considered to include some of the operations of the method for selectively generating a desired product using targetry coupled separation illustrated in FIG. 3. For example, in an embodiment of the source manufacture method 500, the selected radioisotope operation 302, target identification operation 304 and the recoil distance determination operation 306 described with reference to FIG. 3 may be included in the method 500. In the embodiment shown, the method 500 includes a source formation operation 510 in which the grains are formed into a source material. This operation 510 is optional and not necessary in embodiments in which loose grains are used as the form of the source material. This may include combining grains into a solid mass to be incorporated into a source material, such as pelletizing the grains, making a ceramic from the grains, and/or making a solid matrix in which grains are incorporated. The use of nanoparticles in the preparation of ceramics may yield materials with desirable properties as a source for targetry coupled separation. The term nanoparticle refers to grains having a grain size less than 100 nanometers. Compared with traditionally produced ceramics, ceramics created from nanoparticles (nanoparticle-based ceramics) have greater hardness and higher yield strength. Nanoparticles of fissionable materials may be produced, for example, by the precipitation method described above. It is expected that ceramic properties derived from nanoparticles and nanoparticle synthesis routes will be useful in generating porous target material from tetravalent actinides. The following method for generating an actinide (e.g., U, Th, and/or Pu) ceramic is proposed. First, nanoparticles of the tetravalent actinide having a selected particle size are generated in an inert atmosphere. The particles are then sintered, for example at a temperature from 1,000-1,500° C., such as for example 1150° C. An inert atmosphere will be maintained throughout this process to prevent oxidation of the actinide metal. For example, an argon, neon, helium, nitrogen or any suitable inert gas mixture may be used. When this method was applied to zirconia nanoparticles as a surrogate for the actinide, after 2 hours of heating, a density of 93.5% theoretical was found. Density increased to 97.5% with 40 hours of heating and reached 99% at 60 hours. The average grain size was found to be 120 nm after 60 hours of heating at 1150° C. It is anticipated that actinide nanoparticles will have the same or similar properties and be suitable for use in targetry coupled separation sources. It is also expected that actinide oxide nanoparticles could be used to produce ceramic films and membranes and that such actinide oxide nanoparticles will have desirable properties for targetry coupled separations. The following method for generating an actinide oxide ceramic is proposed. First, actinide oxide nanoparticles are created. In an embodiment, this may be done by precipitating actinide oxide from a basic solution. In an alternative embodiment, actinide oxide particles may be synthesized by dissolving an actinide oxycarbonate in acidic nitric solutions, followed by hydrolysis and condensation of polynuclear actinide cations which should promote the formation of nano-sized, polymeric, oxy-hydroxide particles. When applied to zircon as surrogate, the zirconia particles so produced were found to be in the 3-6 nm range. Creation of actinide oxide nanoparticles may also be enhanced through various techniques. In an embodiment, alcohol may also be used as a solvent to generate actinide oxide nanoparticles. The alcohol solvent may induce a faster particle formation rate and produced submicrometer microspheres due to the low solubility of hydroxide species in alcohol solution. In yet another embodiment, adding polyethyleneimine and 2,3-dihydroxybenzoic acid in the precipitation phase may produce particles with a suitable particle size distribution. In yet another embodiment, oxalate precipitation may result nanoparticles within a suitable particle size. A hydrothermal technique may also be suitable for the synthesis of actinide oxide nanoparticles which may be a suitable form for source material in targetry coupled separation. Urea can be used in the synthesis of the nanoparticles produced by hydrothermal conditions. The general method for this technique is actinide precipitation under basic conditions at temperatures above 100° C. in a pressure vessel. Monoclinic nanocrystal nucleation and growth is expected to occur at 1200° C. from powders produced by forced hydrolysis. The particles so produced may then be sintered into a ceramic as described above. In yet another embodiment, near-critical water may be used to form actinide nanocrystalline materials. An aqueous mixture of actinide is brought to near-critical conditions and shock waves are produced by nozzle cavitation to generate actinide oxide particles. Near-critical water has been shown to rapidly hydrolyze and subsequently dehydrate cerium and zirconium salts to form mixed ceria-zirconia nanocrystalline materials. Shock waves produced by nozzle cavitation resulted in nano-sized particles of TiO2 and ZrO2. In yet another embodiment, an emulsion-combustion method may be used to generate actinide oxide particles. In this embodiment, actinide ions in an aqueous phase are mixed into a second, flammable phase to form an emulsion. The emulsion may then be burned which will result in the actinide ions being rapidly oxidized. In the emulsion-combustion method, zirconium ions in flammable solution were rapidly oxidized upon combustion. This method produced hollow, thin-walled particles of sub-micrometer size. In various embodiments, sources in targetry coupled separation may include metal-organic frameworks (“MOFs”). MOFs include coordination solids formed from linking metal ions with organic ligands. The high surface area compound can make suitable sources for the coupled production and separation of radionuclides. Lanthanide-based MOFs have been examined in more detail than actinide MOFs. Most of the actinide-based MOFs are based on the uranyl cation. Varying the combination of ligands and synthetic conditions has generated a large number of solid-state compounds. Molecular templates have been applied to uranyl MOFs. In various embodiments, source materials in targetry coupled separation may include monolayers and aggregates. Photochemical reduction of actinides in organic solvent has been used to produce actinide monolayers and aggregates such as particles of tetravalent uranium phosphates. The product morphology can be varied through treatment to achieve the desired grain size or characteristic lengths for frameworks other than grains. In one monolayer example, a uranium monolayer is formed through the interaction of a pulsed laser with uranyl in a tributyl phosphate organic phase. The monolayer presents as a distinctly different color and can be isolated from the organic phase. Aggregation occurs upon treatment with methanol. The uranium product properties can be tuned through coupling photoreduction parameters and monolayer treatment. In various embodiments, a target in targetry coupled separation may include uranium metal. High surface area uranium metal material can be prepared and used as targets, with or without a container, for specific radioisotope production. In an embodiment, a uranium metal ingot can be used as a starting form for uranium metal grain formation through arc melting. The arc melting parameters can be adjusted to produce metal grains with desirable properties such as a grain size selected based on the recoil distance of the selected radioisotope. Uranium metal can also form a high surface area structure through a hydriding-dehydriding process. In an embodiment, the Kroll process may be modified to produce a high surface area structure in the form of a porous actinide metal sponge from an actinide tetrachloride. The Kroll process involves the reduction of uranium chloride by liquid magnesium or sodium. Electrochemical reduction can also yield uranium metal which may have desirable target properties. After the formation operation 510, the source material may be placed in a container in a packaging operation 512. Packaging of fissionable material into containers has been discussed above. Containers have been discussed in detail above with reference to FIG. 4. Alternative Characterization of Grains FIGS. 6A through 6C illustrate a more detailed means for characterizing grain size than the typical approximation of a grain as a spherical particle with a characteristic diameter. The characteristic length 106 along at least one dimension of one or more grains 104 may include a characteristic length 106 along all dimensions of one or more grains 104 of the source material 100. For example, the grains 104 of the source material 100 may be engineered such that the “height”, represented by “a,” and “width,” represented by “b” are similar in size. Therefore, notwithstanding of factors (e.g., stress or thermal gradients), a radiation product subject to recoil upon creation may efficiently diffuse from the grain interior 110 to the grain boundary 112 along all directions within the grain. In this context, a grain structure may be characterized by the “grain size” of the grains 106 of the source material 100. The “grain size” may be selected such that the grains are small enough to allow for adequate diffusion from the interiors 110 of the one or more grains 104 to the boundaries 112 between the one or more grains 104. As shown in FIG. 6B, the characteristic length 106 along at least one dimension of one or more grains 104 may include a characteristic length 106 along a selected dimension of one or more grains 104. For example, as shown in FIG. 6B, the grains 104 within the source material 100 may be engineered to have a selected characteristic length 106 along a given dimension of the grains 104. For instance, in a grain 104 having an elongated grain structure, the grain may have a selected characteristic length along the “thin” dimension, shown as dimension “a” in FIG. 6B, of the grain 104. In another instance, in a grain 104 having an elongated grain structure, the grain 104 may have a selected characteristic length along the “thick” dimension, shown as dimension “b” in FIG. 6B, of the grain 104. It should be recognized that the grain 104 need only have at least one characteristic length 106 smaller than the distance required for adequate diffusion due to recoil from the interiors 110 of the one or more grains 104 to the boundaries 112 of the one or more grains 104. It is further recognized, however, that all dimensions of a grain 104 may have a characteristic length 106 smaller than or equal to a distance required for adequate diffusion of fission product 108 from the interiors 110 of the one or more grains 104 to the boundaries 112 of the one or more grains 104. As shown in FIG. 6C, the characteristic length 106 along at least one dimension of one or more grains 104 may include a characteristic length 106 along a selected direction 134. For example, the grains 106 within the source material 100 may be engineered to have a selected characteristic length 106 along a given direction in the source material 100. For instance, a grain 104 having an elongated grain structure may have a selected characteristic length 106 along a selected direction 134 within the nuclear fuel. It should be recognized that engineering the grain structures to have a characteristic length 106 along a selected direction 134 smaller than the length required for adequate movement of a radiation product due to recoil from a grain interior 110 to a grain-boundary 112 may supply a more efficient means for transferring fission product such as a radioisotope product from the grain interior 110. In another embodiment, one or more grains 104 may have a characteristic length 104 along a dimension of the one or more grains selected to maximize heat transfer from a grain-interior 110 to a grain-boundary 112. For example, the one or more grains 104 may be oriented such that their narrow dimensions, shown as “a” in FIG. 6C, are aligned substantially perpendicular to a thermal gradient 136 in the source material 100. Such an arrangement aids in the heat transfer from the grain-interior 110 to the grain-boundary, aiding in the diffusion of a fission product 108 from the grain interior 110 to its grain boundary 112. By way of another example (not shown), in a cylindrical pellet fabricated utilizing the source material 100 the grains 104 of the source material 100 may be arranged (i.e., on average the grains of the material may be arranged) to have the narrow dimension substantially perpendicular to the radial thermal gradient of the cylindrical pellet. It should be noted that the illustrations in FIGS. 6C, 6B, and 6A represent simplified conceptual illustrations of a plurality of grains 106 and should not be interpreted as schematic in nature. Further, it should be recognized by those skilled in the art that a variety of materials processing techniques (e.g., cold-working and/or annealing, compression, or extrusion) may be implemented in order to develop the symmetrical grain structure in FIG. 6A, and the deformed elongated grain structure illustrated in FIGS. 6B and 6C. A variety of materials processing techniques are discussed further herein. In another embodiment, the grains 104 of the source material 100 may have an average characteristic length 106 along at least one dimension smaller than or equal to a selected distance necessary for adequate diffusion of a fission product. For example, the grains 106 of the source material 100 may have an average characteristic length along a selected dimension or direction of the grains 104 of the nuclear fuel. It is recognized that there may exist a maximum average grain size which will provide adequate diffusion of fission products from the interiors 110 of the grains 104 to the grain boundaries 112 of the grains 104. In another embodiment, the grains 104 of the source material may have a selected statistical distribution of characteristic lengths. For example, the grains 104 of the source material 100 may have a grain size distribution having a selected percentage of the grains having a grain size below a selected distance. For instance, the source material 100 may have a grain size distribution such that 75% of the grains have a grain size equal to or less than 5 μm, with an average grain size of 3 μm. In another embodiment, the grains 104 of the source material 100 may have multiple statistical distributions of characteristic lengths. For instance, the source material 100 may have a grain size 106 distribution such that 25% of the grains have a grain size equal to or less than 10 μm, 25% of the grains have a grain size 106 equal to or less than 5 μm, and 10% of the grains are below 1 μm. In another instance, the source material 100 may have a grain size 106 distribution such that 25% of the grains have a grain size 106 equal to or less than 10 μm and 25% of the grains have a grain size equal to or greater than 50 μm. In another instance, the source material 100 may have a grain size distribution such that 25% of the grains have a grain size between 1 μm and 5 μm, 50% of the grains have a grain size between 5 μm and 10 μm, and 25% of the grains have a grain size 106 greater than 10 μm. Applicant's co-pending U.S. patent application Ser. No. 13/066,253, filed Apr. 8, 2011 4/8/11, titled Nuclear Fuel and Method of Fabricating the Same, which is hereby incorporated herein by reference, includes embodiments of nuclear fuel manufacture that could be used to create suitable target material for use in targetry coupled separations. Liquid Source Regarding various other embodiments of sources, a liquid source material can be employed and may be coupled with continuous separation to provide radioactive isotopes. As noted above target destruction can be reduced by limiting the phase change of the target (e.g., a liquid target with or without solid or other phase ancillary materials or a suspension of solid target in a liquid phase source material) through separation. In this embodiment, a liquid source can be a molten salt or solution phase. The liquid source can flow through an irradiation location or may be contained in a container that is passed through the irradiation location. The resulting radionuclides produced from a liquid source can be separated, isolated, and purified from the target using conditions and automation procedures as described for the solid source. Such separation may use a liquid-liquid extraction process, a liquid-gas extraction process, an electrochemical process, or, alternatively, a liquid-solid extraction process such as passing the irradiated liquid source over a solid material adapted to remove the desired product(s) from the liquid phase. For example, in a liquid-liquid extraction embodiment, under certain conditions the target may be immiscible in or otherwise separable from the extraction material to facilitate separation of a liquid extraction material from the liquid source material after a sufficient contact time. A liquid source embodiment may have similar benefits from target reuse and waste reduction, but the source configuration and flow may entail additional considerations than sources of solid material. A liquid fuel recycle system for removing fission products from salt-based fuels and recycling the fuels back to the reactor may be chemically similar to the process developed for metallic fuels. Supercritical CO2 separation, in particular, takes advantage of the properties of the salts, which are, by themselves, insoluble in sCO2. Extractants, such as diketones, may be used to draw select metals into the sCO2 phase as described herein. Physically, the liquid fuel recycle system may be made to avoid pressurization of the reactor vessel during a leak in the sCO2 system. Additionally, the salts in their liquid states may be at temperatures high enough to dissociate or degrade the diketones. To avoid both of these obstacles, a liquid fuel recycle system may be designed such that the molten-salt is pumped external to the reactor vessel and injected into a vessel containing the sCO2. A sCO2 system may be maintained at a temperature low enough to solidify the molten-salt, resulting in a high surface area solid. Provided the sCO2 can be maintained at a sufficiently low temperature, the beta-diketones or other appropriate extractant(s) may be co-mixed with the sCO2 during salt injection, avoiding dissociation. Alternatively, the extractant may be injected into an extraction vessel in a batch-wise fashion following salt injection. In either case, the result is a salt solution of (selected) metal-complexes solvated in the sCO2 diketone solution. The salt solution may then be pumped to a secondary system where temperature or pressure is adjusted to remove the metal complexes (product) from the salt solution without substantial destruction of the target in the molten salt fuel. Again, it is likely that the metal complex is removable from the salt solution without dropping the CO2 to a gaseous state (below the critical point) via heating, cooling, or both. Heat may be used to volatilize the metal complexes so that a separate gas phase occurs within the sCO2 solution. The sCO2 may alternatively be cooled or heated near and above the critical point where its solubility typically changes significantly with changes in temperature and pressure, resulting in a separate, liquid-metal complex phase which was forced out of solution due to changes in thermodynamic condition. This phase can then be transferred, such as by way of pumping, from the extraction system to a system designed for interim or long term storage. Whether further heating or cooling is used to separate the metal complex or other product, ultimately further heating can be used to thermally decompose the diketones, leaving behind the metal fission product(s). Separations of Radioisotope(s) from a Source Embodiments suitable for use in one or more of the separation operations described above will now be described in greater detail. As discussed above, embodiments of the separation of the desired product(s) from a source may include exposing at least some of the source material to an extraction material that preferentially extracts the selected radioisotope product from the source material without removing substantially any of the target or requiring the target to be dissolved or to otherwise require a change in the phase or physical form of the target. This allows the target to be reused in a subsequent neutron bombardment with little or no regeneration or post-separation processing. In an embodiment, the separation process generally involves the preferential isolation of the desired product(s) created by neutron bombardment from a solid phase source material. The separation is performed without dissolution of the source material or the target within the source material using a solvent as the extraction material. As mentioned above, the targetry coupled separation can exploit the recoil from the nuclear reaction used to selectively tailor and produce the target nuclei and target material to make the desired product more easily recoverable in the separation operation. Additionally and/or alternatively to the recoil, chemical differences between the target and the product nuclei can be selected, tailored and/or exploited to achieve a preferential separation of the desired product(s). Additional steps may be desired to remove the extracted radioisotope product from the extraction material and, in subsequent steps, further purify the desired product. Additional purification may utilize any one or more appropriate techniques as are known in the art, including column chromatography, precipitation, electrochemistry, ion exchange, sorption, filtration, and solvent extraction. Based on the source material composition, properties, and/or morphology, in various embodiments, the nuclear reaction may separate the product nuclei from the source or may physically move the product nuclei near or onto an available surface of the source material, which causes it to be more accessible to an extraction material, or may induce a chemical change that can be utilized to achieve separation. Appropriate extraction material can be selected, formed, introduced and/or activated in the separation process to exploit differences in the extraction product and the target, with the extraction product being either the direct (selected) product of the neutron bombardment or an indirect (decay daughter) product of the selected radioisotope. The desired product is amenable to separation due to the behavior of its chemical form in a solid, liquid, or gas phase. Additionally, the chemical processes do not appreciably dissolve the source or at least substantially reduce dissolution of the target, thereby leaving the target in a state to be reformed for further irradiation. Various separation treatment options are available. In various embodiments the target may be removed from the irradiation generator and treated. The treatment can use any single and/or combination of any appropriate process including chemical, electrochemical, thermal, filtration, pressure, fluidized bed, and gas phase methods. Solution phases can include any one or more phases including aqueous phases, organic phase, ionic liquids, molten salts, suspensions, and supercritical fluids. The chemical composition of the gas phase can also be varied in composition of gases, temperature, flow rates, pressure, etc. Illustrative separation processes and methods can include any one or more of members of the group comprising extraction, liquid chromatography, gas chromatography, capillary chromatography, crystallization, precipitation, filtration, distillation, fractional distillation, electrophoresis, capillary electrophoresis, magnetic separation, evaporation, flotation, cloud point, micellar, flocculation, electrochemical methods, volatilization, and sublimation. The separation process can be performed in the source's container, thereby precluding removing the target from the irradiation container. Alternatively, if there is no container, the source material may be placed inside a chemical reactor or other container and then removed for subsequent re-irradiation after the separation is complete. If desired, the separation process can utilize an automated chemistry system, such as those produced by Chemspeed Technologies, Skalar, Human Diagnostics, Randox or any other appropriate automated chemistry system. FIG. 7 illustrates an embodiment of general separation method suitable for use with targetry coupled separation. The method 700 obtains a desired product or products from a source material that has previously been irradiated so that at least some of the desired product is distributed throughout the source material. In the embodiment shown, such a source material is provided in operation 702. In a selection operation 704, an extraction material that removes the desired product or products from the source material without substantially dissolving the source material is selected and prepared based on the desired product to be removed and the characteristics of the source material. For example, in an embodiment, the extraction material may be a solvent that dissolves the desired product but does not dissolve fissionable material in the source material. In yet another embodiment, the extraction material may be a solvent containing an extractant, such as a ligand, that will bind to the desired product (thereby making it soluble with respect to the extraction material) but will not bind to fissionable material. If there are multiple desired products, one extractant may be suitable or, alternatively, multiple extractants may be selected. Such a ligand should be soluble in the solvent under temperature and pressure conditions of the contacting operation. In yet another embodiment, and as will be discussed in greater detail below, the solvent may be sCO2 and the selected ligand or ligands form a carbon dioxide soluble chelate with the radioisotope. Again, such a ligand should be soluble in the solvent under temperature and pressure conditions of the contacting operation. For example, for removal using sCO2, the ligand concentration may be up to 0.5 mole/liter and the temperature and contacting time may be varied. However, sufficient removal is anticipated to occur at temperatures below 220° C. at 1 atm with a contacting time of 30 minutes or less. Examples of possible ligands include a fluorinated β-diketone and a trialkyl phosphate, or a fluorinated β-diketone and a trialkylphosphine oxide. Further examples include dithiocarbamates, thiocarbazones, β-diketones and crown ethers. Inorganic ligands, including nitrates, sulfates, thiocynates, cyanates, and other similar compounds may also be used. A ligand may be provided with one or more functional groups selected to enhance the ligands ability to bind and remove desired products. Such functional groups include hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate, isothiocyanate, phosphate, and phosphono groups. Next, the source material is exposed to the extraction material, in a contacting operation 706 which in some cases may include adding the extraction material to the source material. Various actions may be performed to enhance the contact between the extraction material and the source material, again depending on the characteristics of the components involved. For example, if the source material is solid, the contacting operation 706 may include contacting the source material with a liquid extraction material for a residence time. As a result, an extraction material and radioisotope liquid mixture is created as the desired radioisotope product is dissolved from the source material. Alternatively, if the source material is a liquid, the contacting operation 706 may include contacting the source material with an immiscible liquid extraction material for a residence time. This results in a two-phase liquid mixture containing a first phase of bulk material and a second phase of extraction material with the dissolved desired product. The contacting operation 706 may also include other actions to assist in separation. For example, in an embodiment the contacting operation 706 includes agitating one or both of the source material and the extraction material during at least a portion of the residence time. In yet another embodiment, the contacting operation 706 includes changing a temperature of one or both of the source material and the extraction material during at least a portion of the residence time. And, in yet another embodiment, the contacting operation 706 includes changing a pressure of one or both of the source material and the extraction material during at least a portion of the residence time. In yet another embodiment in which the source material is in the form of solid grains stored in a container, the contacting operation 706 includes inserting an amount of the extraction material into the container and retaining the extraction material in the container for some predetermined residence time. After a selected residence time, the extraction material, now including the dissolved desired radioisotope product or products, is removed from contact with the source material in a removing operation 708. This may involve simply draining liquid extraction material from the source material or may require more active processing such as using centrifugal force, heating, cooling, pressurizing, or depressurizing to remove the extraction material. The desired product or products may then be separated from the extraction material and converted into a final product in a separation operation 710, substantially as described above with reference to final processing operation 314 of FIG. 3. Volatility-Based Separations It will also be appreciated that in various embodiments of an extraction process may include a rapid, volatility-based separation that can be used to isolate the desired products from irradiated sources. Volatility-based separation embodiments can exploit the formation of halides (F−, Cl−, Br−, I−), carbonyl (CO), and diketone based ligands such as hexafluoroacetylacetonate (“hfac”) (FIG. 9) to produce volatile metal compounds. The formation of volatile fluorides with nuclear materials is known. Chlorination has also been examined, and found similar to fluoride behavior. The existing differences can be exploited and extended to the other halides for tunable separations. This is readily performed in the Van Arkel process for obtaining pure Zr from ZrI4. Carbonyls are used in the Mond process to form volatile Ni species. The fission products Mo, Tc, Ru, and Rh also form carbonyl species, with Mo(CO)6 being a primary example of a volatile product. The hfac complexes are known to be volatile for a range of elements. This can provide a rapid and selective separation of radionuclides. Formation of halide, carbonyl, or hfac complexes can be exploited for the separation of a range of elements from starting material based on differences in volatility. Targeting the specific formation of volatile species can achieve separations that can be rapid and selective. An additional benefit is that volatile complexes may be used as metal vapor deposition precursors. Thus, a pure sample of the product could be obtained directly from volatile complexes formed in the reaction mixture. It is to be noted that decay of the product or product of the irradiation may require further processing for product generation (e.g., a product precursor is the result of irradiation of the source). In an embodiment, using the Mond process, an irradiated UO2-containing source material in a granular form having 99Mo solid distributed throughout the grains of the source material as the result of the previous irradiation may be exposed to carbon monoxide in a vessel, container or chamber maintained at a pressure from 0.5 to 5 atm and temperature from 50-60° C. According to the Mond process, this will convert at least some of the 99Mo to 99Mo(CO)6. Relatively more 99Mo(CO)6 may be created by extending the exposure time and by other methods such as agitating the source material to provide better contacting of the carbon monoxide gas with the surface of the source material. The boiling point of 99Mo(CO)6 (approximately 156° C.) is substantially less than the melting point of UO2 (approximately 2,865° C.). Therefore, volatilization can be easily achieved by raising the temperature of the source material after the carbon monoxide contacting operation to a temperature above the boiling point of 99Mo(CO)6. Furthermore, by keeping the temperature below the melting point of the UO2 after the 99Mo(CO)6 has been driven off, the source material is unaffected and ready for a subsequent irradiation operation. In another more generalized embodiment, an amount of irradiated source material having desired product distributed throughout the source material as the result of the previous irradiation may be exposed to a F−, Cl−, Br−, I−, CO, or diketone based ligands in a vessel, container or chamber under conditions that cause the desired products to form a volatile compound of the desired product, but that do not alter the target material. Relatively more volatile desired products may be created by extending the exposure time and by other methods such as agitating the source material to provide better contacting with the surface of the source material. Subsequently, as long as the boiling point of the desired product compound is below the melting point of the target material, volatilization can be easily achieved by raising the temperature. The remaining source material is unaffected and ready for a subsequent irradiation operation. Supercritical Carbon Dioxide Separations As mentioned above, another separation technology suitable for use in targetry coupled separations is supercritical carbon dioxide. The sCO2 extraction described herein may also be suitable for use in removing fission products from nuclear fuel in addition to removing desired products from targetry coupled separation sources. Supercritical CO2 has been examined for extraction on metals and metalloids from both aqueous and solid solutions. Accordingly, sCO2 combined with various ionic liquids (ILs) can be utilized as ligands to extract metal ions from solutions. Similar methods may be used to extract metals or metalloids from solid materials, such as contaminated paper, fabrics, or even soils. Current irradiated fissionable material recycling techniques using sCO2 solutions require dissolution of the irradiated material into a solution. Using the sCO2 separation techniques described herein, it may be possible to treat used fuel source material (including nuclear fuels considered for molten-salt reactors) with sCO2 in a manner which does not require dissolution. As an example, metal fuel from a breed and burn reactor such as a traveling wave reactor (TWR) can be treated with a sCO2 system that does not dissolve the U metal but does remove selected fission products (with high cross sections for parasitic absorption). A sCO2 system may be capable of selectively removing these elements and their corresponding isotopes. A list of elements soluble in ILs is shown in Table 2. TABLE 2Occurrence of selected elements inTWR spent fuel and IL solubility.FractionalFractionalElementAbsorptionElementAbsorptionPd2.38%Ru1011.18%Ru1.95%Pd1051.13%Sm1.25%Tc991.02%Mo1.21%Rh1031.02%Cs1.16%Pd460.73%Tc1.02%Cs1330.73%Rh1.02%Mo970.45%Nd0.85%Sm1490.43%Xe0.41%Ru1020.41%Eu0.30%Mo950.41% For ILs, the sCO2 may be useful as a means of introducing uranium into the IL. In other cases, it may be appropriate to have direct dissolution of oxides into an IL. Metals of interest to nuclear waste processing, such as actinides, lanthanides, and transition metals, have been characterized chemically using highly soluble fluorinated β-diketones in sCO2. Extraction can be accomplished by using appropriate chelating agents as extractants. For example, La and Eu extraction with greater than 90% effectiveness has been demonstrated using fluorinated diketones combined with tri-butylphosphate (TBP). In this process, a room temperature ionic liquid, an imidazolium-based 1-butyl-3-methylimidazolium (BMIM) with bis(trifluoromethylsulfonyl)-imide (also known as Tf2N−, which is properly described as (CF3SO2)2N−) was used as a complexing agent because of the complexing agent's ability to solubilize CO2. In this manner, a full water/RTIL/sCO2 system is developed. A similar process with other ionic liquids and metal chelating agents (extraction agents) and is summarized in Table 3. Note that Eu and La are both extracted with all systems except when using thenoyl tri-fluoroacetone (TTA) without TBP. The latter only extracted La while not separating (extracting) Eu. For example, for removal using sCO2, the ligand concentration may be up to 0.5 mole/liter and the temperature and contacting time may be varied. However, sufficient removal is anticipated to occur at temperatures below 220° C. at 1 atm with a contacting time of 30 minutes or less. The extractions performed in Table 3 were carried out with the extractant/sCO2 mixture at 150 atm for one hour at 50° C. The extractions show that sCO2 separation should be suitable for use on irradiated source material including nuclear fuel, nuclear waste material, and targetry coupled separation sources. Further, the extractions show that β-diketones can be used to selectively bind with oxides or metal in the presence of fissionable species such as uranium. Based on this information, it is anticipated that β-diketones can be used to selectively bind with radioisotope oxides or metals while not substantially dissolving fissionable material regardless of its origin. TABLE 3Degree of extraction (%) of EUIII and LaIII fromBMIMTf2N with different beta-diketones (with or without TBP).Eu3+La3+HFA w/o TBP90.590.4HFA w/TBP99.992.6TTA w/o TBP—87.1TTA w/TBP95.590.5HFA = hexafluoroacetylacetone,TTA = 4,4,4-trifluoro-1-(2-thienyl)-1,3-butanedione Further examples of possible ligands include dithiocarbamates, thiocarbazones, β-diketones and crown ethers. Inorganic ligands, including nitrates, sulfate, thiocynates, cyanates, and other similar compounds may also be used. A ligand may be provided with one or more functional groups selected to enhance the ligands ability to bind and remove desired products. Such functional groups include hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate, isothiocyanate, phosphate, and phosphono groups. In general, an obstacle to CO2 solvation is low solvent power of CO2 (non-polar). Metals and metal chelates have low solubility in sCO2 with CO2 solubility parameters in the range of 4-5 cal/cm3. This can be overcome by adding CO2-philic functional groups such as fluoroethers, fluoroacrylates, fluoroalkyls, silicones, and certain phosphazenes. Fluorinated beta-diketones (with and without tributyl phosphate) have been demonstrated in current techniques to extract a variety of metals. Bis(trifluoroethyl) dithiocarbamate exhibits higher solubility than non-fluorinated counterparts; 10−4 mol/L for fluorinated vs. 10−6 to 10−7 mol/L for non-fluorinated. As another example, Diethyldithiocarbamate (DDC) can be 3-800 times less soluble in sCO2 at 100 atm than bis(trifluoroethyl)dithiocarbamate (FDDC). Since sCO2 density change is nearly linear with pressure, the solubility also changes nearly linearly with solubility increasing with increasing pressure. Lanthanides, actinides, copper, arsenic, and antimony (and other products of irradiated sources) can have concentrations on the order of 10−4 mol/L CO2. Water and soil extraction has been demonstrated in current techniques with 1000-10000 molar ratio of chelate to metal in solution. In large scale processes, it may be impractical to transition sCO2 to the gas phase and remain economical since it may require either recompression of the CO2 to the supercritical state or a steady supply of high pressure CO2, not to mention the safety risk inherent to confining a high pressure solution of a highly compressible fluid. Furthermore, the off-gas CO2 may need to be collected in a container capable of further decontamination or disposal, due to some residual radioactive materials or decay products potentially remaining in the carbon dioxide gas. Some current techniques have a ‘back extraction’ process which does not require gasification of the sCO2 as part of the separation of the radioisotopes from the sCO2. In this type of process, metal or metalloid species are removed from solid or liquid solutions by using supercritical fluids to form a metal or metalloid chelate. The supercritical fluid will typically contain a solvent modifier, such as a few percent H2O or MeOH. The metals or metalloids are then back-extracted from the sCO2 solution by using an acidic solution, one which is preferably halogenated. By back extracting to another (aqueous) solution, decompression of the sCO2 is avoided. What is left is the other solution bearing the selected radioisotopes and sCO2 that can be readily reused. This is particularly advantageous in an automated system and in a continuous treatment, although even in a semi-automated, batch treatment system the ability to recycle sCO2 without the added step of repressurization would be cost-advantageous. Back extraction may, or may not remove the ligand with the radioisotope product. In an embodiment, fresh ligand may need to be added to the sCO2 before it can be reused as an extraction material. It should be noted that ILs could also be used for the back extraction process. FIG. 10 illustrates an embodiment of a method of extracting a first radioisotope product from irradiated fissionable source material. The method 1000 begins with an irradiated fissionable source material illustrated by the providing operation 1002. The irradiated fissionable target material may contain a plurality of radioisotopes in addition to the desired radioisotope product. Examples of desired radioisotope products include 99Mo, 238U, 131I, 51Cr, 225Ra, and 225Ac. Based on the desired radioisotope product or products and the characteristics of the target material, a ligand is selected in a ligand selection operation 1006. In an embodiment, a ligand is selected that is soluble in supercritical carbon dioxide (sCO2), forms a chelate with the desired product, and does not form a chelate with the target material. For example, in an embodiment, the desired radioisotope product is 99Mo, the irradiated target material is 235U and the ligand known to complex with molybdenum. Examples of other suitable ligands are provided above. Next, the identified ligand is dissolved into sCO2 to form a sCO2-ligand solution in an extraction material preparation operation 1006. If the selected ligand is not particularly soluble in sCO2, this operation 1006 may also include modifying the ligand to make it more soluble, such as by adding CO2-philic functional groups such as fluoroethers, fluoroacrylates, fluoroalkyls, silicones, and certain phosphazenes. In an embodiment, the ligand may be a fluorinated β-diketone and a trialkyl phosphate, or a fluorinated β-diketone and a trialkylphosphine oxide. In another embodiment, the ligand may be selected from dithiocarbamates, thiocarbazones, β-diketones and crown ethers. The sCO2-ligand solution is then placed in contact with the irradiated source material for a contact time, in a contacting operation 1008. As the selected ligand forms a complex with the desired product, a result of the contacting operation 1008 is a sCO2-radioisotope complex solution. In an embodiment, the irradiated source material is in a container and the contacting operation 1008 includes passing the sCO2-ligand solution through the container. The contacting operation 1008 may also include performing additional actions to enhance the mass transfer of the radioisotope product into the sCO2-ligand extraction material. For example, in an embodiment in which the irradiated source material is in the form of loose or loosely packed grains in a container, the contacting operation 1008 may include passing the sCO2-ligand extraction material through the container, essentially using the container as a packed bed reactor by forcing the solution through the bed of grains. In yet another embodiment, the sCO2-ligand solution may be passed through the container at a flow rate sufficient to fluidize the plurality of grains within the container, in effect using the container as a fluidized bed reactor. In yet another embodiment, the irradiated fissionable source material may be in liquid form and contacting includes agitating the fissionable material/sCO2-ligand solution mixture. After the contact time, the sCO2-radioisotope complex extraction solution is then removed from the irradiated source material in a removal operation 1010. In this operation, care may be taken to prevent the fissionable target material from being removed with the sCO2-radioisotope complex extraction material so that substantially all of the irradiated fissionable target material remains together in its original, physical form, e.g., a powder or ceramic. The reader will understand that a perfect system is not possible, and that some de minimis amount of irradiated material may be removed with the extraction material. However, systems in which less than 1% by weight or less than 0.1%, 0.01%, or 0.001% of the original amount of irradiated material is removed with the sCO2-radioisotope complex solution should be readily achievable. Next, the desired product and/or a further decay daughter product of the desired product is separated from the sCO2 in a separation operation 1012. This may be by back extraction of the sCO2 or may involve reducing the sCO2 to subcritical. This may include removing the ligand-product complex or, alternatively, may include removing only the product. In an embodiment, a back extraction is used in which a sCO2-ligand solution is also generated from the separation operation 1012 that is suitable for reuse without decompressing and repressurizing the sCO2 ligand solution. In an embodiment, this may be achieved by contacting the sCO2-product complex solution with an acidic solution, thereby generating an acid-product solution and a regenerated sCO2-ligand solution. Supercritical Carbon Dioxide Separation for Reformation of Spent Fuel Metallic fuel, including those metal fuels appropriate for vented pin configurations and/or a traveling wave reactor, typically includes metal fuel capable of high burn-up contained within vented, ferritic martensitic stainless steel cladding. At the end of life, the fuel generally has a highly porous matrix of metallic form fuel and solid fission products which precipitated from the fuel during the burn cycle. FIG. 8 illustrates an embodiment of a method for the reformation of nuclear fuel using sCO2. Reformation of fuel after irradiation generally may be designed to allow treatment of the entire fuel assembly for fission product, lanthanide, or actinide removal treatments without modification of the nuclear fuel assembly or fuel pins contained within. Using the example of a sealed vessel with targetry coupled separations, a previously burned nuclear fuel assembly source material may be placed into a sealable pressure vessel in a container operation 802. The vessel is then filled with pressurized sCO2 and one or more extractant (such as diketones, or any other appropriate agent) to create an extraction material in the absence of an IL or aqueous component in operation 804. Because of the presence of a vent in the existing fuel assembly for fission gas venting, and the nature of supercritical fluids, the sCO2-extractant solution will work to fill the fuel pin and the matrix of porous fuel (i.e. supercritical fluids behave as low surface tension, low viscosity fluids which fill the volume they are contained within). The extraction material will begin to solvate targeted fission products (or other materials, if so desired and a proper ligand chosen), leaving the uranium metal matrix unaffected. The fission products will then begin to diffuse out of the fuel source material such that the concentration of the overall system tends toward equilibrium. The extraction material containing the dissolved fission products can then be slowly released from the pressure vessel in an extraction material removal operation 806. New, clean extraction material may or may not be added to the pressure vessel during the removal operation 806. Agitation, heat and/or continued pressurization and depressurization may be applied to the system to enhance the solvation rate. For example, the system may operate at greater than 7.5 MPa (approximate critical point at 51° C.) and be oscillated by +/−0.1 MPa to enhance ‘pumping’ of extraction material in and out of the porous fuel. The extraction material removed from the system, containing the elements and isotopes removed from the used fuel, is directed toward another vessel in a collection operation 808. In the separation vessel, the sCO2 in the extraction material can be brought to below the critical point and converted to the gaseous phase in operation 810. By reducing the CO2 below the critical point, the extractant and the fission products are separated out of the CO2 and collect as a liquid phase in vessel. Next, a volatilization operation 812 can be performed on the collected liquid phase extractant and fission product mixture in which the extractant is brought to above its volatilization temperature and converted to a vapor phase, leaving behind the selected element or isotopes. This may be done in the same separation vessel as the sub-critical operation 810 or the extractant-fission product liquid mixture may be moved to a different vessel for this operation. Variations of this scheme may be used as appropriate. For example, lowering the solution to below the liquidus point of the carbon dioxide may be preferred if the chosen extractant and liquid CO2 are insoluble. Another alternative may be to raise the temperature of the supercritical solution to above the volatilization point of the extractant (e.g. greater than 100° C. to 200° C.) or to above the decomposition temperature (e.g. greater than 200° C. to 300° C.). In either case, the metal may substantially or partially precipitate from the sCO2 once the extractant is lost. Removal of the extractant vapor or decomposition product can be accomplished by a gas phase separation or, as above, by converting the CO2 to a liquid phase. Furthermore, the solution may change temperature or pressure from a first supercritical condition to a second supercritical condition, the second condition having a solubility of the extractant lower than the solubility of the first condition. By this process, all or a portion of the extractant may be recovered without leaving the supercritical state. Removing fission products from the fuel assembly may greatly enhance the disposability of the fuel assemblies, as >90% of targeted fission products may be removed with >90% capable of being removed with multiple sCO2 solution treatments. In some cases, it may be advantageous to apply multiple cycles such as repeated treatments or multiple different treatments, each with a different extraction material, to increase the removal of fission products. For example, in some cases, two treatments could give 99% removal of accessible fission products whereas three would give 99.9% and so forth. Any appropriate factors may be used to determine the number and/or type of processing treatments and may be based on fission products dissolved or stuck inside the solid fuel matrix where sCO2 solution cannot penetrate. It should be noted, however, that it may be possible to operate at temperature and timescales which would allow for diffusion of solution soluble metals out of the fuel matrix and into solution. This may lower the short term heat load of the spent fuel assembly, decrease the dangers of handling and transporting the assembly, and make it more suitable for long-term disposal. An alternative to spent fuel disposal would be to re-use the fuel assembly once the fissions products are removed as a source in a targetry coupled separation method, such as described above. The fuel assembly could be transported to a targetry coupled separation facility for this or processed in the same facility that created the spent fuel. A fuel assembly may be used as a source without modification or it may be processed to improve the targetry couple separation effects, such as by converting the spent fuel into grains of an appropriate size for the radioisotope products of interest to the targetry coupled separation facility. For example, in an embodiment the facility is a breed and burn type reactor such as a TWR. In this embodiment, the fission products may be removed and then a thermo-mechanical treatment is performed within the pressure vessel used for solvation. The thermo-mechanical treatment modifies the structural material for continued in-reactor use. To enhance the treatment, after the fission products are removed, the vessel and contained assembly may be brought to significantly higher temperatures (which could be made to exceed the fuel melting point) and pressures (10's of MPa's). A system using targetry coupled separation may remove fission products prior to the end of life by incorporating the separation process such as sCO2 process into the fuel management or ‘shuffling’ cycle to remove fission products periodically during irradiation (operation of the reactor). For example, some TWR re-fueling systems incorporate a sealed enclosure for raising the assembly out of the vessel. In such systems, the existing enclosure also contains cooling capability to manage assembly decay heat. These systems may be made more robust such that fission products may be removed, in containment, with minimal system modifications. This allows sCO2 extraction to be done as an integral part of the shuffling operation. Such a system would not require large vessels and piping, due to the high density of sCO2. Concentrations of greater than 10-4 kg metal/kg solution are possible. At end of life, each assembly contains the maximum amount of fission products, on the order of 50 kg. The solution density is on the order of 1000 kg/m3. Therefore only 5 m3 of sCO2 solution would be needed in some cases to contain all the fission products in a single assembly. Treating the assembly at more frequent intervals would obviously reduce this maximum volume. Furthermore, since the CO2 may be separated from the fission products and re-entered into the system, the inventory can be additionally reduced. Irradiated Material Reprocessing FIG. 12 illustrates an alternative embodiment of a method for selectively generating a desired radioisotope using targetry coupled separation. The method of FIG. 12 differs from that of FIG. 3 in that the irradiated target material is provided as the starting material, thus limiting the options of which desired products may be selected. This may occur, for example, when a quantity of spent nuclear fuel is available and it is desired to use targetry coupled separation to recover some value from the spent fuel. Such an example includes the production of 223Ra from 223U which is a waste product from the thorium fuel cycle. The method 1200 begins in operation 1201 with provision of an amount of irradiated source material, which may include some amount of both target and ancillary material, to be used in targetry coupled separation. The initial source material may be spent nuclear fuel, nuclear waste containing some amount of fissionable material, or some other material and may include any target material as described above. The initial source material is then characterized to determine what radioisotopes are within the material in a characterization operation 1202. The initial source material may or may not be suitable for targetry coupled separation without further processing and/or its incorporation into a source material. Thus, the characterization operation 1202 also determines if the form of the initial material can be modified to enhance the separation of any particular radioisotopes. A selection operation 1204, similar to that described above with reference to FIG. 3, is then performed. In this operation 1204, however, because the initial source material is known, the range of radioisotopes that may be selected is limited to those that can be obtained from the initial material. In an embodiment, more than one radioisotope may be selected. As already noted, some desirable radioisotopes may not be direct products of an irradiation operation. In those situations, the selection operation 1204 may be equally considered a selection of the decay chain or a selection of any of the radioisotopes in the decay chain. A material processing operation 1206 may then be performed. The initial source material is processed into one or more sources. In an embodiment where re-irradiation is to occur, this processing may be done based on the recoil distance of the selected radioisotope, as described with reference to FIG. 3. The processing operation 1206 may be as simple as placing the initial source material in a container. In another embodiment, the initial source material may be processed, physically and/or chemically, to make the form of the source material more suitable for the separation operation. For example, an initial material may be crushed and sieved to generate particulates having a selected particle size. As mentioned above, if the initial source material is to be re-irradiated, this sizing may be done based on the recoil distance of the selected radioisotope. Such processing may further include sintering the particulate into a ceramic, as described herein. Additional detail regarding embodiments of targets, source materials, and the processing operation 1206 are discussed with reference to FIG. 5, below. The processing operation 1206 may include further selecting, creating and/or providing a suitable container for the source material. In an embodiment in which re-irradiation using neutrons will occur, the container may be made of a neutronically-translucent material, so that neutrons are capable of passing through the container. If re-irradiation will not occur, then a container of neutron-absorbing material may be selected. A container may be in any suitable shape and form and may be provided with one or more valves for allowing the easy introduction and/or removal of the extraction material. After the source or sources have been created, a separation operation 1208 is performed, extracting atoms of the desired product or products from the source material. As mentioned above, the desired product may be the selected radioisotope, a decay daughter of the selected radioisotope, or, as is the case with 99Mo, both. This operation 1208 may include transporting the source(s) to a separation facility/equipment, for example by conveyor system as described above. In an embodiment of the separation operation 1208, the source material is exposed to an extraction material such as a solvent that preferentially extracts the desired product from the source without substantially dissolving the remaining target in the source material. For example, in one embodiment the source is in a solid phase and remains in the solid phase throughout the irradiation and separation operations. In the embodiment illustrated in FIG. 12, an optional re-irradiation operation 1210 may be performed. In that operation, the sources are exposed to neutrons for some irradiation period in an irradiation operation 1210. This operation 1210 may include transporting the source(s) to the irradiation facility/equipment for safe irradiation, for example by conveyor belt as described above. In the irradiation operation 1210, source material is exposed to neutrons, thereby causing at least some atoms of the source material to undergo nuclear fission or neutron capture to create atoms of the selected radioisotope. This results in a re-irradiated source material that contains some amount of the selected radioisotope product within a reduced amount of unreacted target as discussed with reference to FIG. 11. In addition, because of the recoil from the fission reaction, at least some of the newly created atoms of the selected radioisotope move the recoil distance relative to the remaining, unreacted target within the source material. As described above, the recoil of the selected radioisotope product may make that radioisotope more available to the extraction material such as by making the radioisotope product closer to an available surface of the source material, which may then improve extraction by the extraction material. In embodiments in which sources include a container, the source material may or may not be removed from the container during the separation operation 1208. The separation operation 1208 may further include regeneration of the target to prepare it for subsequent irradiation. This may involve one or more washing operations to remove extraction material from the source material prior to subsequent irradiation. After a separation operation 1208, the same source may be re-irradiated to create more of the selected radioisotope allowing the irradiation and separation operations 1210, 1208 to be repeated multiple times without substantially dissolving, changing or removing any of the mass of remaining target material in the source (except as a result of the fusion reaction). As discussed above, this allows the fissionable material to be more efficiently converted into the desired product than would be possible with a single neutron exposure. The method 1200 further includes a final processing operation 1214 that converts the extracted radioisotope product into a final product or final form suitable for commercial use, as described with reference to FIG. 3. In an embodiment, the final processing operation 1214 includes incorporating the radioisotope into a daughter isotope generator as described above. It will be clear that the systems and methods described herein are well adapted to attain the ends and advantages mentioned as well as those inherent therein. Those skilled in the art will recognize that the methods and systems within this specification may be implemented in many manners and as such is not to be limited by the foregoing exemplified embodiments and examples. In this regard, any number of the features of the different embodiments described herein may be combined into one single embodiment and alternate embodiments having fewer than or more than all of the features herein described are possible. While various embodiments have been described for purposes of this disclosure, various changes and modifications may be made which are well within the scope of the technology described herein. For example, targetry coupled separation may be adapted to remove fission products including poisons or other nuclear contaminants from sources made of solid nuclear waste. Numerous other changes may be made which will readily suggest themselves to those skilled in the art and which are encompassed in the spirit of the disclosure and as defined in the appended claims. |
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description | The present invention relates to a system for injecting mortar into a container. The invention relates in particular to introducing mortar into a drum containing harmful waste, in particular radioactive waste resulting from operations for conditioning material during the fabrication of mixed oxide (MOX) fuel (U,Pu)O2, and operations of decontaminating or dismantling a glovebox. Patents FR 2 605 788 and U.S. Pat. No. 5,246,287 describe an apparatus for introducing mortar into a drum containing radioactive waste. The apparatus comprises a receiver fitted with a mixer and into which water and the materials necessary for making a slurry are introduced. The apparatus includes a pump extracting the slurry from the receiver and delivering the slurry to the drum via ducts for conveying the slurry and including a three-port valve. A compressed air duct opens out into the valve, and a return duct connects the valve to the receiver. The quantity of slurry delivered by the pump is controlled by load cells fitted to the receiver, and excess slurry is sent to the receiver via the return duct. The compressed air serves to facilitate injecting the slurry into the drum. A drawback of that slurry injection method is that the compressed air used for injection is subjected to contamination and must subsequently be decontaminated. Furthermore, although that method is adapted to injecting a predetermined quantity of slurry into a drum containing waste, which drum must therefore present a volume that is known accurately, it is on the contrary ill-adapted to circumstances in which the volume of waste contained in the drum is poorly known. Another drawback of that method is that it does not make it possible, once the filling of the drum has been terminated, to be certain that no contaminated slurry or air has migrated to the three-port valve and to the apparatus as a whole. An object of the invention is to propose a device and a method for injecting mortar into a container containing waste, enabling the container to be filled accurately without knowing exactly the volume of the waste, by providing dynamic confinement. An object of the invention is to propose a device and a method for injecting mortar into a container containing waste, minimizing the quantities of the materials used (in particular air and mortar) that are subjected to contamination and that consequently need to be subjected to subsequent decontamination. An object of the invention is to propose a device and a method for injecting mortar into a container containing waste, making it possible to be certain, once filling of the container has terminated, that no materials that might be contaminated (in particular air and mortar) migrate to the device as a whole. An object of the invention is to propose a device and a method for injecting mortar into a container containing waste, facilitating emptying and cleaning of the ducts for conveying mortar. An object of the invention is to propose a device and a method for injecting mortar into a container containing waste, enabling the container to be filled accurately so as to be sure that the empty space at the top thereof is of substantially zero volume. An object of the invention is to propose a device and a method for injecting mortar into a container containing waste, improving and/or remedying, at least in part, the shortcomings and drawbacks of prior art devices and methods for injecting mortar. In an aspect of the invention, there is provided a method of injecting mortar into a container containing waste, the method comprising the following operations: causing a first stream of mortar to circulate continuously in a circulation loop; during the continuous circulation, extracting from the circulation loop a second stream of mortar that is smaller than the first stream of mortar; and introducing the second stream of mortar into the container containing waste. Thus, by extracting only a fraction of the flow of mortar circulating in the loop, any risk of introducing air into the mortar for injection is avoided. The invention also provides a device for injecting mortar into a container containing waste, which device comprises: a mortar circulation loop comprising a mortar storage receiver, a mortar transfer pump connected to the storage receiver, an outlet duct for conveying the mortar leaving the pump, and a return duct for conveying mortar to the storage receiver; and an injection duct extending the outlet duct. According to a characteristic of the mortar injection device of the invention, the circulation loop includes an extractor member connecting together the outlet duct, the return duct, and the injection duct, and the injection duct and the loop are isolated by a valve having a single passageway; this isolation valve is placed at the inlet to the injection duct, and the absence of a valve, other than an optional mortar flow regulator, in the outlet and return ducts of the loop makes it possible to ensure a continuous flow of mortar, the extractor member serving to extract a fraction of the mortar stream circulating in the loop and to introduce it into the injection duct. Preferably, the extractor member is in the form of a Y junction or coupling presenting three duct portions: a first duct portion and a second duct portion connected respectively to the outlet duct and to the return duct; and the third duct portion is placed (connected) tangentially to the first duct portion and is connected to the injection duct. For this purpose, at least one of the three duct portions is curved. In an embodiment, the section of the first duct portion is substantially the same as the section of the second duct portion, while the section of the third duct portion is less than the section of the first and second duct portions. In other words, and according to another aspect of the invention, a method is provided of injecting mortar into a container containing waste, the method comprising the following operations: mortar is caused to circulate under pressure in a circulation loop; at an extraction point of the circulation loop, mortar is extracted at a pressure that is sufficient to compensate for the head loss resulting from conveying the (extracted) mortar via an injection duct connecting the extraction point to the container; and the mortar extracted from the loop is introduced into the container containing waste in such a manner as to avoid introducing any propellant (solid, liquid, gaseous, or other) into the injection duct—and consequently into the container containing waste. To this end, in a device of the invention, the lengths and the diameters of the return duct and of the injection duct, and the through diameters of the members, such as valves, located in said ducts are selected in such a manner that the head loss in the injection duct, corrected for variations of position between the inlet and the outlet of the injection duct, is close to or less than the head loss in the return duct, corrected for variations in position between the inlet and the outlet of the return duct. Furthermore, and also for this purpose, the valve(s) fitted to the injection duct are selected to give rise to low head loss; the valve(s) is/are preferably selected from “full flow” valves, in particular from sleeve valves and plug valves. Preferably, the altitude position of the inlet orifice of the injection duct is higher than the altitude position of the outlet orifice of said duct, so as to encourage the mortar to flow under gravity along the duct. In other words, in yet another aspect of the invention, there is provided a method of injecting mortar into a container containing waste and including a first orifice and a second orifice, a first vessel secured to the container communicating with the container via the first orifice, and a second vessel secured to the container communicating with the container via the second orifice, the method comprising the following operations: extracting mortar from the mortar circulation loop; introducing the mortar extracted from the mortar circulation loop into the container; vibrating the container to facilitate flow inside it; and monitoring the appearance of mortar in the second vessel, and when said appearance is detected, ceasing to extract mortar from the circulation loop. It is thus possible to introduce into the waste container the quantity of mortar that is strictly necessary for filling it without there being any need to know this quantity in advance. To this end, in a device of the invention, the altitude positions of the first and second vessels are preferably similar (substantially identical); the respective capacities of these vessels may also be substantially identical. The device of the invention also preferably includes a sensor sensitive to the appearance of mortar in the second vessel, such as a radar sensor. In a preferred embodiment, each of the vessels presents an upwardly-flared shape, in particular an upwardly-flared frustoconical shape in order to facilitate subsequent unmolding. Preferably, after mortar has ceased to be extracted from the loop, the mortar contained in an injection duct connecting the mortar circulation loop to the first vessel (and to the container) is expelled into the first vessel so that it is subsequently possible to clean the injection duct prior to filling another waste container. To this end, in a device of the invention, the sum of the capacities, or the useful volume, of the first and second vessels is preferably not less than the capacity, or volume, of the injection duct. Also preferably, the mortar contained in the injection duct is expelled by introducing compressed air into the injection duct and then, after connecting the injection duct to a rinsing pot, causing a rinsing liquid such as water to flow in the duct so as to entrain and remove any mortar residue that might have collected on the walls of the injection duct. Also preferably, the second vessel is connected to a circuit for extracting and filtering air and the contaminated air that is expelled from the container while mortar is being introduced therein is extracted from the second vessel. To this end, a device of the invention may include a receptacle for collecting the rinsing liquid, a collector of shape adapted to the shape of the second vessel for collecting the gaseous effluents, essentially air, leaving the vessel, and a duct connected to the collector to deliver the effluents to a gaseous effluent decontamination circuit. After the mortar has dried, and shrunk while drying, it is possible to separate the two vessels and the “lumps” of mortar they contain from the waste container, and then to close both orifices of the container with stoppers. Other aspects, characteristics, and advantages of the invention appear from the following description given with reference to the accompanying drawings that show preferred embodiments of the invention having no limiting character. To ensure the present application is clear, the terms “receiver” and “hopper” are used to designate a container adapted to contain a sufficient supply of mortar to fill the space left empty by the waste placed in a waste container. For the same purpose, the terms “vessel” and “cone” are used in the present application to designate a container suitable for containing surplus mortar delivered to the waste container. Also for the same purpose, the terms “receptacle” and “capacity” are used in the present application to designate a container adapted to contain effluents resulting from cleaning the mortar injection system. Consequently, and unless stated explicitly or implicitly to the contrary, the term “container” is used in the present application solely to designate the container that contains waste. With reference to FIG. 1 in particular, the mortar injection system is intended to ensure that waste contained in a container 87 is locked in place. The mortar injection device comprises: a mortar circulation loop BA comprising a mortar storage receiver TM12, a positive displacement pump P11 for transferring mortar, which pump is connected to the storage receiver, an outlet duct CD for transporting the mortar from the outlet of the pump, and a return duct CR for transporting mortar back to the storage receiver; and an injection duct CI extending the outlet duct. The circulation loop includes an extractor member OP connecting the outlet duct, the return duct, and the injection duct together. The injection duct and the loop are isolated by a valve V1 having a single passageway and located at the inlet to the injection duct. The mortar for injection into the container is prepared and then stored temporarily in a hopper TM12 prior to being taken by pipework to a glovebox BAG in which the container 87 is located. The device has an injection duct CI fitted with a system of three valves V1, V2, and V3 adapted to injection and to rinsing the duct. The valves V1, V2, and V3 are sleeve valves or plug valves with full-flow. The injection duct terminates in an injection pipe CAI located in the glovebox and supported by a mechanism MD for moving the injection pipe, which mechanism is operated by an operator. In order to fill the container 87, the injection pipe CAI is inserted into a first vessel R1 referred to as a “filler cone” and fastened to a top wall 88 of the contain 87, which top wall is pierced by a first orifice 89 used for filling. Detecting when the container has been filled with mortar is performed via a second vessel R2, referred to as a “vent cone”, that is also fastened to the top wall 88 of the container that is pierced by a second orifice 90 serving as a vent and overflow. The first and second vessels R1 and R2 are secured to the container in register with the orifices 89, 90 provided through its wall 88, with the height positions of the first and second vessels being similar. The container 87 is set into vibration while the mortar is flowing in. The mortar is constituted by a mixture of sand, cement, and water, possibly having added thereto one or more additives, in particular a plasticizing agent. The mortar may present a density close to 2.25 kilograms per cubic decimeter (kg/dm3), fluidity measured using a Marsh cone close to 200 centipoise (cP) to 500 cP, and a duration of utilization before setting of no more than three hours. The mortar is prepared in a mixer (not shown) and then placed in the buffer hopper TM12 that presents a working volume that is sufficient to fill a container 87 that contains little waste. The mortar injection installation comprises three portions: a feed loop BA between the hopper TM12 and a member OP for extracting mortar into the loop; an injection duct CI between the member OP for extracting mortar and the glovebox BAG; and a device DAE for introducing air and water into the injection duct. The mortar injection installation serves to perform the following functions: causing the mortar to circulate around the loop BA and to flow along the duct CI leading to the cementing glovebox BAG where the container 87 for filling is located; filling one or more containers 87 per day, while guaranteeing the quality of the mortar injected into the containers 87; being capable of being emptied and rinsed simply, limiting the quantity of waste that is generated; avoiding mortar overflowing in the glovebox; and guaranteeing compliance with safety requirements associated with the danger level of the waste. The safety requirements are as follows: ensuring confinement between the roof of the container 87 and the ambient atmosphere of the treatment premises housing the device; ensuring confinement between the ambient atmosphere in the cementing glovebox BAG and the ambient atmosphere in the treatment premises; ensuring confinement relative to the outside; and recovering the suspect waste generated under confinement to avoid any dispersion in the treatment premises. For this purpose, it is useful to take account of the respective dimensions of the point OP where the mortar is extracted from the loop and the top 88 of the container, and the diameter and the length of the mortar injection duct, so that, at the desired injection flow rate, the head of mortar and the in-line head losses are in balance and the mortar can flow to the outlet of the injection duct at a pressure that is substantially zero, without said duct becoming emptied. Thus, for example, it is possible to determine the mean diameter of the injection duct as a function of these dimensions, length, and flow rate. Choosing a larger diameter could lead to the downwardly sloping portion of the duct CI being emptied under gravity, thereby putting the ambient atmosphere of the glovebox into communication with the pipework outside the confined zone, and also leading to a greater volume of contaminated mortar (when emptying and rinsing the injection duct). Choosing a smaller diameter would increase the risks of the duct CI becoming blocked and that would require a higher pressure for the mortar in order to cause it to flow. For the injection pipe, it is preferable to select a diameter that is adapted to the fluidity and the viscosity of the composition and to the slope between the extraction point and the high level of the container. The mortar is put under pressure and caused to circulate around the loop BA by a peristaltic pump P11, and it is transferred to the container by the injection duct CI that is connected to the loop BA via the member OP. Providing the mortar flows continuously in the feed loop (and thus providing the flow rate in the loop is greater than the injected flow rate), this makes it possible to have a loop that is filled and under moderate pressure at the level of the branching point OP of the injection duct; this makes it possible to restrict the volume of contaminated/suspect mortar to the volume of the injection duct, with the mortar that is present in the feed loop constituting waste that is conventional (i.e. non-suspect). The flow rate of the mortar flowing in the return duct of the loop BA may for example be about 10% of the flow rate of the mortar passing through the pump P11, with 90% of that flow rate that passes through the pump being extracted from the loop and injected into the container. The pressure of the mortar in the extractor member may for example be adjusted to a value of the order of about 0.5 bar to about 1 bar. The height dimension of the inlet orifice to the injection duct, i.e. of the member OP, is higher than the height dimension of the outlet orifice of said duct, i.e. of the pipe CAI, so as to encourage the mortar to flow in said duct under gravity. The injection duct preferably presents a downward slope so as to avoid the presence of any bottom point that might retain mortar or rinsing water or moisture. The confinement between the feed loop together with the rinsing device relative to the injection duct is provided by a motor-driven isolating valve V2; the confinement of the feed loop is provided by the two motor-driven isolation valves V1 and V2. The valves may be of the sleeve type that withstand abrasion (full flow when the valve is open, closure by flattening the membrane), with pneumatic motor drive. The duct segments may be made of stainless steel; flexible duct portions may be provided to connect both the peristaltic pump and the return duct to the buffer hopper TM12, and also within the glovebox to connect the injection pipe to the injection duct, so as to allow said pipe to be moved and avoid transmitting the vibration of the container 87 to the glovebox. Two cones R1, R2 are put into place on the filler and vent orifices 89 and 90 of the container 87. The volume of each of these cones is not less than half the volume of the injection duct; the volume of said duct between the valve V1 and the outlet orifice of the pipe CAI may be of the order of one or several cubic decimeters (dm3). The injection pipe is supported by a bracket MD enabling the pipe to be moved in translation along axes x and z, and also in rotation about the axis z. The end of the pipe includes a system that provides sealing when the pipe comes to bear against the cone R1. The vent cone R2 is fitted with a radar detector DRA for detecting the presence of mortar in the cone. This cone is connected by a collector CO and a flexible hose CS to a system for extracting air from the cementing glovebox BAG so as to avoid contaminating the inside of the glovebox with air that has passed through the container 87. Mortar injection is stopped as a result of the presence of mortar in the vent cone being detected by the radar sensor; the feed loop is then isolated from the injection duct by closing the valve V1. The feed loop is then emptied by expelling the mortar using compressed air delivered by the source S20. The mortar contained in the feed loop is recovered in the hopper TM12. The flexible connection LS1 of the return duct of the feed loop is then connected to a tank for recovering rinsing water and the delivery from the mortar pump P11 is connected to the industrial waste water network. Water delivered by the source S21 is then introduced into the loop BA together with a sponge ball via an insertion lock S1, which ball is driven by the compressed air so as to clean the loop. The part OP for extracting mortar to the injection duct, and having the valve V1 connected thereto, enables the portion of said valve that is upstream relative to the flow direction of the mortar to be rinsed. The residue of mortar remaining in the injection duct (between the valve V1 and the injection pipe) is emptied into the mortar insertion cone R1 and into the container 87 by thrust from the compressed air delivered by the compressed air source S22 of the device DAE, or by a foam ball, after the pipe has been raised in order to vent the cones. The volume of this “emptied-out” mortar is shared between the filler cone R1 and the vent cone R2. The injection pipe is then moved and positioned by the mechanism MD over a third cone R3 connected to a capacity S14 situated in the glovebox BAG and serving to recover the water used for rinsing the injection portion. Emptying is performed by expelling the water coming from a diaphragm reservoir R13 put under air pressure by the source S22 and filled with water by the source S23. On each occasion after the injection duct has been rinsed, the capacity S14 is emptied to a tank for suspect effluents via a duct provided with a valve V24. With reference to FIGS. 2 to 4 in particular, the extractor member OP is in the form of a Y junction or coupling presenting three duct portions: a first duct portion OP1 and a second duct portion OP2 are connected respectively to the outlet duct CD and to the return duct CR; the third duct portion OP3 is placed (connected) tangentially to the first duct portion and is connected to the injection duct CI. In the embodiment shown in FIG. 2, the third duct portion OP3 extends along an axis OP5 that coincides with the axis of the first portion OP1, the second duct portion OP2 being curved. In the embodiment shown in FIG. 3, the second duct portion OP2 extends along an axis OP4 that coincides with the axis of the first portion OP1, the third duct portion OP3 being curved. In the embodiment of FIG. 4, both the second duct portion OP2 and the third duct portion OP3 are curved. In the embodiments of FIGS. 2 to 4, all three duct portions OP1, OP2, and OP3 of the member OP present substantially identical sections (and/or diameters). In a variant embodiment that is not shown, the section of the first duct portion may be substantially the same as the section of the second duct portion, with the section of the third duct portion being less than the section of the first and second duct portions. The valve V3 serves to isolate the segment in the glovebox during maintenance operations or when changing pipework between the valves V1 and V3. |
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description | This application claims the benefit under 35 U.S.C. § 119(e) of the priority of the following U.S. Provisional Applications filed on Apr. 3, 2013, the entire disclosures of which are hereby incorporated by reference: U.S. Provisional Application No. 61/808,136, entitled “MAGNETIC FIELD PLASMA CONFINEMENT FOR COMPACT FUSION POWER”; U.S. Provisional Application No. 61/808,122, entitled “MAGNETIC FIELD PLASMA CONFINEMENT FOR COMPACT FUSION POWER”; U.S. Provisional Application No. 61/808,131, entitled “ENCAPSULATION AS A METHOD TO ENHANCE MAGNETIC FIELD PLASMA CONFINEMENT”; U.S. Provisional Application No. 61/807,932, entitled “SUPPORTS FOR STRUCTURES IMMERSED IN PLASMA”; U.S. Provisional Application No. 61/808,110, entitled “RESONANT HEATING OF PLASMA WITH HELICON ANTENNAS”; U.S. Provisional Application No. 61/808,066, entitled “PLASMA HEATING WITH RADIO FREQUENCY WAVES”; U.S. Provisional Application No. 61/808,093, entitled “PLASMA HEATING WITH NEUTRAL BEAMS”; U.S. Provisional Application No. 61/808,089, entitled “ACTIVE COOLING OF STRUCTURES IMMERSED IN PLASMA”; U.S. Provisional Application No. 61/808,101, entitled “PLASMA HEATING VIA FIELD OSCILLATIONS”; and U.S. Provisional Application No. 61/808,154, entitled “DIRECT ENERGY CONVERSION OF FUSION PLASMA ENERGY VIA CYCLED ADIABATIC COMPRESSION AND EXPANSION”. This disclosure generally relates to fusion reactors and more specifically to heating plasma for compact fusion power using antennas. Fusion power is power that is generated by a nuclear fusion process in which two or more atomic nuclei collide at very high speed and join to form a new type of atomic nucleus. A fusion reactor is a device that produces fusion power by confining and controlling plasma. Typical fusion reactors are large, complex, and cannot be mounted on a vehicle. According to one embodiment, a fusion reactor includes two internal magnetic coils suspended within an enclosure, a center magnetic coil coaxial with the two internal magnetic coils and located proximate to a midpoint of the enclosure, a plurality of encapsulating magnetic coils coaxial with the internal magnetic coils, and two mirror magnetic coil coaxial with the internal magnetic coils. The fusion reactor further includes one or more electromagnetic wave generators operable to inject a beam of electromagnetic waves into the enclosure. Technical advantages of certain embodiments may include providing a compact fusion reactor that is less complex and less expensive to build than typical fusion reactors. Some embodiments may provide a fusion reactor that is compact enough to be mounted on or in a vehicle such as a truck, aircraft, ship, train, spacecraft, or submarine. Some embodiments may provide a fusion reactor that may be utilized in desalination plants or electrical power plants. Other technical advantages will be readily apparent to one skilled in the art from the following figures, descriptions, and claims. Moreover, while specific advantages have been enumerated above, various embodiments may include all, some, or none of the enumerated advantages. Fusion reactors generate power by confining and controlling plasma that is used in a nuclear fusion process. Typically, fusion reactors are extremely large and complex devices. Because of their prohibitively large sizes, it is not feasible to mount typical fusion reactors on vehicles. As a result, the usefulness of typical fusion reactors is limited. The teachings of the disclosure recognize that it is desirable to provide a compact fusion reactor that is small enough to mount on or in vehicles such as trucks, trains, aircraft, ships, submarines, spacecraft, and the like. For example, it may be desirable to provide truck-mounted compact fusion reactors that may provide a decentralized power system. As another example, it may be desirable to provide a compact fusion reactor for an aircraft that greatly expands the range and operating time of the aircraft. In addition, it may desirable to provide a fusion reactor that may be utilized in power plants and desalination plants. The following describes an encapsulated linear ring cusp fusion reactor for providing these and other desired benefits associated with compact fusion reactors. FIG. 1 illustrates applications of a fusion reactor 110, according to certain embodiments. As one example, one or more embodiments of fusion reactor 110 are utilized by aircraft 101 to supply heat to one or more engines (e.g., turbines) of aircraft 101. A specific example of utilizing one or more fusion reactors 110 in an aircraft is discussed in more detail below in reference to FIG. 2. In another example, one or more embodiments of fusion reactor 110 are utilized by ship 102 to supply electricity and propulsion power. While an aircraft carrier is illustrated for ship 102 in FIG. 1, any type of ship (e.g., a cargo ship, a cruise ship, etc.) may utilize one or more embodiments of fusion reactor 110. As another example, one or more embodiments of fusion reactor 110 may be mounted to a flat-bed truck 103 in order to provide decentralized power or for supplying power to remote areas in need of electricity. As another example, one or more embodiments of fusion reactor 110 may be utilized by an electrical power plant 104 in order to provide electricity to a power grid. While specific applications for fusion reactor 110 are illustrated in FIG. 1, the disclosure is not limited to the illustrated applications. For example, fusion reactor 110 may be utilized in other applications such as trains, desalination plants, spacecraft, submarines, and the like. In general, fusion reactor 110 is a device that generates power by confining and controlling plasma that is used in a nuclear fusion process. Fusion reactor 110 generates a large amount of heat from the nuclear fusion process that may be converted into various forms of power. For example, the heat generated by fusion reactor 110 may be utilized to produce steam for driving a turbine and an electrical generator, thereby producing electricity. As another example, as discussed further below in reference to FIG. 2, the heat generated by fusion reactor 110 may be utilized directly by a turbine of a turbofan or fanjet engine of an aircraft instead of a combustor. Fusion reactor 110 may be scaled to have any desired output for any desired application. For example, one embodiment of fusion reactor 110 may be approximately 10 m×7 m and may have a gross heat output of approximately 100 MW. In other embodiments, fusion reactor 110 may be larger or smaller depending on the application and may have a greater or smaller heat output. For example, fusion reactor 110 may be scaled in size in order to have a gross heat output of over 200 MW. FIG. 2 illustrates an example aircraft system 200 that utilizes one or more fusion reactors 110, according to certain embodiments. Aircraft system 200 includes one or more fusion reactors 110, a fuel processor 210, one or more auxiliary power units (APUs) 220, and one or more turbofans 230. Fusion reactors 110 supply hot coolant 240 to turbofans 230 (e.g., either directly or via fuel processor 210) using one or more heat transfer lines. In some embodiments, hot coolant 240 is FLiBe (i.e., a mixture of lithium fluoride (LiF) and beryllium fluoride (BeF2)) or LiPb. In some embodiments, hot coolant 240 is additionally supplied to APUs 220. Once used by turbofans 240, return coolant 250 is fed back to fusion reactors 110 to be heated and used again. In some embodiments, return coolant 250 is fed directly to fusion reactors 110. In some embodiments, return coolant 250 may additionally be supplied to fusion reactors 110 from APUs 220. In general, aircraft system 200 utilizes one or more fusion reactors 110 in order to provide heat via hot coolant 240 to turbofans 230. Typically, a turbofan utilizes a combustor that burns jet fuel in order to heat intake air, thereby producing thrust. In aircraft system 200, however, the combustors of turbofans 230 have been replaced by heat exchangers that utilize hot coolant 240 provided by one or more fusion reactors 110 in order to heat the intake air. This may provide numerous advantages over typical turbofans. For example, by allowing turbofans 230 to operate without combustors that burn jet fuel, the range of aircraft 101 may be greatly extended. In addition, by greatly reducing or eliminating the need for jet fuel, the operating cost of aircraft 101 may be significantly reduced. FIGS. 3A and 3B illustrate a fusion reactor 110 that may be utilized in the example applications of FIG. 1, according to certain embodiments. In general, fusion reactor 110 is an encapsulated linear ring cusp fusion reactor in which encapsulating magnetic coils 150 are used to prevent plasma that is generated using internal cusp magnetic coils from expanding. In some embodiments, fusion reactor 110 includes an enclosure 120 with a center line 115 running down the center of enclosure 120 as shown. In some embodiments, enclosure 120 includes a vacuum chamber and has a cross-section as discussed below in reference to FIG. 7. Fusion reactor 100 includes internal coils 140 (e.g., internal coils 140a and 140, also known as “cusp” coils), encapsulating coils 150, and mirror coils 160 (e.g., mirror coils 160a and 160b). Internal coils 140 are suspended within enclosure 120 by any appropriate means and are centered on center line 115. Encapsulating coils 150 are also centered on center line 115 and may be either internal or external to enclosure 120. For example, encapsulating coils 150 may be suspended within enclosure 120 in some embodiments. In other embodiments, encapsulating coils 150 may be external to enclosure 120 as illustrated in FIGS. 3A and 3B. In general, fusion reactor 100 provides power by controlling and confining plasma 310 within enclosure 120 for a nuclear fusion process. Internal coils 140, encapsulating coils 150, and mirror coils 160 are energized to form magnetic fields which confine plasma 310 into a shape such as the shape shown in FIGS. 3B and 5. Certain gases, such as deuterium and tritium gases, may then be reacted to make energetic particles which heat plasma 310 and the walls of enclosure 120. The generated heat may then be used, for example, to power vehicles. For example, a liquid metal coolant such as FLiBe or LiPb may carry heat from the walls of fusion reactor 110 out to engines of an aircraft. In some embodiments, combustors in gas turbine engines may be replaced with heat exchangers that utilize the generated heat from fusion reactor 110. In some embodiments, electrical power may also be extracted from fusion reactor 110 via magnetohydrodynamic (MHD) processes. Fusion reactor 110 is an encapsulated linear ring cusp fusion device. The main plasma confinement is accomplished in some embodiments by a central linear ring cusp (e.g., center coil 130) with two spindle cusps located axially on either side (e.g., internal coils 140). These confinement regions are then encapsulated (e.g., with encapsulating coils 150) within a coaxial mirror field provided by mirror coils 160. The magnetic fields of fusion reactor 110 are provided by coaxially located magnetic field coils of varying sizes and currents. The ring cusp losses of the central region are mitigated by recirculation into the spindle cusps. This recirculating flow is made stable and compact by the encapsulating fields provided by encapsulating coils 150. The outward diffusion losses and axial losses from the main confinement zones are mitigated by the strong mirror fields of the encapsulating field provided by encapsulating coils 150. To function as a fusion energy producing device, heat is added to the confined plasma 310, causing it to undergo fusion reactions and produce heat. This heat can then be harvested to produce useful heat, work, and/or electrical power. Fusion reactor 110 is an improvement over existing systems in part because global MHD stability can be preserved and the losses through successive confinement zones are more isolated due to the scattering of particles moving along the null lines. This feature means that particles moving along the center line are not likely to pass immediately out of the system, but will take many scattering events to leave the system. This increases their lifetime in the device, increasing the ability of the reactor to produce useful fusion power. Fusion reactor 110 has novel magnetic field configurations that exhibit global MHD stability, has a minimum of particle losses via open field lines, uses all of the available magnetic field energy, and has a greatly simplified engineering design. The efficient use of magnetic fields means the disclosed embodiments may be an order of magnitude smaller than typical systems, which greatly reduces capital costs for power plants. In addition, the reduced costs allow the concept to be developed faster as each design cycle may be completed much quicker than typical system. In general, the disclosed embodiments have a simpler, more stable design with far less physics risk than existing systems. Enclosure 120 is any appropriate chamber or device for containing a fusion reaction. In some embodiments, enclosure 120 is a vacuum chamber that is generally cylindrical in shape. In other embodiments, enclosure 120 may be a shape other than cylindrical. In some embodiments, enclosure 120 has a centerline 115 running down a center axis of enclosure 120 as illustrated. In some embodiments, enclosure 120 has a first end 320 and a second end 330 that is opposite from first end 320. In some embodiments, enclosure 120 has a midpoint 340 that is substantially equidistant between first end 320 and second end 330. A cross-section of a particular embodiment of enclosure 120 is discussed below in reference to FIG. 8. Some embodiments of fusion reactor 110 may include a center coil 130. Center coil 130 is generally located proximate to midpoint 340 of enclosure 120. In some embodiments, center coil 130 is centered on center line 115 and is coaxial with internal coils 140. Center coil 130 may be either internal or external to enclosure 120, may be located at any appropriate axial position with respect to midpoint 340, may have any appropriate radius, may carry any appropriate current, and may have any appropriate ampturns. Internal coils 140 are any appropriate magnetic coils that are suspended or otherwise positioned within enclosure 120. In some embodiments, internal coils 140 are superconducting magnetic coils. In some embodiments, internal coils 140 are toroidal in shape as shown in FIG. 3B. In some embodiments, internal coils 140 are centered on centerline 115. In some embodiments, internal coils 140 include two coils: a first internal coil 140a that is located between midpoint 340 and first end 320 of enclosure 120, and a second internal coil 140b that is located between midpoint 340 and second end 330 of enclosure 120. Internal coils 140 may be located at any appropriate axial position with respect to midpoint 340, may have any appropriate radius, may carry any appropriate current, and may have any appropriate ampturns. A particular embodiment of an internal coil 140 is discussed in more detail below in reference to FIG. 7. Encapsulating coils 150 are any appropriate magnetic coils and generally have larger diameters than internal coils 140. In some embodiments, encapsulating coils 150 are centered on centerline 115 and are coaxial with internal coils 140. In general, encapsulating coils 150 encapsulate internal coils 140 and operate to close the original magnetic lines of internal coils 140 inside a magnetosphere. Closing these lines may reduce the extent of open field lines and reduce losses via recirculation. Encapsulating coils 150 also preserve the MHD stability of fusion reactor 110 by maintaining a magnetic wall that prevents plasma 310 from expanding. Encapsulating coils 150 have any appropriate cross-section, such as square or round. In some embodiments, encapsulating coils 150 are suspended within enclosure 120. In other embodiments, encapsulating coils 150 may be external to enclosure 120 as illustrated in FIGS. 3A and 3B. Encapsulating coils 150 may be located at any appropriate axial position with respect to midpoint 340, may have any appropriate radius, may carry any appropriate current, and may have any appropriate ampturns. Fusion reactor 110 may include any number and arrangement of encapsulating coils 150. In some embodiments, encapsulating coils 150 include at least one encapsulating coil 150 positioned on each side of midpoint 340 of enclosure 120. For example, fusion reactor 110 may include two encapsulating coils 150: a first encapsulating coil 150 located between midpoint 340 and first end 320 of enclosure 120, and a second encapsulating coil 150 located between midpoint 340 and second end 330 of enclosure 120. In some embodiments, fusion reactor 110 includes a total of two, four, six, eight, or any other even number of encapsulating coils 150. In certain embodiments, fusion reactor 110 includes a first set of two encapsulating coils 150 located between internal coil 140a and first end 320 of enclosure 120, and a second set of two encapsulating coils 150 located between internal coil 140b and second end 330 of enclosure 120. While particular numbers and arrangements of encapsulating coils 150 have been disclosed, any appropriate number and arrangement of encapsulating coils 150 may be utilized by fusion reactor 110. Mirror coils 160 are magnetic coils that are generally located close to the ends of enclosure 120 (i.e., first end 320 and second end 330). In some embodiments, mirror coils 160 are centered on center line 115 and are coaxial with internal coils 140. In general, mirror coils 160 serve to decrease the axial cusp losses and make all the recirculating field lines satisfy an average minimum-β, a condition that is not satisfied by other existing recirculating schemes. In some embodiments, mirror coils 160 include two mirror coils 160: a first mirror coil 160a located proximate to first end 320 of enclosure 120, and a second mirror coil 160b located proximate to second end 330 of enclosure 120. Mirror coils 160 may be either internal or external to enclosure 120, may be located at any appropriate axial position with respect to midpoint 340, may have any appropriate radius, may carry any appropriate current, and may have any appropriate ampturns. In some embodiments, coils 130, 140, 150, and 160 are designed or chosen according to certain constraints. For example, coils 130, 140, 150, and 160 may be designed according to constraints including: high required currents (maximum in some embodiments of approx. 10 MegaAmp-turns); steady-state continuous operation; vacuum design (protected from plasma impingement), toroidal shape, limit outgassing; materials compatible with 150C bakeout; thermal build-up; and cooling between shots. Fusion reactor 110 may include one or more heat injectors 170. Heat injectors 170 are generally operable to allow any appropriate heat to be added to fusion reactor 110 in order to heat plasma 310. In some embodiments, for example, heat injectors 170 may be utilized to add neutral beams in order to heat plasma 310 within fusion reactor 110. In particular embodiments, plasma 310 may be heated using electromagnetic (EM) waves injected through heat injectors as described further below with respect to FIGS. 3C and 3D. Plasma heating can be performed in any suitable fashion according to particular embodiments, and may include neutral beams injections and/or EM wave generation. Although illustrated in a particular position of fusion reactor 110, heat injectors 170 may be coupled to fusion reactor in any suitable location. In operation, fusion reactor 110 generates fusion power by controlling the shape of plasma 310 for a nuclear fusion process using at least internal coils 140, encapsulating coils 150, and mirror coils 160. Internal coils 140 and encapsulating coils 150 are energized to form magnetic fields which confine plasma 310 into a shape such as the shape shown in FIGS. 3B and 5. Gases such as deuterium and tritium may then be reacted to make energetic particles which heat plasma 310 and the walls of enclosure 120. The generated heat may then be used for power. For example, a liquid metal coolant may carry heat from the walls of the reactor out to engines of an aircraft. In some embodiments, electrical power may also be extracted from fusion reactor 110 via MHD. In order to expand the volume of plasma 310 and create a more favorable minimum-β geometry, the number of internal coils can be increased to make a cusp. In some embodiments of fusion reactor 110, the sum of internal coils 140, center coil 130, and mirror coils 160 is an odd number in order to obtain the encapsulation by the outer ‘solenoid’ field (i.e., the magnetic field provided by encapsulating coils 150). This avoids making a ring cusp field and therefor ruining the encapsulating separatrix. Two internal coils 140 and center coil 130 with alternating polarizations give a magnetic well with minimum-β characteristics within the cusp and a quasi-spherical core plasma volume. The addition of two axial ‘mirror’ coils (i.e., mirror coils 160) serves to decrease the axial cusp losses and more importantly makes the recirculating field lines satisfy average minimum-β, a condition not satisfied by other existing recirculating schemes. In some embodiments, additional pairs of internal coils 140 could be added to create more plasma volume in the well. However, such additions may increase the cost and complexity of fusion reactor 110 and may require additional supports for coils internal to plasma 310. In the illustrated embodiments of fusion reactor 110, only internal coils 140 are within plasma 310. In some embodiments, internal coils 140 are suspending within enclosure 120 by one or more supports, such as support 750 illustrated in FIG. 7. While the supports sit outside the central core plasma well, they may still experience high plasma fluxes. Alternatively, internal coils 140 of some embodiments may be amenable to levitation, which would remove the risk and complexity of having support structures within plasma 310. FIGS. 3C and 3D illustrate example configurations that may be used to heat plasma 310 in fusion reactor 110. In order to create the hot plasma condition needed for fusion energy release, energy (e.g., heat) is added to the plasma. In particular embodiments, heating of plasma 310 may be accomplished using electromagnetic (EM) waves, including radio frequency (RF) and/or microwave EM waves. This may be in addition to, or in lieu of, heating plasma 310 using heat injectors 170 (such as with neutral beams, as described above). Because the ability to propagate through low field regions is an important aspect in encapsulated linear ring cusp configurations such as fusion reactor 110 due to their near-zero field regions, RF and microwave EM wave injection may be considered to be a particularly good option for heating plasma 310 within fusion reactor 110. As discussed below, these EM wave generators used in such embodiments may be operable to launch the EM waves on the main device axis (i.e., centerline 115) as well as off-axis into the center of fusion reactor 110. In addition, injection of EM waves with field reversal may allow for energy to be added in resonant zones located on the interior cusp surfaces of fusion reactor 110. Moreover, RF and microwave EM waves with tight beams may be able to propagate through small internal openings between magnetic coils of fusion reactor 110. It will be recognized by those of skill in the art that EM wave generator 381 and antenna 391 of FIGS. 3C and 3D, respectively, are not necessarily drawn to scale. For example, it will be recognized that the appropriate size of EM wave generator 381 and antenna 391 may depend on the wavelength of EM waves used for heating plasma 310 in fusion reactor 110. FIG. 3C illustrates an example EM wave generator 381 coupled to fusion reactor 110. EM wave generator 381 may be any suitable device for generating EM waves in the microwave spectrum for use in heating plasma 310 in fusion reactor 110. EM wave generator 381 may be coupled to fusion reactor 110 through a number of microwave circuit components such as a circulator, tuner, and/or waveguide. In some embodiments, EM wave generator 381 may generate linear rectangularly polarized TE10 EM waves that are converted to right hand circularly polarized TE11 EM waves prior to being injected into fusion reactor 110. In particular embodiments, EM wave generator 381 may be configured on-axis with fusion reactor 110 (i.e. on the same axis as centerline 115, as shown in FIG. 3C). The complex field geometry and internal structures of an encapsulated linear ring cusp field configuration (such as fusion reactor 110 of FIGS. 3A-3D) may pose unique considerations. As a result, in some embodiments, the EM wave beam may be narrowed to tailor it to the field pattern in fusion reactor 110. For example, the beam may be tailored such that it propagates along the desired trajectory into the center of fusion reactor 110. One form of heating plasma 310 using microwave EM waves includes on-axis microwave electron cyclotron resonance heating (ECRH). ECRH may be accomplished, in some embodiments, by propagating the EM waves on the axis of centerline 115 using circularly polarized EM waves. In some embodiments, the circularly polarized waves may heat electrons when the waves are right-hand circularly polarized. However, configurations with right-hand circularly polarized EM waves may have a magnetic field that changes on-axis, so an EM wave injected as a left-hand circularly polarized wave may heat electrons in those reversed field regions of fusion reactor 110 with a reversed magnetic field. To heat the center of fusion reactor 110, the proper EM wave may accordingly be injected in such a way that it reaches the correct condition (i.e., right-hand circularly polarized) in the center of fusion reactor. Using on-axis ECRH may also provide the ability to selectively heat different regions of plasma 310 within fusion reactor 110 by choosing the polarization of the injected waves. Another form of heating plasma 310 using microwave EM waves includes off-axis microwave ECRH. In particular embodiments, it may be desired to heat the center of plasma 310 without depositing heat on or near the outside of fusion reactor 110. This can be accomplished by injecting an extraordinary wave from an off-axis position (i.e., not aligned with centerline 115) so that it is injected between the internal coils and reaches the device center. An example injection location for such off-axis microwave ECRH may include the location of heat injectors 170 in FIGS. 3A-3D. High frequency EM waves may be useful for electron heating in plasma 310. In addition, high frequency EM waves may also be able to propagate in fusion reactor 110 in tight beams that may be required in some embodiments in order to pass through the gaps between the internal coils. Yet another form of heating plasma 310 using microwave EM waves includes off-axis cavity mode microwave ECRH and ion cyclotron resonance heating (ICRH). This form of plasma heating may avoid complications seen in some embodiments that are caused by intricate magnetic field lines. In particular embodiments, off-axis cavity mode microwave ICRH may include injecting an extraordinary mode EM wave from an off-axis location and letting it reflect around the vacuum chamber inside of fusion reactor 110. In such configurations, the EM waves may eventually be absorbed locally in plasma 310 where the resonance conditions are satisfied. Yet another form of heating plasma 310 using microwave EM waves includes helicon-based ECRH and ICRH. In this form of plasma heating, electrons and ions may be heated using injection of right-hand or left-hand EM waves from helical antennas. An example of a helical antenna is illustrated in FIG. 3D. In particular embodiments, the EM waves from the helical antenna may be Whistler or Alfven waves and may be operable to propagate in the low magnetic field regions on the axis of centerline 115 in fusion reactor 110. In some embodiments, the antennas used with helicon-based ECRH and ICRH may be sized to produced resonant conditions with much higher energies than is typically seen for plasma heating and/or processing work. FIG. 3D illustrates an example antenna 391 coupled to fusion reactor 110. Antenna 391 may be any suitable antenna for generating EM waves in the RF spectrum for use in heating plasma 310 in fusion reactor 110. In particular embodiments, antenna 391 may comprise a helical antenna. In some embodiments, antenna 391 may generate circularly polarized EM waves (such as right-hand circularly polarized EM waves) prior to being injected into fusion reactor 110. Plasma heated using antennas may be suited to provide ionization of plasmas as the energy is preferentially pumped into a desired energy level, which is useful for fusion power generation as well as plasma processing. Furthermore, because the energy is pumped somewhat directly into a desired energy level, less energy is wasted on undesired energy levels. Using antennas such as a helical antenna may be well-suited in particular embodiments because propagation of the EM waves may begin in low field regions of fusion reactor 110 and continue through magnetic field reversals and minimum-β (i.e., low magnetic field) regions. The size of antenna 391 may, in particular embodiments, be scaled in order to produce a desired energy population at a given density by maximizing Landau damping of the energetic particles in plasma 310. In particular embodiments using a helicon antenna, antenna 391 may be made to produce circularly polarized transverse electric fields, which may allow both ions and electrons to be resonantly heated. In such embodiments, EM waves may be injected in a Whistler or Alfven mode and may accordingly propagate through low magnetic field regions in fusion reactor (such as those that may exist at the exterior of the fusion reactor 110). As the EM waves propagate toward the center of fusion reactor 110, they may cause resonant heating of plasma 310. These areas may include locations where the magnetic field is such that the gyrofrequency of the plasma particles matches the propagating EM waves. In some embodiments, as the EM waves travel through field reversals in fusion reactor 110, the direction of the particle rotation relative to the handedness may change. Thus, it may be possible to selectively heat ions in the core of fusion reactor 110 since particles of the plasma 310 are only resonant with one of the handedness. When heating ions using antenna 391, the useful frequency of the EM waves may be very low. This may present difficulties in matching antennas with sources. To address these difficulties, magnetic materials such as toroidal ferrites may be added to provide reactance to allow better matching with the antennas 391. Resonant heating using helical antennas 291 may be implemented, in some embodiments, by using glass cylinders mounted to the vacuum vessel with copper-tube and with water-cooled antennas wrapped around the glass. These glass tubes in some embodiments may be positioned on-axis, for example, to match with axial magnetic fields in fusion reactor 110. The glass tubes in other embodiments may be positioned off-axis or on diagonals to direct the EM waves between the internal vacuum structures (e.g., magnetic coils). In particular embodiments, the glass tubes and antennas may be sized such that the internal geometries of fusion reactor 110 are taken into account to achieve the desired resonant energies in fusion reactor 110. FIG. 4 illustrates a simplified view of the coils of fusion reactor 110 and example systems for energizing the coils. In this embodiment, the field geometry is sized to be the minimum size necessary to achieve adequate ion magnetization with fields that can be produced by simple magnet technology. Adequate ion magnetization was considered to be ˜5 ion gyro radii at design average ion energy with respect to the width of the recirculation zone. At the design energy of 100 eV plasma temperature there are 13 ion diffusion jumps and at full 20 KeV plasma energy there are 6.5 ion jumps. This is the lowest to maintain a reasonable magnetic field of 2.2 T in the cusps and keep a modest device size. As illustrated in FIG. 4, certain embodiments of fusion reactor 110 include two mirror coils 160: a first mirror coil 160a located proximate to first end 320 of the enclosure and a second magnetic coil 160b located proximate to second end 330 of enclosure 120. Certain embodiments of fusion reactor 110 also include a center coil 130 that is located proximate to midpoint 340 of enclosure 120. Certain embodiments of fusion reactor 110 also include two internal coils 140: a first internal coil 140a located between center coil 130 and first end 320 of enclosure 120, and a second internal coil 140b located between center coil 130 and second end 330 of enclosure 120. In addition, certain embodiments of fusion reactor 110 may include two or more encapsulating coils 150. For example, fusion reactor 110 may include a first set of two encapsulating coils 150 located between first internal coil 140a and first end 320 of enclosure 120, and a second set of two encapsulating coils 150 located between second internal coil 140b and second end 330 of enclosure 120. In some embodiments, fusion reactor 110 may include any even number of encapsulating coils 150. In some embodiments, encapsulating coils 150 may be located at any appropriate position along center line 115 other than what is illustrated in FIG. 4. In general, encapsulating coils 150, as well as internal coils 140 and mirror coils 160, may be located at any appropriate position along center line 115 in order to maintain magnetic fields in the correct shape to achieve the desired shape of plasma 310. In some embodiments, electrical currents are supplied to coils 130, 140, 150, and 160 as illustrated in FIG. 4. In this figure, each coil has been split along center line 115 and is represented by a rectangle with either an “X” or an “O” at each end. An “X” represents electrical current that is flowing into the plane of the paper, and an “O” represents electrical current that is flowing out the plane of the paper. Using this nomenclature, FIG. 4 illustrates how in this embodiment of fusion reactor 110, electrical currents flow in the same direction through encapsulating coils 150, center coil 130, and mirror coils 160 (i.e., into the plane of the paper at the top of the coils), but flow in the opposite direction through internal coils 140 (i.e., into the plane of the paper at the bottom of the coils). In some embodiments, the field geometry of fusion reactor 110 may be sensitive to the relative currents in the coils, but the problem can be adequately decoupled to allow for control. First, the currents to opposing pairs of coils can be driven in series to guarantee that no asymmetries exist in the axial direction. The field in some embodiments is most sensitive to the center three coils (e.g., internal coils 140 and center coil 130). With the currents of internal coil 140 fixed, the current in center coil 130 can be adjusted to tweak the shape of the central magnetic well. This region can be altered into an axial-oriented ‘bar-bell’ shape by increasing the current on center coil 130 as the increase in flux ‘squeezes’ the sphere into the axial shape. Alternatively, the current on center coil 130 can be reduced, resulting in a ring-shaped magnetic well at midpoint 340. The radius of center coil 130 also sets how close the ring cusp null-line comes to internal coils 140 and may be chosen in order to have this null line close to the middle of the gap between center coil 130 and internal coils 140 to improve confinement. The radius of internal coils 140 serves to set the balance of the relative field strength between the point cusps and the ring cusps for the central well. The baseline sizes may be chosen such that these field values are roughly equal. While it would be favorable to reduce the ring cusp losses by increasing the relative flux in this area, a balanced approach may be more desirable. In some embodiments, the magnetic field is not as sensitive to mirror coils 160 and encapsulating coils 150, but their dimensions should be chosen to achieve the desired shape of plasma 310. In some embodiments, mirror coils 160 may be chosen to be as strong as possible without requiring more complex magnets, and the radius of mirror coils 160 may be chosen to maintain good diagnostic access to the device center. Some embodiments may benefit from shrinking mirror coils 160, thereby achieving higher mirror ratios for less current but at the price of reduced axial diagnostic access. In general, encapsulating coils 150 have weaker magnetic fields than the other coils within fusion reactor 110. Thus, the positioning of encapsulating coils 150 is less critical than the other coils. In some embodiments, the positions of encapsulating coils 150 are defined such that un-interrupted access to the device core is maintained for diagnostics. In some embodiments, an even number of encapsulating coils 150 may be chosen to accommodate supports for internal coils 140. The diameters of encapsulating coils 150 are generally greater than those of internal coils 140, and may be all equal for ease of manufacture and common mounting on or in a cylindrical enclosure 120. In some embodiments, encapsulating coils 150 may be moved inward to the plasma boundary, but this may impact manufacturability and heat transfer characteristics of fusion reactor 110. In some embodiments, fusion reactor 110 includes various systems for energizing center coil 130, internal coils 140, encapsulating coils 150, and mirror coils 160. For example, a center coil system 410, an encapsulating coil system 420, a mirror coil system 430, and an internal coil system 440 may be utilized in some embodiments. Coil systems 410-440 and coils 130-160 may be coupled as illustrated in FIG. 4. Coil systems 410-440 may be any appropriate systems for driving any appropriate amount of electrical currents through coils 130-160. Center coil system 410 may be utilized to drive center coil 130, encapsulating coil system 420 may be utilized to drive encapsulating coils 150, mirror coil system 430 may be utilized to drive mirror coils 160, and internal coil system 440 may be utilized to drive internal coils 140. In other embodiments, more or fewer coil systems may be utilized than those illustrated in FIG. 4. In general, coil systems 410-440 may include any appropriate power sources such as battery banks. FIG. 5 illustrates plasma 310 within enclosure 120 that is shaped and confined by center coil 130, internal coils 140, encapsulating coils 150, and mirror coils 160. As illustrated, an external mirror field is provided by mirror coils 160. The ring cusp flow is contained inside the mirror. A trapped magnetized sheath 510 that is provided by encapsulating coils 150 prevents detachment of plasma 310. Trapped magnetized sheath 510 is a magnetic wall that causes plasma 310 to recirculate and prevents plasma 310 from expanding outward. The recirculating flow is thus forced to stay in a stronger magnetic field. This provides complete stability in a compact and efficient cylindrical geometry. Furthermore, the only losses from plasma exiting fusion reactor 110 are at two small point cusps at the ends of fusion reactor 110 along center line 115. This is an improvement over typical designs in which plasma detaches and exits at other locations. The losses of certain embodiments of fusion reactor 110 are also illustrated in FIG. 5. As mentioned above, the only losses from plasma exiting fusion reactor 110 are at two small point cusps at the ends of fusion reactor 110 along center line 115. Other losses may include diffusion losses due to internal coils 140 and axial cusp losses. In addition, in embodiments in which internal coils 140 are suspended within enclosure 120 with one or more supports (e.g., “stalks”), fusion reactor 110 may include ring cusp losses due to the supports. In some embodiments, internal coils 140 may be designed in such a way as to reduce diffusion losses. For example, certain embodiments of fusion reactor 110 may include internal coils 140 that are configured to conform to the shape of the magnetic field. This may allow plasma 310, which follows the magnetic field lines, to avoid touching internal coils 140, thereby reducing or eliminating losses. An example embodiment of internal coils 140 illustrating a conformal shape is discussed below in reference to FIG. 7. FIG. 6 illustrates a magnetic field of certain embodiments of fusion reactor 110. In general, fusion reactor 110 is designed to have a central magnetic well that is desired for high beta operation and to achieve higher plasma densities. As illustrated in FIG. 6, the magnetic field may include three magnetic wells. The central magnetic well can expand with high Beta, and fusion occurs in all three magnetic wells. Another desired feature is the suppression of ring cusp losses. As illustrated in FIG. 6, the ring cusps connect to each other and recirculate. In addition, good MHD stability is desired in all regions. As illustrated in FIG. 6, only two field penetrations are needed and MHD interchange is satisfied everywhere. In some embodiments, the magnetic fields can be altered without any relocation of the coils by reducing the currents, creating for example weaker cusps and changing the balance between the ring and point cusps. The polarity of the currents could also be reversed to make a mirror-type field and even an encapsulated mirror. In addition, the physical locations of the coils could be altered. FIG. 7 illustrates an example embodiment of an internal coil 140 of fusion reactor 110. In this embodiment, internal coil 140 includes coil windings 710, inner shield 720, layer 730, and outer shield 740. In some embodiments, internal coil 140 may be suspending within enclosure 120 with one or more supports 750. Coil windings 710 may have a width 715 and may be covered in whole or in part by inner shield 720. Inner shield 720 may have a thickness 725 and may be covered in whole or in part by layer 730. Layer 730 may have a thickness 735 and may be covered in whole or in part by outer shield 740. Outer shield may have a thickness 745 and may have a shape that is conformal to the magnetic field within enclosure 120. In some embodiments, internal coil 140 may have an overall diameter of approximately 1.04 m. Coil windings 710 form a superconducting coil and carry an electric current that is typically in an opposite direction from encapsulating coils 150, center coil 130, and mirror coils 160. In some embodiments, width 715 of coils winding is approximately 20 cm. Coil windings 710 may be surrounded by inner shield 720. Inner shield 720 provides structural support, reduces residual neutron flux, and shields against gamma rays due to impurities. Inner shield 720 may be made of Tungsten or any other material that is capable of stopping neutrons and gamma rays. In some embodiments, thickness 725 of inner shield 720 is approximately 11.5 cm. In some embodiments, inner shield 720 is surrounded by layer 730. Layer 730 may be made of lithium (e.g., lithium-6) and may have thickness 735 of approximately 5 mm. Layer 730 may be surrounded by outer shield 740. Outer shield 740 may be made of FLiBe and may have thickness 745 of approximately 30 cm. In some embodiments, outer shield may be conformal to magnetic fields within enclosure 120 in order to reduce losses. For example, outer shield 740 may form a toroid. FIG. 8 illustrates a cut-away view of enclosure 120 of certain embodiments of fusion reactor 110. In some embodiments, enclosure 120 includes one or more inner blanket portions 810, an outer blanket 820, and one or more layers 730 described above. In the illustrated embodiment, enclosure 120 includes three inner blanket portions 810 that are separated by three layers 730. Other embodiments may have any number or configuration of inner blanket portions 810, layers 730, and outer blanket 820. In some embodiments, enclosure 120 may have a total thickness 125 of approximately 80 cm in many locations. In other embodiments, enclosure 120 may have a total thickness 125 of approximately 1.50 m in many locations. However, thickness 125 may vary over the length of enclosure 120 depending on the shape of the magnetic field within enclosure 120 (i.e., the internal shape of enclosure 120 may conform to the magnetic field as illustrated in FIG. 3b and thus many not be a uniform thickness 125). In some embodiments, inner blanket portions 810 have a combined thickness 815 of approximately 70 cm. In other embodiments, inner blanket portions 810 have a combined thickness 815 of approximately 126 cm. In some embodiments, inner blanket portions are made of materials such as Be, FLiBe, and the like. Outer blanket 820 is any low activation material that does not tend to become radioactive under irradiation. For example, outer blanket 820 may be iron or steel. In some embodiments, outer blanket 820 may have a thickness 825 of approximately 10 cm. FIG. 9 illustrates an example computer system 900. In particular embodiments, one or more computer systems 900 are utilized by fusion reactor 110 for any aspects requiring computerized control. Particular embodiments include one or more portions of one or more computer systems 900. Herein, reference to a computer system may encompass a computing device, and vice versa, where appropriate. Moreover, reference to a computer system may encompass one or more computer systems, where appropriate. This disclosure contemplates any suitable number of computer systems 900. This disclosure contemplates computer system 900 taking any suitable physical form. As example and not by way of limitation, computer system 900 may be an embedded computer system, a system-on-chip (SOC), a single-board computer system (SBC) (such as, for example, a computer-on-module (COM) or system-on-module (SOM)), a desktop computer system, a laptop or notebook computer system, an interactive kiosk, a mainframe, a mesh of computer systems, a mobile telephone, a personal digital assistant (PDA), a server, a tablet computer system, or a combination of two or more of these. Where appropriate, computer system 900 may include one or more computer systems 900; be unitary or distributed; span multiple locations; span multiple machines; span multiple data centers; or reside in a cloud, which may include one or more cloud components in one or more networks. Where appropriate, one or more computer systems 900 may perform without substantial spatial or temporal limitation one or more steps of one or more methods described or illustrated herein. As an example and not by way of limitation, one or more computer systems 900 may perform in real time or in batch mode one or more steps of one or more methods described or illustrated herein. One or more computer systems 900 may perform at different times or at different locations one or more steps of one or more methods described or illustrated herein, where appropriate. In particular embodiments, computer system 900 includes a processor 902, memory 904, storage 906, an input/output (I/O) interface 908, a communication interface 910, and a bus 912. Although this disclosure describes and illustrates a particular computer system having a particular number of particular components in a particular arrangement, this disclosure contemplates any suitable computer system having any suitable number of any suitable components in any suitable arrangement. In particular embodiments, processor 902 includes hardware for executing instructions, such as those making up a computer program. As an example and not by way of limitation, to execute instructions, processor 902 may retrieve (or fetch) the instructions from an internal register, an internal cache, memory 904, or storage 906; decode and execute them; and then write one or more results to an internal register, an internal cache, memory 904, or storage 906. In particular embodiments, processor 902 may include one or more internal caches for data, instructions, or addresses. This disclosure contemplates processor 902 including any suitable number of any suitable internal caches, where appropriate. As an example and not by way of limitation, processor 902 may include one or more instruction caches, one or more data caches, and one or more translation lookaside buffers (TLBs). Instructions in the instruction caches may be copies of instructions in memory 904 or storage 906, and the instruction caches may speed up retrieval of those instructions by processor 902. Data in the data caches may be copies of data in memory 904 or storage 906 for instructions executing at processor 902 to operate on; the results of previous instructions executed at processor 902 for access by subsequent instructions executing at processor 902 or for writing to memory 904 or storage 906; or other suitable data. The data caches may speed up read or write operations by processor 902. The TLBs may speed up virtual-address translation for processor 902. In particular embodiments, processor 902 may include one or more internal registers for data, instructions, or addresses. This disclosure contemplates processor 902 including any suitable number of any suitable internal registers, where appropriate. Where appropriate, processor 902 may include one or more arithmetic logic units (ALUs); be a multi-core processor; or include one or more processors 902. Although this disclosure describes and illustrates a particular processor, this disclosure contemplates any suitable processor. In particular embodiments, memory 904 includes main memory for storing instructions for processor 902 to execute or data for processor 902 to operate on. As an example and not by way of limitation, computer system 900 may load instructions from storage 906 or another source (such as, for example, another computer system 900) to memory 904. Processor 902 may then load the instructions from memory 904 to an internal register or internal cache. To execute the instructions, processor 902 may retrieve the instructions from the internal register or internal cache and decode them. During or after execution of the instructions, processor 902 may write one or more results (which may be intermediate or final results) to the internal register or internal cache. Processor 902 may then write one or more of those results to memory 904. In particular embodiments, processor 902 executes only instructions in one or more internal registers or internal caches or in memory 904 (as opposed to storage 906 or elsewhere) and operates only on data in one or more internal registers or internal caches or in memory 904 (as opposed to storage 906 or elsewhere). One or more memory buses (which may each include an address bus and a data bus) may couple processor 902 to memory 904. Bus 912 may include one or more memory buses, as described below. In particular embodiments, one or more memory management units (MMUs) reside between processor 902 and memory 904 and facilitate accesses to memory 904 requested by processor 902. In particular embodiments, memory 904 includes random access memory (RAM). This RAM may be volatile memory, where appropriate. Where appropriate, this RAM may be dynamic RAM (DRAM) or static RAM (SRAM). Moreover, where appropriate, this RAM may be single-ported or multi-ported RAM. This disclosure contemplates any suitable RAM. Memory 904 may include one or more memories 904, where appropriate. Although this disclosure describes and illustrates particular memory, this disclosure contemplates any suitable memory. In particular embodiments, storage 906 includes mass storage for data or instructions. As an example and not by way of limitation, storage 906 may include a hard disk drive (HDD), a floppy disk drive, flash memory, an optical disc, a magneto-optical disc, magnetic tape, or a Universal Serial Bus (USB) drive or a combination of two or more of these. Storage 906 may include removable or non-removable (or fixed) media, where appropriate. Storage 906 may be internal or external to computer system 900, where appropriate. In particular embodiments, storage 906 is non-volatile, solid-state memory. In particular embodiments, storage 906 includes read-only memory (ROM). Where appropriate, this ROM may be mask-programmed ROM, programmable ROM (PROM), erasable PROM (EPROM), electrically erasable PROM (EEPROM), electrically alterable ROM (EAROM), or flash memory or a combination of two or more of these. This disclosure contemplates mass storage 906 taking any suitable physical form. Storage 906 may include one or more storage control units facilitating communication between processor 902 and storage 906, where appropriate. Where appropriate, storage 906 may include one or more storages 906. Although this disclosure describes and illustrates particular storage, this disclosure contemplates any suitable storage. In particular embodiments, I/O interface 908 includes hardware, software, or both, providing one or more interfaces for communication between computer system 900 and one or more I/O devices. Computer system 900 may include one or more of these I/O devices, where appropriate. One or more of these I/O devices may enable communication between a person and computer system 900. As an example and not by way of limitation, an I/O device may include a keyboard, keypad, microphone, monitor, mouse, printer, scanner, speaker, still camera, stylus, tablet, touch screen, trackball, video camera, another suitable I/O device or a combination of two or more of these. An I/O device may include one or more sensors. This disclosure contemplates any suitable I/O devices and any suitable I/O interfaces 908 for them. Where appropriate, I/O interface 908 may include one or more device or software drivers enabling processor 902 to drive one or more of these I/O devices. I/O interface 908 may include one or more I/O interfaces 908, where appropriate. Although this disclosure describes and illustrates a particular I/O interface, this disclosure contemplates any suitable I/O interface. In particular embodiments, communication interface 910 includes hardware, software, or both providing one or more interfaces for communication (such as, for example, packet-based communication) between computer system 900 and one or more other computer systems 900 or one or more networks. As an example and not by way of limitation, communication interface 910 may include a network interface controller (NIC) or network adapter for communicating with an Ethernet or other wire-based network or a wireless NIC (WNIC) or wireless adapter for communicating with a wireless network, such as a WI-FI network. This disclosure contemplates any suitable network and any suitable communication interface 910 for it. As an example and not by way of limitation, computer system 900 may communicate with an ad hoc network, a personal area network (PAN), a local area network (LAN), a wide area network (WAN), a metropolitan area network (MAN), or one or more portions of the Internet or a combination of two or more of these. One or more portions of one or more of these networks may be wired or wireless. As an example, computer system 900 may communicate with a wireless PAN (WPAN) (such as, for example, a BLUETOOTH WPAN), a WI-FI network, a WI-MAX network, a cellular telephone network (such as, for example, a Global System for Mobile Communications (GSM) network), or other suitable wireless network or a combination of two or more of these. Computer system 900 may include any suitable communication interface 910 for any of these networks, where appropriate. Communication interface 910 may include one or more communication interfaces 910, where appropriate. Although this disclosure describes and illustrates a particular communication interface, this disclosure contemplates any suitable communication interface. In particular embodiments, bus 912 includes hardware, software, or both coupling components of computer system 900 to each other. As an example and not by way of limitation, bus 912 may include an Accelerated Graphics Port (AGP) or other graphics bus, an Enhanced Industry Standard Architecture (EISA) bus, a front-side bus (FSB), a HYPERTRANSPORT (HT) interconnect, an Industry Standard Architecture (ISA) bus, an INFINIBAND interconnect, a low-pin-count (LPC) bus, a memory bus, a Micro Channel Architecture (MCA) bus, a Peripheral Component Interconnect (PCI) bus, a PCI-Express (PCIe) bus, a serial advanced technology attachment (SATA) bus, a Video Electronics Standards Association local (VLB) bus, or another suitable bus or a combination of two or more of these. Bus 912 may include one or more buses 912, where appropriate. Although this disclosure describes and illustrates a particular bus, this disclosure contemplates any suitable bus or interconnect. Herein, a computer-readable non-transitory storage medium or media may include one or more semiconductor-based or other integrated circuits (ICs) (such, as for example, field-programmable gate arrays (FPGAs) or application-specific ICs (ASICs)), hard disk drives (HDDs), hybrid hard drives (HHDs), optical discs, optical disc drives (ODDs), magneto-optical discs, magneto-optical drives, floppy diskettes, floppy disk drives (FDDs), magnetic tapes, solid-state drives (SSDs), RAM-drives, SECURE DIGITAL cards or drives, any other suitable computer-readable non-transitory storage media, or any suitable combination of two or more of these, where appropriate. A computer-readable non-transitory storage medium may be volatile, non-volatile, or a combination of volatile and non-volatile, where appropriate. Herein, “or” is inclusive and not exclusive, unless expressly indicated otherwise or indicated otherwise by context. Therefore, herein, “A or B” means “A, B, or both,” unless expressly indicated otherwise or indicated otherwise by context. Moreover, “and” is both joint and several, unless expressly indicated otherwise or indicated otherwise by context. Therefore, herein, “A and B” means “A and B, jointly or severally,” unless expressly indicated otherwise or indicated otherwise by context. The scope of this disclosure encompasses all changes, substitutions, variations, alterations, and modifications to the example embodiments described or illustrated herein that a person having ordinary skill in the art would comprehend. The scope of this disclosure is not limited to the example embodiments described or illustrated herein. Moreover, although this disclosure describes and illustrates respective embodiments herein as including particular components, elements, functions, operations, or steps, any of these embodiments may include any combination or permutation of any of the components, elements, functions, operations, or steps described or illustrated anywhere herein that a person having ordinary skill in the art would comprehend. Furthermore, reference in the appended claims to an apparatus or system or a component of an apparatus or system being adapted to, arranged to, capable of, configured to, enabled to, operable to, or operative to perform a particular function encompasses that apparatus, system, component, whether or not it or that particular function is activated, turned on, or unlocked, as long as that apparatus, system, or component is so adapted, arranged, capable, configured, enabled, operable, or operative. |
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abstract | As typically embodied, the inventive method features bombardment of atomic nuclei with 3He ions in order to effect transmutation of atoms from a first atomic element to a second atomic element. Two notable inventive genres describe transmutation of: oxygen to nitrogen in an oxygen-containing target (e.g., including ZnO film); and, carbon to boron in a carbon-containing target (e.g., including SiC film). According to the former, transmutation of 16O to 15N occurs; more specifically, transmutation of 16O to 15O occurs via nuclear bombardment, and then transmutation of 15O to 15N occurs via decay by positron emission. According to the latter, transmutation of 12C to 11B occurs; more specifically, transmutation of 12C to 11C occurs via nuclear bombardment, and then transmutation of 11C to 11B occurs via decay by positron emission. Inventive practice frequently results in significant alteration of at least one physical property among: electronic carrier concentration; resistivity; photoconductivity; luminescence; morphology. |
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062663895 | description | DETAILED DESCRIPTION OF THE INVENTION The present invention will now be described based on the preferred embodiments. This does not intend to limit the scope of the present invention, but exemplify the invention. All of the features and the combinations thereof described in the embodiments are not necessarily essential to the invention. In the description below, an X-ray refers to an electromagnetic wave having a wave length of about 1-50 nm. The first embodiment of the present invention will be described below with reference to the accompanying drawings. FIG. 1 shows the exposure apparatus EX of the first embodiment of the present invention. The exposure apparatus EX is a step and scan type projection exposure apparatus using an EUV radiation having a wave length about 5-15 nm as an exposure light. In FIG. 1 the Z-axis is parallel to the optical axis of the projection system 9 which projects a reduction image of a mask 8 onto a wafer 10. The Y-axis is perpendicular to the Z-axis in the paper plane, and the X-axis is perpendicular to the paper plane. The exposure apparatus EX scans the reflection type reticle 8 and the wafer 10 against the projection system 9 along the one-dimensional direction (Y direction in this embodiment), as the partial image of the circuit pattern formed on a reflection type reticle 8 as a mask is projected through the projection system 9 onto a wafer 10 as a substrate. It then transcribes the entire circuit pattern formed on the reflection type reticle 8 as a mask onto each of a plurality of shooting areas of a wafer 10 by means of the step and scan method. The soft X-rays which used as an exposure light in this embodiment are very poorly transmitted through the air. The optical path of the EUV radiation is therefore encased in the vacuum chamber 1. The illumination system of this embodiment is described below. The laser light source 100 supplies a laser light having a wave length range from infra-red to visible light. For example, the YAG laser using the excitation by the semiconductor laser, or the excimer laser can be used as the laser light source 100. The laser light is condensed at the position 3 by the condensing optical system 101. The nozzle 2 expels gaseous material to the position 3 where the gaseous material contacts with the highly condensed laser light. The gaseous material then becomes hot due to the energy of the laser light and excited to a plasma. EUV radiation is radiated during the transition from plasma to the lower potential level. Around the position 3 an elliptic mirror 4 constituting a condensing optical system is positioned. The elliptic mirror 4 is placed at the position where its first focal point nearly coincides with the position 3. The inner surface of the elliptic mirror 4 is coated with a multi-layer membrane for reflecting the EUV radiation. The EUV radiation reflected by the inner surface of the elliptic mirror 4 is condensed at the second focal point of the elliptic mirror 4, and goes to the paraboloid mirror 5 constituting another condensing optical system. The focal point of the paraboloid mirror 5 is in nearly coincides with the second focal position of the elliptic mirror 4. The inner surface of the paraboloid mirror 5 is also coated with a multi-layer membrane for reflecting the EUV radiation. The EUV radiation reflected by the paraboloid mirror 5 is almost collimated and goes to the reflection fly-eye optical system 6 as an optical integrator. The reflection fly-eye optical system 6 comprises the first reflection element 6a having a plurality of reflection planes and the second reflection element 6b having a plurality of reflection planes corresponding to the reflection planes of the first reflection element 6a. The surfaces of the plurality of reflection planes of the first and second reflection elements 6a, 6b are also coated by multi-layer membrane for reflecting the EUV radiation. The wavefront of the EUV radiation collimated by the paraboloid mirror 5 is split by the first reflection element 6a, and a plurality of light source images condensed by each reflection plane is formed. The second reflection element 6b is deposited at a position near the light source images. The reflection planes of the second reflection elements 6b act as a field of mirrors. The reflection fly-eye optical system 6 forms a plurality of light source images as the second light source on the basis of the nearly parallel luminous flux from the paraboloid mirror 5. The reflection fly-eye optical system was disclosed in a similar manner in the Japanese Laid-Open Publication 10-47400. The EUV radiation from the second light source formed by the reflection fly-eye optical system 6 goes to the condenser mirror 7 whose focal point is near the position of the second light source. It is then reflected and condensed by the condenser mirror 7, finally arriving at the reflection type reticle 8 through the reflection mirror 7a. The surfaces of the condenser mirror 7 and the reflection mirror 7a are coated with a multi-layer membrane for reflecting the EUV radiation. The condenser mirror 7 uniformly illuminates the predetermined area of the reflection type reticle 8 by condensing and superimposing the EUV radiation from the second light source. In this embodiment, the illumination system is a non-telecentric system and the projection system 9 is also a non-telecentric system on the reticle side. This splits spatially the optical path to the reflection type reticle 8 and the optical path from the reflection type reticle 8 to the projection system 9. The multi-layer reflection membrane having a shape according to the predetermined pattern to be transferred onto the wafer 10 is deposited on the reflection type reticle 8. The EUV radiation is reflected by the reflection type reticle 8 and directed to the projection system 9, including the information of the pattern formed on the reflection type reticle 8. The projection system 9 of this embodiment comprises four reflection mirrors. The first mirror 91 has a concave shape, the second mirror 92 has a convex shape, the third mirror 93 has a convex shape, and the fourth mirror 94 has a concave shape. The surface of each mirror 91-94 is coated with a multi-layer membrane for reflecting the EUV radiation. Each mirror is placed at the position where the optical axes of mirrors are coaxial. The first mirror 91, the second mirror 92, and the fourth mirror 94 have a notch so as not to block the optical path formed by each mirror 91-94. The non-illustrated aperture stop is deposited at the third mirror 93. The EUV radiation reflected by the reflection type reticle 8 is further reflected by each mirror 91-94 in order of precedence. This forms the reduction image of the pattern of the reflection type reticle 8 onto the projection area of the wafer 10 with the predetermined reduction magnification .beta. (for example, .vertline..beta..vertline.=1/4, 1/5, 1/6). The projection system 9 is telecentric on the side of the image (wafer 10). The reflection type reticle 8 is held by the reticle stage which is movable along at least the direction Y and the wafer 10 is held by the wafer stage movable along the X-Y-Z direction, though these stages is not illustrated in FIG. 1. During the exposure process, as the illumination area of the reflection type reticle 8 is illuminated with the EUV radiation from the illumination system, the reflection type reticle 8 and the wafer 10 are shifted against the projection system 9 with the velocity ratio determined by the reduction magnification of the projection system 9. The pattern of the reflection type reticle 8 is exposed to the predetermined shooting area of the wafer 10 by scanning process. Described below with reference to FIG. 2--FIG. 4 is the method for manufacturing the projection system of this embodiment. FIG. 2 is a flowchart showing the steps for manufacturing the projection system of this embodiment. FIG. 3 shows the schematic frame of the interferometer for measuring the image property of the projection system. In the step S101, the mirrors constituting the projection system, the holder and the barrel holding the mirrors (which is not illustrated in FIG. 1) are manufactured according to the design factor of the predetermined data of the optical architecture. The surface of the mirrors are coated with a multi-layer membrane for reflecting the EUV radiation. In the step S102, the surface shapes of the mirrors manufactured in the step S101 are measured by the interferometer. An interferometer using a non-exposure light such as visible light or an interferometer using an exposure light (the EUV radiation) can be used for this measuring step. When the interferometer is using a non-exposure light (for example the well-known Fizeau interferometer and Twyman-Green interferometer) to measure the surface shape of the mirrors, the measurement result may be different from the measurement result obtained by using the EUV radiation. This is because the non-exposure light is reflected by the surface of the multi-layer membrane while the EUV radiation exposure light can be reflected by the interior of the multi-layer membrane. This deviation is added to the measured data as the offset in this case. The interferometer disclosed in the U.S. Pat. No. 5,076,695 by the present applicant can be used for the above mentioned measuring step using the exposure light as the measuring light. The above mentioned measurement of the surface shape of the mirrors using the interferometer is performed on each mirror (91-94) of the projection system 9. The measured data of the surface shape is memorized in the memory member connected to the operation member in much the same way as a calculator through an input system such as a console. In step S102, each mirror 91-94 held by a holder is built into the barrel of the projection system 9. In this embodiment, the holder comprises an apparatus for adjusting the spatial position of each mirror (91-94) and a position measuring apparatus such as a micrometer for measuring the position of each mirror based on the predetermined reference position. In the step S103, the information of the position of the reflection plane in connection with the position of each mirror 91-94 is memorized in the above mentioned memory member in parallel with the step for composing the projection system 9. In step S104, as shown in FIG. 3, the imaging performance (wave aberration) of the projection system 9 composed in the step S103 is measured. The measurement is performed by attaching the projection system 9 to the wave aberration measuring apparatus shown in FIG. 3. The wave aberration measuring apparatus shown in FIG. 3 can be attached to the body of the exposure apparatus. In FIG. 3, the laser light source 20 illuminates the laser light having a predetermined wave length range as nearly plane waves. This laser light is directed to the condensing optical system 22 after going through the beam splitter 21. The condensing optical system 22 has negligible aberration because the spherical aberration and the offense against the sine condition is lowered to be negligible compared with the measurement error of the wave aberration measuring apparatus. The incidence plane 22a of the condensing optical system 22 has the same shape as the incident plane wave (that is the plane shape). The plane wave entering the incidence plane 22a is subject to amplitude splitting because the incidence plane 22a is a half-mirror. The reflected light goes back to the beam splitter 21 with the plane wave remaining as the reference light. The light transmitted by the incidence plane 22a is condensed by the condensing optical system 22, converted to a spherical wave, and goes to the projection system 9. It is correct that the final plane of the condensing optical system 22 can be used as the reference plane. The position of the projection system 9 is such that the position of the mask plane of the projection system 9 is coincides to the position of the focal point of the spherical wave in the wave aberration measuring apparatus. The spherical wave emanated from the mask plane enters the projection system 9. If the wave aberration does not exist in the projection system 9, that is, if the projection system 9 is an ideal optical system, a spherical wave is emanated from the projection system 9 and condensed at the position of the image plane. The spherical mirror 23 is deposited on the emanating side of the projection system 9. The surface shape of the spherical mirror 23 is equal to the spherical wave emitted by the projection system 9 when the projection system 9 is an ideal optical system. Therefore, if the projection system 9 is an ideal optical system, the spherical wave having the same shape as the spherical wave emitted by the projection system 9 returns again to the projection system 9. If the projection system 9 has a wave aberration, a wave having the shape according to the wave aberration returns to the projection system 9. The measuring light reflected by the spherical mirror 23 and returned to the projection system 9 emanates from the projection system 9 and goes to the beam splitter 32 through the condensing optical system 22. The reference light reflected by the incidence plane of the condensing optical system 9 also goes to the beam splitter 21 as described above. The measuring light and the reference light are reflected by the beam splitter 21 and directed to the light receiving plane made of a light receiving material 24. It consists of a photoelectric conversion element such as CCD. If the projection system 9 has a wave aberration, the interference fringes are measured on the light receiving plane in accordance with the wave aberration. The shapes of these interference fringes relate to the difference between the reference wave plane and the wave plane of the luminous flux shuttled to and from the projection system 9. The wave aberration of the projection system 9 is therefore calculated by image analysis of these interference fringes. The wave aberrations at a plurality of the positions in the field of the projection system 9 (or the exposure area) can be measured by moving the projection system 9, the condensing optical system 21 and the spherical mirror 23 relative to each other. The laser light source 20 is not limited to just a laser. A light source emitting a plane wave having much longer coherent length can be used as the laser light source 20. In the above mentioned example, the measurement is performed using a light having a different wave length from the exposure light of the projection system 9. This is because the projection system 9 is a reflection type optical system consisting of only mirrors, and the chromatic aberration can be neglected. In this case, a difference between the measured wave aberration and the wave aberration under the EUV radiation may arise because the measuring light is reflected by the surface of the multi-layer membrane on each mirror 91-94 in the projection system 9 while the EUV radiation as the exposure light may be reflected by the interior of the multi-layer membrane. This deviation between the measured wave aberration and the wave aberration under the exposure light is added to the measured wave aberration as the offset. The measurement can be performed by using exposure light instead of non-exposure light. In the case of using exposure light, the principle of the interferometer disclosed by the present applicant in the Japanese Laid-Open Publication 57-64139 and U.S. Pat. No. 5,076,695 can be used. When using the interferometer disclosed in the Japanese Laid-Open Publication 57-64139, a reflection optical element is used instead of the refraction optical element, and the element for converting the EUV radiation to the detective light such as UV-visible-IR light (such as the fluorescence board) is deposited at the position of the detector. For example, the method for detecting the photo current of the metal blocks arranged in the matrix form can be used, and an element sensitive to the EUV radiation can be used as the detector. In step S105, the wave aberration W(.rho., .theta.) expanded by the orthogonal functions such as Zernike's cylindrical functions is calculated from the wave aberration obtained in above mentioned step S104. In the following equation r denotes the normalized pupil radius and q denotes the radius vector angle at the pupil (positive for anti clockwise direction). The wave aberration W(.rho., .theta.) is represented by the formula below. W(.rho., .theta.)=Z0 +Z1.rho.cos.theta. PA2 +Z2.rho.sin.theta. PA2 +Z3(2.rho.2-1) PA2 +Z4(.rho.2cos2.theta.) PA2 +Z5(.rho.2sin2.theta.): wherein Z0, Z1, Z2, Z3, Z4, Z5 . . . are the coefficients of each term. Each term of the above mentioned formula of the wave aberration represents an aberration such as distortion, focus, third order astigmatism, third order coma, third order spherical aberration, coma having three times rotation symmetry, fifth order astigmatism, fifth order coma, fifth order spherical aberration, and similar. Thus, the wave aberration is derived into a plurality of aberration components. In step S106, each aberration component requiring correction is linked to the reflection plane constituting the projection system 9. Then, the reflection plane to be reprocessed is selected. Prior to step S106, the amount of change is calculated for every aberration component by simulation. This is necessary when the design factor according to the optical design data of the projection system 9 (the shape and the position of each reflection plane) requires microscopic changes. The information about the calculated amount of change is memorized in the memory member connected to the operation member in much the same way as a calculator through an input system such as a console. The residual component of aberration is calculated by subtracting the aberration component which can be corrected by changing the position of the reflection plane (the direction of the optical axis, the direction perpendicular to the optical axis, the direction of rotation about the axis perpendicular to the optical axis, the direction of rotation about the optical axis) from a plurality of aberration components which should be corrected. This residual component of the aberration is the aberration component which can only be corrected by changing the shape of plurality of reflection planes. The details of the residual component of aberration are also memorized in the memory member connected to the operation member in much the same way as a calculator through an input system such as a console. When the system is non-telecentric on the side of the object plane, the magnification error can be corrected by changing the distance from the object plane to the projection system along the optical axis. The residual component of the aberration is therefore calculated by subtracting the magnification error. Next, the reflection plane which requires a change in its shape to effectively correct the residual component of aberration is selected on the basis of the degree of the change of the aberration component on each reflection plane which is calculated by simulation, and the residual component of aberration. The information about the relationship between the residual component of aberration and the reflection plane is also memorized in the memory member connected to the operation member in much the same way as a calculator through an input system such as a console. It is preferable that the aberration component relative to the displacement of the image is corrected by the reflection plane near the object plane or the image plane. The aberration component relative to the displacement of the image can be corrected by the reflection plane near the image plane with little influence compared to the aberration component relative to the displacement of the image plane caused by a numerical aperture. There is advantage therefore that the calculation in S106 and S107 is very simple. It is preferable that the projection system 9 is deposited near the object plane when the projection system 9 has reduced magnification. This is because the width of the luminous flux is narrower on the side of the mask plane compared to the side of the image plane and the aberration given to images adjoining each other is easy to control independently. For the above mentioned reason in the embodiment shown in FIG. 1 the mirror 91 placed near the object (reflection type reticle 8) plane is selected as the reprocessing (changing) reflection plane. This allow to make the correction of image displacement aberration component independently of the other aberration components. The mirror placed near the object plane (image plane) refers to the mirror placed at the position where the fluctuation of the aberration component except the aberration component relative to the displacement of the image from the ideal image on the image plane (such as isotropic or anisotropic magnification error, isotropic or anisotropic distortion) is less than half of the fluctuation of the aberration component relative to the displacement of the image from the ideal image on the image plane when the shape of the mirror is modulated by reprocessing or changing the mirror. For the above mentioned reason, in the embodiment shown in FIG. 1 the mirror 93 placed near the position of the aperture stop is selected as the reprocessing (changing) reflection plane so as to correct mainly the aberration component relative to the displacement of the image caused by the numerical aperture. The mirror placed near the position of the aperture stop refers to the mirror placed at the position where the fluctuation of the aberration component except the aberration component relative to the displacement of the image caused by the numerical aperture (the aberration component dependent on the pupil coordinate of the luminous flux passing through the pupil such as spherical aberration, coma, astigmatism) is less than half of the fluctuation of the aberration component relative to the displacement of the image caused by the numerical aperture when the shape of the mirror is modulated by reprocessing, or changing the mirror. When selecting a reflection plane near the mask plane or the image plane and/or a reflection plane near the aperture stop as the reflection plane to change the shape, the reflection plane can be selected on the basis of calculated residual component and the amount of the change of the reflection plane mentioned above. In the step S107, the shape of each selected reflection plane is calculated so as to correct the residual aberration. This step S107 comprises a plurality of sub-steps SS1071.about.SS1078 described below. In description of sub-steps below, m denotes the total number of the reflection planes selected in the step S106, and n denotes the reflection plane which shape is calculated. In the sub-step SS1071 n=1 which denotes the number of the reflection plane. In the sub-step SS1072, the shape data of each reflection plane constituting the projection system 9 measured in the step S102 and the information of the position of each reflection plane memorized in the step S103 are read from the memory. In the sub-step SS1073, the residual component of the aberration is read from the memory. In the sub-step SS1074, the shape of the reflection plane n is optimized so as to correct the aberration component considering the shape data of the reflection plane and the position information of the reflection plane read in the sub-step SS1072 as the initial data. In this optimizing step, the parameter to be changed is the shape of the reflection plane n, and the evaluation value is the adjustable aberration component. The aberration is calculated by the optical path tracking simulation using the corrected shape and the position of the reflection plane. The residual component of the aberration as the evaluation value is calculated by subtracting the aberration component which can be reduced by adjusting the position of each reflection plane from the aberration calculated by the simulation. This step is performed until the evaluation value becomes less than set point. A local optimization method such as the DLS (Damped Least Square Method) and the global optimization method such as the GA (Generic Algorithm) can be used in the sub-step SS1704. As the shape of the reflection plane is the parameter used in the sub-step SS1704, not only the rotation symmetric parameters such as the radius of curvature, the aspherical coefficient, but also the gradient of the plane in a plurality of coordinates on the reflection plane can be used. In this embodiment both of these methods are used. In the sub-step SS1075, the shape of the reflection plane calculated in the sub-step SS1074 is memorized in the memory. The shape data of the reflection plane read in the sub-step SS1072 in the next loop will be the shape data memorized in the sub-step SS1075, not the measured shape data. In the sub-step SS1076, the residual component of the aberration calculated in the sub-step SS1074 is memorized in the memory. The residual component of the aberration used in the sub-step SS1073 in the next loop will be the data memorized in this sub-step. In the sub-step SS1077, the decision is made whether the n which is the number of the reflection plane whose shape is calculated overreaches the m which is the total number of the selected reflection planes. If n is less than m, then sub-step SS1078 is performed so as to calculate the shape of the reflection plane having the number of n+1. Following this, the sub-step SS1072 is performed again. Alternatively, the sub-step SS1079 is performed. In the sub-step SS1079, the decision is made whether the residual component of the aberration memorized in the memory is less than the set point. If the residual component of the aberration is less than the set point, the sub-steps SS1071-SS1079 are completed and the step S108 is performed. If not, the sub-step SS1071 is performed again so as to optimize the shape of the selected reflection plane. The sub-step SS1079 is effective in case where the change of the shape of one of the plurality of the reflection planes physically influences the aberration component to be corrected by changing the shape of another reflection plane when a plurality of reflection planes are selected in step S106. As described in step S106, there is an advantage that the divergence of the solution is suppressed if the reflection plane can correct the predetermined aberration component among the image properties in the projection system 9 without physically influencing other aberration components. In step S107, the shape to physically correct the residual component of aberration can be calculated by proportional calculations from the data about the value of the change calculated in step S106 instead of the above mentioned optimization method. In this case it is preferable that the change of the surface shape of the selected reflection plane does not influence the aberration component to be corrected by the change of the surface shape of another reflection plane. In the step S108, the mirrors are outside of the projection system 9, and processed so that the surface shape of the mirror becomes the shape calculated in the step S107. In this embodiment, the surface shape of the mirror is nearly equal to the surface shape of the multi-layer membrane placed on the surface of the mirror. In the step S108, the distribution of the thickness of the membrane on the mirror is changed either by a method of removing partially the multi-layer membrane disposed on the mirror, the method for laminating partially the multi-layer membrane on the mirror, or the method combining above mentioned methods. The method for removing partially the multi-layer membrane is disclosed in the Japanese Laid-Open Publication 7-84098 by the present applicant, and the method for laminating partially the multi-layer membrane is disclosed in the Japanese Laid-Open Publication 10-30170 by the present applicant. In step S108, the reprocessing to change the surface shape can be performed on another mirror which has the same design factor as the mirror to be taken out of the projection system for reprocessing. In this case, the real shape of the surface of the provided mirror is measured in advance. In step S109, the mirrors processed in step S108 are incorporated into the projection system 9. In step S110, the position of the mirror is adjusted based on the information of the position of the reflection plane memorized in the memory. In step S111, the wave aberration of the projection system 9 is measured as in step S104. In step S112, the decision is made whether the wave aberration measured in the step S111 is up to the standard. If the wave aberration is up to the standard, the adjustment of the projection system 9 is completed, or alternatively the next step S113 is performed. In the step S113, the wave aberration measured in the step S112 is derived into a plurality of aberration components as the step S105. The residual component of aberration is calculated by subtracting the aberration component that can be corrected by changing the position of each reflection plane from these aberration components. The decision is then made whether the reprocessing of the reflection plane is needed by comparing the residual component of the aberration with the threshold value. If the residual component of the aberration is less than the threshold value, the next step S110 is performed because there is no need to reprocess the reflection plane. Alternatively the next step S105 is performed to reprocess the reflection plane. The reflection type projection system having excellent image properties is manufactured by above mentioned steps S101-S113. Additionally, the exposure apparatus which can transcribe accurately microscopic patterns by embodying this projection system, the illuminating system, the alignment system, the stage, and similar components in the body of the exposure apparatus are also manufactured by the above mentioned steps S101.about.S113. Next, the second embodiment of the present invention will be described with reference to FIG. 5. The second embodiment includes the modified projection system 9. All the other components are similar to the components of the first embodiment. The drawings and the description of other components will therefore be omitted. In FIG. 5, the projection system 9 comprises two mirrors 95, 96 having a plane shape, a mirror 91 having a concave shape, a mirror 92 having a convex shape, a mirror 93 having a convex shape, and a mirror 94 having a concave shape. Each mirror is deposited at the position where their optical axes (normal axis of the reflection plane in the case of the plane mirror) are coaxial. Each mirror 91-96 consists of the substrate coated by the multi-layer membrane for reflecting the EUV radiation. The mirrors 91, 92, 94 and 95 have a notch so as not to block the optical path formed by each mirror 91-96. Each mirror 91-96 is deposited at the position where the EUV radiation reflected by the reflection type reticle 8 is directed to the image plane by the mirror 95, the mirror 96, the mirror 91, the mirror 92, the mirror 93, and the mirror 94 in this order. The non-illustrated aperture stop is deposited at the third mirror 93, and the telecentric system on the image plane side (wafer side) is produced by this aperture stop. Within the projection system 9 shown in FIG. 5 is also the non-telecentric system on the side of the reflection type reticle 8. This makes it easy to isolate the optical path to the reflection type reticle 8 from the optical-path from the reflection type reticle 8 to the projection system 9. FIG. 6 illustrates the method for determining the surface shape of the mirror 95. In FIG. 6, it is supposed that the main beam perpendicular to the image plane in the image side (the side of the wafer 10) enters the projection system 9 from the image side. The main beams PR1, PR2 pass through the mirrors 91-94 and 96 and are reflected at the control points C1 and C2 on the reflection plane of the mirror 95. When the surface shape of the projection mirror 95 is plane in the initial state, the main beams reach the points P11, P12 on the reflection type reticle 8. If the projection system 9 has no aberration relative to the displacement of the image, this main beam should reach the ideal image points P12 and P22. The reflection plane is inclined near the control points C1 and C2 so that the main beams reach the ideal image points P12 and P22. The gradients of the inclined reflection planes 95a1 and 95a2 are calculated on the basis of the difference between the ideal image points P12 and P22 and the measured image points P11 and P21, the distance between the reflection type reticle 8 and the mirror 95, and the telecentricity at each ideal image point P12 and P22 (the gradient between the designed main beam reaching the ideal point and the optical axis). The surface shape of the mirror 95 is calculated by binding the partial reflection plane 95a1 and 95a2 after calculating the gradient at each control point. The above mentioned method may be interchanged with the steps S106.about.S108 in the first embodiment. In the case of selecting (interchanging) the plane mirror in the projection system 9, the aberration relative to the displacement of the image can be changed with practically no influence on other aberrations. In the projection system 9 shown in FIG. 5, if the mirror 95 and/or the mirror 96 are totally given the curvature, the Petzval's sum of the projection system 9 can be adjusted, as is effective for adjusting the curvature of the field. If the plane mirror 95 and/or the plane mirror 96 are changed to have the toric shape or the cylindrical shape, the magnification error of the image can be adjusted along each X or Y direction independently. If the plane mirror 95 and/or the plane mirror 96 is given the local power distribution, the gradient or the curvature of the image plane can be adjusted because the distribution of the focal point on the image plane along the optical axis can be changed to the desired value. If the surface shape of the reflection plane is given so that the local power along the sagittal direction is different from the local power along the meridian direction, the astigmatism at each point on the image plane can be adjusted. In case the mirror 96 in addition to the mirror 95 is selected (changed), the telecentricity at each point on the image plane as well as the displacement of the image can be controlled, though only the aberration relative to the displacement of the image is adjusted in the example described above. In this case, the telecentricity is controlled so that the telecentricity is equal to the designed telecentricity. When adjusting the magnification error by changing the distance between the reflection type reticle 8 and the projection system 9 along the optical axis, the telecentricity can be controlled so that the gradient of the main beam projecting to the YZ plane is equal to the gradient of the main beam projecting to the XZ plane. In this case, when changing the distance between the reflection type reticle 8 and the projection system 9 along the optical axis the magnification ratio along the scanning direction can be made equal to the magnification ratio along the direction perpendicular to the scanning direction. In the above example, the method for manufacturing an optimized projection system under certain exposure condition is described. There can be a case, however, in which the system optimized under certain conditions is not optimized under other conditions. This occur when the illuminating condition, the type of resist, the pattern rule, the environment of the system, and the similar conditions may vary substantially. In such case, both the mirror optimized under certain condition and the mirror optimized under other conditions can be made changeable. FIG. 7 shows the exposure apparatus comprising a plurality of mirrors changeable according to the exposure condition. The difference between the exposure apparatus shown in FIG. 7 and the exposure apparatus shown in FIG. 1 is described below. In FIG. 7, the .sigma. lens stop AS1 for varying the shape of the optical path is deposited near the second reflection element 6b of the reflection type fly-eye optical system 6. This forms the second light source consisting of a plurality of light source images. The shape of the optical path can be selected from the circular shape having the first diameter (for large .sigma. illumination), the circular shape having a shorter diameter than the first diameter (for small .sigma. illumination), the zone shape (for zone illumination), and four circular or fan shapes placed at the eccentric position from the optical axis (for quadropole illumination). This .sigma. lens stop AS1 is controlled by the .sigma. lens stop controlling unit CU1 so that the shape of the optical path can be changed. The variable aperture stop AS2 which can change the diameter of the optical path is deposited near the mirror 93 in the projection system 9. The diameter of the optical path of the variable aperture stop AS2 is controlled by the variable aperture stop controlling unit CU2. In FIG. 7, the mirror 91 nearest the reflection type reticle 8 is interchangeable with the mirror 97 although their surface shapes are different. This interchange is controlled by the mirror interchange controlling unit CU3. The main controlling unit MCU is connected with the .sigma. lens stop controlling unit CU1, the variable aperture stop controlling unit CU2, the mirror interchange controlling unit CU3, and the input unit IU which inputs information about the exposure condition. An input apparatus such as a console and a bar code reader detecting the bar code on the reflection type reticle 8 or similar can be used as the input unit IU. After receiving information about the exposure condition from the input unit, the main controlling unit MCU transmits the controlling signal about the exposure condition such as the .sigma. value to the .sigma. lens stop controlling unit CU1 and the variable aperture stop controlling unit CU2. The table containing the relation information between the exposure condition and the type of the mirror is memorized in the memory in the main controlling unit MCU. The main controlling unit MCU transmits the controlling signal regarding the mirror to be used to the mirror interchange controlling unit CU3. The .sigma. lens stop controlling unit CU1 drives the .sigma. lens stop AS1 so that the shape of the second light source is equal to the predetermined shape. The variable aperture stop controlling unit CU2 drives the variable aperture stop AS2 so that the numerical aperture of the projection system is equal to the predetermined value. An optimized image property under any condition is achieved by above mentioned method. In FIG. 7, the distribution of the optical property of the EUV radiation (such as the amplitude distribution, the phase distribution) can be changed at the position near the variable aperture stop AS2 according to the type of the pattern to be transcribed (such as line and space pattern, contact hole pattern). In this case, the mirrors 91 and 97 can be changed so as to adjust the aberration fluctuation by changing the distribution of the optical property of the EUV radiation. The reflectance distribution or the shape distribution of the mirror 93 can be changed so as to change the distribution of the optical property of the EUV radiation near the pupil. In this case, the mirror 93 can be also changed to a mirror having another reflectance distribution or another shape distribution. The nonlinear stretch of wafer caused by the wafer production process and associated magnification error and distortion etc. which is asymmetric along the rotation axis caused by the difference of the exposure apparatus could be adjusted if the mirror 96 has a toric shape or a cylindrical shape and the mirror 96 is able to rotate around the optical axis. In the first embodiment, the step S112 can be performed between the step S104 and the step S105. In this case, if the image property is up to grade in step S112, then the process is finished. If not, step S105 is performed. If the residual component of the aberration is negligible in step S106, the step S110 is performed in order to adjust the position of the reflection plane. In the first embodiment, when selecting the reflection plane to change the shape in step S106, only one predetermined reflection plane can be selected. Following this, steps S107-S112 are performed. If the residual component of the aberration relative to the selected reflection plane is over the predetermined threshold in step S113, then the same steps as in the first embodiment are performed. If the residual component of the aberration relative to the other reflection plane is excessive, then another reflection plane is selected and step S107 is performed. If the aberration component relative to the displacement of the image is dominant in step S105, only the mirror 91 nearest to the mask in the projection system shown in FIG. 1 can be selected. If the aberration component relative to the displacement of the image caused by a numerical aperture is dominant, only the mirror 93 near the aperture stop can be selected. The embodiment for forming a predetermined circuit pattern onto a wafer by the above mentioned exposure apparatus is described below with reference to flowchart shown in FIG. 8. Initially, in step 101 in FIG. 8, a metal membrane is evaporated onto one lot of the wafer. In the next step 102, the photo resist is painted onto the metal membrane on one lot of the wafer. In the step 103, the pattern image of the reticle R is exposed stepwise onto each shooting area of one lot of the wafer through the projection system C of the exposure apparatus shown in FIGS. 1, 5, or 7. In the step 104, the photo resist on one lot of the wafer is developed. In step 105, the circuit pattern corresponding to the pattern on the reticle R is formed on each shooting area of each wafer by etching. A device having the supersubtle circuit such as semiconductor device is manufactured by forming the circuit pattern on the upper layer after the above mentioned processes. Although the present invention has been described by way of exemplary embodiments, it should be understood that many changes and substitutions may be made by those skilled in the art without departing from the spirit and the scope of the present invention which is defined only by the appended claims. An exposure apparatus having good image properties can be manufactured by the method of the present invention in cases where the projection system comprises a small number of optical members. The exposure apparatus manufactured by the method of the present invention can form a supersubtle pattern on a substrate. A device having high resolution circuit pattern on a substrate can be manufactured by the method of the present invention. |
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abstract | A small and compact infrared scene projector allows for projection of infrared images from a variety of sources. The unit utilizes an external power supply to reduce the weight of the unit and has several input pathways such as an integrated RS-232 serial port and a video port for receiving RS-170 formatted video signals. The projector uses a digital signal processor to control its internal electronics and for on-board generation of pre-programmed infrared images. A processor electronics card and a scene generator electronics card communicate internally to generate images and to control a thermoelectric cooling device mounted to a semiconductor infrared emitter array to project flicker free, high resolution infrared images. Optics in the form of an interchangeable lens or a mirrored collimator allow for projection of a generated image onto a test object. Internal high speed memory and electrically erasable firmware, both externally programmable, allow for on-the-fly programming and self-sustaining and continued operation after disconnection from a separate programming computer. An external, user programmable interface allows for the download of commands and images into the scene projector. Once programmed by an external computer, such as a personal notebook computer, the scene projector can be placed in the field for live, self test of military electronics that rely on infrared sensors to make decisions. |
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056687281 | abstract | A deflector is disposed in a steam vent region above the upper ends of part-length fuel rods to deflect upwardly flowing liquid laterally outwardly into the interstices of the full-length fuel rods adjacent the steam vent region. The deflector is supported by a spacer or from the upper tie plate and can be removed from the fuel bundle for access to the part-length fuel rods. The deflector may comprise a flat plate, an inverted pyramid, an inverted cone, a multi-sided pyramidal configuration or a swirl device wherein the horizontally projected area of the deflector is substantially coextensive with the horizontal cross-sectional area of the steam vent area. In this manner, a higher density liquid is provided in the interstices of the full-length rods while the lower density steam flows into the steam vent volume for flow upwardly out of the fuel bundle. |
abstract | One of methods for carrying out a reactor vessel according to the present invention comprises the steps of removing an overhead traveling crane in a reactor containment vessel of a pressurized water reactor, or the steps of, in an area where an overhead traveling crane is installed, operating the overhead traveling crane to move aside for creating a space, through which the reactor vessel is able to pass, and then carrying out the reactor vessel through an opening provided in a top portion of the reactor containment vessel. With the present method, the reactor vessel of the pressurized water reactor can be carried out in a short time with high efficiency. |
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description | The embodiments of the present disclosure generally relate to storage of hazardous radioactive materials and, more particularly, to dry storage, spent nuclear fuel casks for containing spent nuclear fuel or other hazardous radioactive material(s). Spent nuclear fuel has historically been stored in deep reservoirs of water, called “spent fuel pools,” within nuclear power plants. This spent fuel storage technology is often termed “wet storage.” Spent fuel pools at reactors are reaching their spent fuel capacity limits, causing concerns about the need to shut down reactors because there is no more room for the spent fuel. Dry nuclear spent fuel storage technology (termed “dry storage”) is deployed throughout the world to expand the capabilities of nuclear power plants to discharge and store nuclear spent fuel external to a reactor's spent fuel pool, thereby extending the operating lives of the power plants. There are two fundamental classes of technology used in dry spent fuel storage: (a) metal casks with final closure lids that are bolted closed at the power plants after loading with spent fuel, and (b) concrete storage casks containing metal canisters having canister final closure lids that are welded closed or sealed with mechanical methods at the power plants following spent fuel loading. This latter dry storage technology is referred to as “canister-based concrete spent fuel storage.” The concrete cask serves as an enclosure, or “overpack” that provides mechanical protection, heat removal features, and radiation shielding for the inner metal canister that encloses the radioactive materials. The use of this technology tends to have significant capital cost and other economic advantages over the use of metal cask technology for storage. Embodiments of a thermal divider insert and method for a dry storage, spent nuclear fuel cask are disclosed. The thermal divider insert enables safe storage of the hazardous nuclear material when one or more air inlets have been fully or partially blocked to an extent that insufficient air flows into the air inlets and through the cask for adequate cooling of the hazardous nuclear material. In one embodiment, among others, a cask comprises a metal canister having a top, bottom, and sidewall. The canister contains the hazardous nuclear material. A concrete overpack contains the metal canister with the hazardous nuclear material. The overpack has a top, bottom, and sidewall. The overpack has an inside surface that is spaced from an outer surface of the canister to create an annular region that permits flow of air between the surfaces for cooling the canister. One or more air inlets near the bottom of the overpack communicates air from an outside environment into the annular region. One or more outlet vents near the top of the overpack communicates air from the annular region to the outside environment. The thermal divider insert extends through a respective outlet vent and into the annular region and is designed to establish two separate and opposite air flows (i.e., inward air flow and outward air flow) through the respective vent and the annular region when the overpack air inlets have been blocked. When not blocked in normal operation, the two air flows both flow upwardly through the annular region and outwardly from the vent. An embodiment of the thermal divider insert, among others, comprises (a) a planar horizontal radial plate and (b) a curved vertical plate extending from the radial plate, in order to establish the two separate and opposite air flows through the vent. The horizontal radial plate extends through the overpack outlet vent. The radial plate has a curvature along its inside and outside edges that corresponds to a curvature associated with the overpack outlet vent. The redial plate establishes a lower air flow region and an upper air flow region within the overpack outlet vent. When the one or more air inlets are blocked, then the lower air flow region enables inward air flow from the outside environment, and the upper air flow region enables outward air flow to the outside environment. When the one or more air inlets are not blocked, then the upper and lower air flow regions enable outward air flow to the outside environment. As for the curved vertical plate, it extends downwardly at a right angle from the inside edge of the radial plate and has a curvature that corresponds to a curvature associated with the annular region. The curved vertical plate essentially establishes an outer annular region and an inner annular region. When the one or more inlets are blocked, the outer annular region enables inward air flow from the lower air flow region within the vent, and the inner annual region enables outward air flow to the upper air flow region of the vent. When the one or more air inlets are not blocked, the outer and inner annular regions enable upward air flow into the lower and upper air flow regions, respectfully, and then outwardly from the vent into the outside environment. An embodiment of a method, among others, for safely storing hazardous nuclear material when one or more air inlets have been fully or partially blocked to an extent that insufficient air flows into the air inlets and through the cask for adequate cooling of the hazardous nuclear material, comprises the steps of: when the one or more air inlets are not blocked, enabling air flow into the air inlets, through the annular region, and then through and out of the one or more air vents; and when the air inlets are blocked, enabling air flow inwardly through the vents, then through the annular region, and then through and out of the vents. Furthermore, another embodiment is an apparatus having a means for performing each of the foregoing steps. Other embodiments, apparatus, methods, features, and advantages of the present invention will be apparent to one with skill in the art upon examination of the following drawings and detailed description. It is intended that all such additional embodiments, apparatus, methods, features, and advantages be included within this disclosure, be within the scope of the present invention, and be protected by the accompanying claims. FIG. 1 shows a cross-sectional view of a typical, prior art, dry storage, spent nuclear fuel cask 10 having an overpack 12 with canister 14 containing a radioactive material(s) 15 stored therein with the typical air 16 into one or more air inlets 17, through an annular region 18 between the overpack 12 and the canister 14, and then out of one or more air outlet vents 22. The canister 14 in the preferred embodiment is primarily (or substantially) metal, such as stainless steel, and generally cylindrical in shape with a flat top, a flat bottom, and cylindrical sidewall. The overpack 12 in the preferred embodiment is primarily (or substantially) concrete and generally cylindrical in shape with a flat top, a flat bottom, and cylindrical sidewall. The overpack heat removal function associated with canister-based spent fuel storage relies upon natural circulation of air though the annular region 18 between the overpack vertical inner boundary and the vertical outer boundary of the metal canister 14 containing the radioactive material stored within the overpack 12. The cooler, more dense air 16 is introduced into the annular region 18 via the one of more inlets 17 where the air 16 absorbs heat which is being emitted from the radioactive material 15 in the canister 14, thereby becoming less dense and more buoyant. This increased buoyancy results in the less dense air 16 rising upward through the annular region 18 until the air 16 reaches the upper area where is exits the overpack 12 via the one or more outlet vents 22. The movement of air 16 through the annular region 18, as described, is a continuous process that results in the removal of heat from the radioactive material 15 stored within the canister 14, thereby ensuring that the temperature of the radioactive material 15 is maintained below a predetermined limit. With reference to FIG. 2, in the unlikely event that water flooding of the area where the cask 10 is stored, it is conceivable that the flood water 24 could cover the one or more overpack air inlets 17, in whole or in part, thereby interrupting the introduction of air 16 (FIG. 1) into the annular region 18. Generally, FIG. 2 shows a cross-sectional view of the cask 10 with stagnant air 16′ due to substantial blockage of the overpack inlets 17 by the flood water 24. This interruption of the flow of air 16 could result in an undesirable and dangerous increase in temperature of the radioactive material 15, potentially above desirable and/or allowable levels. The annular region 18 within the overpack 12 serves to act as a single column for air 16 to travel upward through as the air 16 absorbs heat, becoming less dense. With the blockage of the normal introduction path for cooler, less dense air 16 at the bottom of the overpack 12, this single column for air 16 becomes stagnated, thereby resulting in no means to create a thermally induced driving force based on different air densities. As illustrated in FIGS. 3 through 5, the present disclosure provides a thermal divider insert 26 that is devised specifically to address this stagnant air condition when the air inlets 17 are blocked. One of more of the thermal divider inserts 26 are installed in the overpack 12. Each thermal divider insert 26 extends through a respective air outlet vent 22 and into the annular region 18. As shown in FIG. 5, each thermal divider insert 26 is an angular plate configured in such a manner so as to have a complex right-angle appearance that is concurrently radially shaped to conform to the inner radial dimension of the overpack 12 along both vertical and horizontal surfaces. The thermal divider insert 26 can be made from any suitable materials, but in the preferred embodiment, is primarily metal, such as stainless steel. The thermal divider insert 26 can be any suitable thickness. The material and thickness should give sufficient rigidity to the structure. Furthermore, the thermal divider insert 26 is mounted via bolts, welding, or some other suitable known method. With reference to FIG. 3, the thermal divider insert 26 is installed in the overpack 12 and is configured in such a manner that the horizontal portion 28 of the insert 26 effectively divides the overpack outlet vent 22 into two distinct areas: a lower area and an upper area. The vertical plate 32 of the thermal divider insert 26 is aligned in the overpack annular region 18 between the inner boundary wall of the overpack 12 and the outer wall of the canister 14 containing radioactive material 15 stored within the overpack 12, thereby dividing the annular region 18 into two distinct areas: an inner annular region and an outer annular region. The curved vertical plate 32 extends a substantial vertical distance downwardly through the annular region 18, preferably at least half the vertical span of the annular region 18. In the preferred embodiment, the vertical plate 32 extends about sixty percent of the vertical distance of the annular region. The thermal divider insert 26 acts as a thermal material shield during normal system operation (i.e., no flood condition present that blocks the overpack inlets 17). When the one or more air inlets are not blocked, then the outer and inner annular regions enable upward air flow into the lower and upper air flow regions, respectfully, of the vents 22, and then outward air flow from the vents 22 into the outside environment. Upon blockage of the overpack inlets 17 due to flood waters (or any other postulate condition that prevents or otherwise inhibits the introduction of cooler, more dense air 16 into the overpack inlets 17), a thermal imbalance is initially encountered within the annular region 18, resulting initially in a stagnant air condition. Since the radioactive material 15 within the canister 14 will continue to emit heat, the air 16 closest to the canister 14 will continue to absorb heat, thereby creating a difference in density as compared to the air 16 closest to the inner surface of the overpack 12. As shown by the arrows in FIG. 3, due to the presence of the thermal divider insert 26, a separation of the different density air masses will be established such that the air 16 closest to the canister 14 will begin to rise due to buoyancy and will exit the overpack 12 via the upper region of the overpack outlet vent 22. Conversely, relatively cooler, more dense air will enter into the overpack 12 via the lower region of the overpack outlet vent 22, travelling downward into the outer annular region of the annular region 18, then turning inward and travelling upward within the inner annular region of the annular region 18 that has been established by the thermal divider insert 26, thereby re-establishing air flow through the annular region 18 and removing heat being emitted from the radioactive material 15 stored within the canister 14. Finally, it should be emphasized that the above-described embodiments of the present invention, particularly, any “preferred” embodiments, are merely possible nonlimiting examples of implementations, merely set forth for a clear understanding of the principles of the invention. Many variations and modifications may be made to the above-described embodiment(s) of the invention without departing substantially from the spirit and principles of the invention. All such modifications and variations are intended to be included herein within the scope of this disclosure and the present invention. As an example, it is envisioned that other embodiments of the thermal divider insert 26 of FIG. 5 can designed with a different configuration, shape, size, etc., as compared to the preferred embodiment to achieve the desired goal of establishing two separate and opposite air flows (inward air flow and outward air flow) through the respective vent and the annular region when the overpack air inlets 17 have been blocked. |
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056075197 | abstract | An impermeable, stainless steel, high chrome content container, with an inside surface coating of photo-voltaic elements comprising PN junctions upon which photons and/or electrons are impinged for conversion to electrons to produce useful electric energy; and containing an ionizing radiation energy source surrounded by an ionizable material to induce the ionizable material to emit a photon and/or electron for deposition on the PN junction elements to convert into electrical energy. |
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039506517 | claims | 1. The method of shaping a radiation beam for the treatment of patients so that the beam cross-section conforms to the target area to be treated, comprising, in combination, the steps of mixing a volume of fine grain, heavy metal powder with a quantity of pressure sensitive, non-curing adhesive so as to form a mass of plastically deformable, radiation absorbable material, compressing said material into a block-like member of sufficient area to block said beam and sufficient thickness, considering the compressed density of the material, to absorb the radiation energy of said beam, said compression having the effect of increasing the density of said member and making that member geometrically stable, removing material from said member to form an aperture therethrough having a peripheral contour corresponding to the contour of said target area, and mounting said member in said beam so that said aperture permits a beam of the desired cross sectional shape and size to fall on said target area. 2. The method of claim 1 including the step of facilitating handling of said plate by supporting the member on a radiation transparent support plate, and covering said member with a radiation transparent cover plate, so as to protect the formed periphery of said aperture. 3. The method of claim 1 in which said powder is tungsten in a grain size on the order of 200-325 mesh, and said adhesive is present in an amount in the order of 1 to 10% by weight. |
050531905 | description | Two prior art integral water cooled nuclear reactors are shown in FIGS. 1A and 1B. FIG. 1A shows a self pressurised PWR of the saturated water type. FIG. 1B shows an indirect cycle BWR variant. In both cases the nuclear reactor 10 comprises a pressure vessel 12 within which is positioned a reactor core 14. The reactor core includes a system of neutron absorbing movable control rods linked to drive mechanisms (not shown). The reactor core 14 is positioned substantially at the lower region of the pressure vessel 12, and the reactor core 14 is surrounded by a neutron reflector 16. A thermal shield 18 is positioned below the reactor core 14, and thermal shields 20 are positioned so as to surround the neutron reflector 16. The thermal shields 18,20 protect the pressure vessel 12 from radiation emanating from the reactor core 14. A primary water coolant circuit is used to cool the reactor core 14, and the primary water coolant circuit uses natural circulation. The primary coolant circuit may be provided with pumps to drive the water around the coolant circuit. The primary water coolant circuit comprises a hollow cylindrical member 22 which is aligned with and positioned vertically above the reactor core 14 to define a riser passage 24 therein for the natural vertically upward flow of relatively hot coolant from the reactor core 14, and an annular downcomer passage 40 is defined with the pressure vessel 12 for the natural vertically downward return flow of relatively cool coolant to the reactor core 14. The cylindrical member 22 does not extend completely to the top of the pressure vessel 12, and the upper region of the cylindrical member 22 is provided with apertures 26 for distributing the flow of the water coolant from the riser passage 24 to the annular downcomer passage 40. The upper end of the cylindrical member 22 has a large aperture 28 which connects the riser passage 24 to a steam space 30 formed in the upper region of the pressure vessel 12, to allow vapour from the reactor core 14 to flow into the steam space 30. A secondary coolant circuit takes heat from the primary water coolant circuit for driving an electrical turbo-generator (not shown). The secondary coolant circuit may also take heat from the primary water coolant circuit for other purposes for example process heat, district heating or propulsion systems. The secondary coolant circuit comprises a heat exchanger 32 which is annular and positioned coaxially in the upper region of the annular downcomer passage 40. The heat exchanger 32 comprises one or more tubes which are arranged in an annulus, which receive secondary coolant from a supply of secondary coolant via a supply pipe (not shown) and inlet header (not shown), and which supply heated secondary coolant to an outlet header (not shown). The outlet header supplies the heated secondary coolant via a supply pipe (not shown), for any of the above mentioned purposes. Primary water coolant descends the downcomer passage 40 passing over the outer surface of the heat exchanger tubes, and heat is transmitted to the secondary coolant inside the heat exchanger tubes. The secondary coolant used in the embodiment is water, and the heat exchanger 32 is a steam generator which comprises one or more steam tubes. The steam generator could be a once through type or a recirculatory type which has a recirculating downcomer between the outlet header and the inlet header. The steam tubes are of any suitable configuration for example as is well known in the art the steam tubes could be helically coiled tubes which extend between the inlet header and the outlet header. The helically coiled tubes may be arranged in tube bundles arranged circumferentially within the upper portion of the annular downcomer 40. British Patent No. 1386813 discloses a pressurised water reactor which has helically coiled tubes arranged in an annular steam generator, although the primary water coolant is pumped therethrough normally, natural water circulation takes place if there is a pump failure, this arrangement does not have an integral pressuriser. The steam space 30 is formed above the water level 46 in the pressure vessel 12, and a water space 44 is formed below the water level 46 in the pressure vessel 12. In the integral PWR type of nuclear reactor, shown in FIG. 1A, the heat exchanger 32 is positioned in the downcomer passage 40 wholly below the water level 46. In the indirect cycle BWR type of nuclear reactor, shown in FIG. 1B, the heat exchanger 32 is positioned in the downcomer passage 40, but an upper portion of the heat exchanger 32 protrudes above the water level 46 into the steam space 30. This promotes condensation of the primary coolant vapour in the steam space 30 on the exposed tubes of the heat exchanger 32 and bulk boiling of the primary coolant flowing through the core 14. In the integral PWR type of nuclear reactor the primary water level 46 is above the heat exchanger 32, preventing condensing heat transfer to the heat exchanger 32 and bulk boiling in the reactor core 14. In the operation of the water cooled nuclear reactor 10 the fission of nuclear fuel in the reactor core 14 produces heat. The heat is carried away from the reactor core 14 by the primary water coolant circuit. The heating of the water in the vicinity of the reactor core 14 causes the water to flow in an upwards direction as shown by arrows A through the riser passage 24, the primary water then flows through the flow distribution apertures 26 in the cylindrical member 22 into the steam generator 32 to pass over the steam generator tubes as shown by arrows B. The primary water gives heat to the secondary water in the steam generator tubes on passing through the steam generator 32. The primary water then returns to the reactor core 14 through the annular downcomer passage 40 as shown by arrow C. The steam space 30 pressurises the primary water coolant to a pressure corresponding to the saturation temperature of the reactor core primary coolant outlet temperature. As mentioned previously while operating the pressurised water reactors or boiling water reactors with this integral type of pressuriser have a certain amount of primary coolant voidage. The voidage in the primary coolant is variable and can cause unwanted perturbations in the core power level and flow distribution. Also, pressure transients can be exacerbated by positive reactivity feedback through the influence of pressure and power level on these voids with adverse effects on the pressure control and load following characteristics of the nuclear reactor. A water cooled PWR nuclear reactor 50A with integral pressuriser 94A according to the present invention is shown in FIG. 2A. The water cooled PWR nuclear reactor 50A comprises a pressure vessel 52 within which is positioned a reactor core 54. The reactor core 54 is positioned substantially at the lower region of the pressure vessel 52, and the reactor core 54 is surrounded by a neutron reflector 56. The reactor core 54 includes a system of movable neutron absorbing control rods linked to drive mechanisms (not shown). A thermal shield 58 is positioned below the reactor core 54, and thermal shields 60 are positioned so as to surround the neutron reflector 56. The thermal shields 58,60 protect the pressure vessel 52 from radiation emanating from the reactor core 54. A primary water coolant circuit is used to cool the reactor core 54, and the primary water coolant circuit uses a pumped flow or a natural circulating arrangement. The primary water coolant circuit comprises a hollow cylindrical member 62 which is aligned with and positioned vertically above the reactor core 54 to define a riser passage 64 therein for the natural vertically upward flow of relatively hot primary coolant from the reactor core 54, and an annular downcomer passage 81 is defined with the pressure vessel 52 for the natural vertically downward return flow of relatively cool primary coolant to the reactor core 54. The primary coolant circuit is also provided with one or more pumps 78, which are driven by motor 80. The pumps 78 are positioned in the downcomer passage 81. A casing 70 is positioned in the pressure vessel 52, and divides the pressure vessel 52 into a first vertically upper chamber 69 and a second vertically lower chamber 71. The reactor core 54 and the primary water coolant circuit are arranged in the lower chamber 71. The cylindrical member 62 extends towards but is spaced from the top of the lower chamber 71 defined by the casing 70, and the upper region of the cylindrical member 62 is provided with apertures 66 for the distribution of flow of the primary water coolant from the riser passage 64 to the upper part of the downcomer annular passage 81. The upper end of the cylindrical member 62 has a large aperture 68 which connects the riser passage 64 to the downcomer passage 81 and allows water and vapour from the reactor core to flow over the top of the cylindrical member 62 into the downcomer passage 81. A secondary coolant circuit takes heat from the primary water coolant circuit. The secondary coolant circuit comprises a heat exchanger 74 which is annular and positioned coaxially in the upper region of the annular downcomer passage 81. The heat exchanger 74 comprises one or more tubes which are arranged in an annulus 76, which receive secondary coolant from a supply of secondary coolant via a supply pipe (not shown) and inlet header (not shown), and which supply heated secondary coolant via an outlet header (not shown) and a supply pipe (not shown) for driving an electrical turbo-generator, for district heating, process heat or a propulsion system. The heat exchanger 74 in this example is a steam generator, and the secondary coolant used is water. The steam generator could be a once through type or a recirculatory type with downcomer pipes between the outlet and inlet headers. The casing 70 has an annular member 91 which extends vertically downwards from the peripheral region of the casing 70. The annular member 91 is spaced from the pressure vessel 52 by a small annular passage 92. The annular member 91 extends downwards to a position in the downcomer of the primary coolant circuit. The annular member 91 as shown terminates above the bottom of the heat exchanger 74. However it is equally practical for the annular member 91 to terminate at the bottom of the heat exchanger 74, at any suitable location in the downcomer passage 81 between the heat exchanger 74 and the thermal shields 60 or beneath the level of the thermal shields 60. The pressuriser 94A is positioned within the pressure vessel 52 in the vertically upper chamber 69 formed between the casing 70 and the pressure vessel 52. The pressuriser 94A forms a surge tank which contains water 104, and steam 102 separated by a water/steam interface or water level 106. One or more electrical immersion heaters 95 are provided in the pressuriser 94A positioned below the water level 106. The annular passage 92 between the annular member 91 of the casing 70 and the pressure vessel 52 forms an extension to the pressuriser water space 104 and also acts as a surge flow path for the passage of water between the pressuriser 94A and the primary water coolant circuit. The lower end of the annular member 91 contains surge ports 98 which have hydraulic diodes (not shown) to effect a relatively low resistance to the flow of water from the pressuriser 94A water space 104 to the primary water coolant circuit downcomer 81, and a relatively high resistance to the flow of water from the primary water coolant circuit to the pressuriser 94A water space 104 through the annular passage 92. It may be equally possible to arrange the size, geometry and location of the lower end of the annular member 91 so that the annular passage 92 formed with the pressure vessel 52 has these characteristics. One or more large diameter vent pipes 93 interconnect the steam space 102 of the pressuriser 94A with an upper portion of the primary coolant circuit, as shown the vent pipe 93 extends from the pressuriser steam space 102 through the pressuriser water space 104 and through the casing 70. A water cooled indirect cycle BWR nuclear reactor 50B with integral pressuriser 94B according to the present invention is shown in FIG. 2B. The arrangement of the water cooled BWR nuclear reactor 50B with integral pressuriser 94B is substantially the same as the arrangement of water cooled PWR nuclear reactor 50A with integral pressuriser 94A shown in FIG. 2A, and like parts are denoted by like numerals. The main difference is that condensing heat transfer from the primary coolant and bulk boiling in the reactor core is facilitated in the BWR variant in FIG. 2B but prevented in the PWR variant in FIG. 2A. The water cooled indirect cycle BWR nuclear reactor 50B with integral pressuriser 94B differs in that an upper portion of the heat exchanger 74 protrudes above an effective primary water coolant level 86 within the primary coolant circuit into a steam space 82 defined by the casing 70, whereas in the water cooled PWR nuclear reactor 50A the heat exchanger 74 is completely below the water level. A further difference is that the vent pipes 93 in the water cooled PWR nuclear reactor 50A with integral pressuriser 94A protrude into the primary circuit beneath the casing 70. The vent pipes are labelled 108,110,112 and 90 in FIG. 2B. The vent pipes 108,110 and 112 are of differing lengths and diameters and interconnect the steam space 102 of the pressuriser 94B with an upper portion of the primary coolant circuit, as shown the vent pipes extend into the heat exchanger region 74 but could equally well extend into the riser region 64. The vent pipes 108,110 and 112 extend to different depths into the heat exchanger 74 region or riser 64 of the primary coolant circuit, for reasons discussed later, but each one rises to the same elevation in the pressuriser 94B steam space 102. A small diameter vent pipe 90 extends from the casing 70 to the pressuriser 94B steam space 102 for the passage of incondensibles from the primary water coolant circuit steam space 82 to the pressuriser steam space 102. In the water cooled BWR nuclear reactor the annular member 91 of the casing 70 extends downwards to a position below the normal effective water level 86 in the primary circuit. The water cooled BWR nuclear reactor 50B with integral pressuriser 94B shown in FIG. 2B is not provided with pumps in the downcomer passage 81. However if there is insufficient natural circulation in the primary coolant circuit pumps may be provided to augment the natural circulation. In operation of the water cooled nuclear reactors 50A and 50B the fission of nuclear fuel in the reactor core 54 produces heat. The heat is carried away from the reactor core 54 by the primary water coolant circuit. The heating of the water in the vicinity of the reactor core 54 causes the water to flow in an upwards direction as shown by arrows D through the riser passage 64, the primary water then flows through the flow distribution apertures 66 in the cylindrical member 62 and into in the steam generator 74 cavity to pass over the steam generator tubes as shown by arrows E. The primary water gives heat to the secondary water in the steam tubes on passing through the steam generator 74. The primary water then returns to the reactor core 54 through the annular downcomer passage 81 as shown by arrow F. The function of the integral pressurisers 94A and 94B of the water cooled PWR nuclear reactor 50A and water cooled BWR nuclear reactor 50B respectively is to control the primary water coolant circuit pressure during steady state and transient conditions, to act as a surge tank for transient variations in the primary water coolant circuit inventory during normal and upset conditions, to vent excessive vapour from the primary water coolant circuit and to provide a secure reserve supply of coolant to the primary water coolant circuit automatically by gravity drain under accident conditions. In addition for the water cooled BWR nuclear reactor 50B the pressuriser 94B also regulates the effective water level in primary water coolant circuit side of the heat exchanger and hence controls the amount of boiling in the reactor core during normal operation. The casing 70 prevents the water coolant in the primary water coolant circuit from mixing with the water coolant in the pressuriser water space 104, and hence prevents interference with the performance of the pressurisers. In steady state operation, conditions in the pressurisers are essentially stagnant. Fluid communication between the primary water coolant circuit and the pressuriser occurs during transient conditions via the surge port 98 and annular passage 92 or the vent pipes 93 or vent pipes 90, 108,110, and 112. In the integral water cooled PWR nuclear reactor 50A with integral pressuriser 94A the temperature in the water space 104 of the pressuriser 94 is maintained at a higher level than that at the reactor core 54 exit by means of the electrical immersion heaters 95. Under steady state conditions thermodynamic equilibrium prevails across the steam/water interface or water level 106 of the pressuriser 94A and the primary pressure is the saturation pressure corresponding to the temperature at the water level 106. This is substantially higher than the saturation pressure corresponding to the bulk core outlet temperature. In the integral water cooled BWR nuclear reactor 50B with integral pressuriser 94B the system pressure is essentially the saturation pressure corresponding to the bulk coolant temperature at the outlet from the reactor core 54. Because of its higher elevation and because of heat losses from the steam space 102 to the surroundings, conditions in the pressuriser of 94B of the integral BWR will be at a slightly lower pressure and slightly lower temperature than in the primary water coolant circuit. However, saturation conditions prevail there also. Under steady state conditions heat losses from the pressuriser 94B are compensated by heat transfer to the pressuriser 94B from the primary water coolant circuit across the casing 70, driven by the slight temperature difference between the saturation conditions in the pressuriser 94B and in the riser passage 64 of the primary water coolant circuit. The difference in saturation pressures corresponding to the these temperatures in the pressuriser 94B and primary water coolant circuit is sufficient to cause a vapour bubble, and hence an effective water level 86, to form in the upper region of the primary water coolant circuit beneath the casing 70. The vents 108,110,112 and 90 in FIG. 2B limit or prevent the natural tendency for any steam bubble in the primary circuit beneath the casing 70 to grow indefinitely thereby pushing an excessive volume of water from the primary circuit into the pressuriser and causing the primary circuit and core to become blanketed in steam. An equilibrium is reached when the pressuriser water level and the effective water level in the primary circuit are separated by an elevation H, essentially given by: ##EQU1## where T.sub.2, T.sub.2 =Saturation temperatures in the primary water coolant circuit and pressuriser respectively. ##EQU2## .rho.=Mean density of the water in the pressuriser and surge annulus. g Acceleration due to gravity. H.sub.D =Head loss due to flow in the downcomer between the effective steam generator water level and the surge port 98 elevation. Only a small temperature difference (T.sub.2 -T.sub.1) is required to produce a large value of H. The vents 108,110,112 regulate the effective water level 86 in the primary water coolant circuit riser and steam generator cavity as follows. Vents 110,112 which protrude beneath the effective water level 86 in the steam generator region are maintained full of water by the difference in pressure between the pressuriser 94B and the primary water coolant circuit. The water column in such flooded vents extends to an elevation Hv greater than the pressuriser water level by an amount H.sub.V =H.sub.D equal to the head loss due to flow in the downcomer 81 between the effective steam generator water level 86 and the surge port 98 elevation. The vents 108,110 and 112 must protrude into the pressuriser steam space 102 by at least this amount. Thus to minimise the length of the vents the surge port 98 is located at the highest practical elevation below the effective water level 86 in the steam generator region to minimise the downcomer 81 head loss component H.sub.D. Under steady state conditions flow of coolant through such flooded vents is precluded. Vents which terminate above the effective water level 86 in the steam generator region are empty of water, and vapour in the primary water coolant circuit can flow into the pressuriser 94B steam space 102 under the action of the difference in pressure between the primary water coolant circuit and the pressuriser 94B. This flow of vapour, and enthalpy, increases the pressure and temperature in the pressuriser 94B with a resultant tendency to push water from the pressuriser 94B into the primary water coolant circuit via the annular passage 92 and surge ports 98 to increase the effective water level 86 in the primary water coolant circuit until the vent is covered or until an equilibrium water level is reached when the mass flows through the uncovered vents and surge ports 98 are balanced and the concomitant enthalpy flows and other pressuriser heat losses and gains are also balanced. The vents are sized in length, diameter, i.e. flow capacity, and in number to control the effective water level 86 in the steam generator region at an appropriate steady state level with the desired amount of intrinsic negative feedback during transients. The energy balance on the pressuriser 94B may be augmented by suitable immersion heater capacity in the pressuriser 94B water space 104 and by suitable heat removal systems in the pressuriser 94B steam space 102 for supplementary control of effective water levels 86 in the primary water coolant circuit. In the integral water cooled BWR nuclear reactor 50B with integral pressuriser 94B the vents protrude into the steam generator region of the primary water coolant circuit to control the effective primary water level 86 at an elevation lower than the top of the steam generator to promote condensing heat transfer from the primary coolant to the secondary coolant and hence bulk boiling in the reactor core 54. In the integral water cooled PWR nuclear reactor 50A, with integral pressuriser 94A a water level is prevented from forming in the primary water coolant circuit by maintaining the pressuriser 94A at a higher temperature than that at the core exit, as described earlier. In this case the vents 93 need not protrude into the steam generator region. They terminate where they enter the casing 70. In the integral PWR the vents 93 are sized to provide a means of venting vapour from the primary water coolant circuit under all accident conditions likely to be encountered. In this safety function the vents 93 prevent the primary water coolant circuit and reactor core 54 from becoming steam blanketed in the event that the pressuriser 94A heaters fail to maintain a sufficient overpressure to prevent substantial void formation in the primary water coolant in the primary water circuit. Loss of pressuriser heaters 95, excessive steam generator secondary feed water supply, secondary steam pipe rupture or other excessive secondary steam demands, loss of primary water coolant circuit pressure due to primary water coolant leaks, and excessive primary to secondary power imbalance are accident conditions which may result in void formation in the primary water coolant circuit which the vents 93 are designed to protect against. The vents 93 release such voidage and prevent primary coolant water being forced out of the primary water coolant circuit into the pressuriser water space 104 by the natural tendency for a vapour bubble to expand leading to the possible steam blanketing of the primary circuit and core 54, a characteristic of prior art pressuriser systems which exacerbates the accident conditions and impairs the ability of the primary water coolant circuit to maintain core cooling during severe accidents. The vents 93 facilitate the draining of water from the pressuriser to the primary water coolant circuit through the annular passage 92 and surge ports 98 to maintain core cooling under the action of gravity immediately and continuously as such accidents start and develop. This may preclude reactor core 54 damage during the time period before engineered safety systems can be brought into effect. To effect this safety function the pressuriser water space 104 may be designed to carry a sufficiently large reserve capacity of primary water coolant. Provision may be made for controlled depressurisation of the primary water coolant circuit during some accident conditions by venting steam in the pressuriser steam space 102 and for additional primary water coolant make up supplies to the pressuriser water space 104 or to the downcomer of the primary circuit to effect long term bleed and feed cooling of the reactor core 54 in accident conditions. The vents 108,110,112 in the integral water cooled BWR nuclear reactor 50B with integral pressuriser also have a similar safety function as in the integral PWR variant. However, in this case the deepest vents are sized to prevent vapour blanketing of the primary water coolant circuit and reactor core 54 during all accident conditions and must protrude below the range of water levels encountered during normal operation. The surge ports 98 and the annular passage 92 between the casing 70 and the pressure vessel 52 may be designed to serve this purpose by terminating them at an appropriately high elevation below the normal range of water levels 86. Again the pressuriser 94B doubles as a reserve primary water coolant supply under accident conditions. In the integral water cooled BWR nuclear reactor 50B with integral pressuriser 94B one small vent 90 is provided from the top of the primary water coolant circuit to the pressuriser steam space 102 to vent any incondensible gases which may otherwise accumulate in the steam generator cavity and impede condensation heat transfer. A further function of the vents 93 in the PWR case, or 108,110,112 in the BWR case is to provide a desuperheating spray of subcooled water into the pressuriser steam space 102 during rapid surges of primary water coolant from the primary water coolant circuit to the pressuriser 94B. Flow of water from the primary water coolant circuit to the pressuriser 94B steam space 102 is facilitated by endowing the alternative flow path via the surge ports 98 and annular passage 92 with a relatively high resistance to flow from the primary water coolant circuit to the pressuriser water space 102. This may be effected by hydraulic diodes in the annular passage 92. A rapid flow of water into the water space 104 of the pressuriser 94B results in a piston-like compression of the pressuriser steam space 102 by the rising water level 106. Surge flow diverted through the vents 93 or 108,110,112 lessens the piston effect and facilitates desuperheating of the steam space 102 by mixing. To facilitate mixing of the surge flow from the vents into the steam space the vents may be fitted with spray nozzles. This arrangement effectively taps the power imbalance during a positive primary coolant volume surge to drive the desuperheating spray flow through the vents 93 or 108,110,112 to the pressuriser steam space. The annular passage 92 and surge ports 98 have a low resistance to flow of water from the pressuriser water space 104 to the primary water coolant circuit. This facilitates augmentation of the primary water coolant inventory during negative volume surges accompanying transient reductions in primary water coolant circuit temperature or transient increases in primary water coolant circuit effective water level. A low flow resistance also facilitates gravity draining of water coolant from the water space 104 of the pressuriser into the primary water coolant circuit during accident conditions. FIGS. 3,4 and 5 show further embodiments of water cooled indirect cycle BWR nuclear reactor 50C,50D, and 50E with integral pressurisers. FIGS. 3,4 and 5 could equally apply to integral PWR variants with immersion heaters in the water spacers 102 of their integral pressurisers 94C,D and E to maintain the pressure well above that core 54 outlet saturation conditions. These three embodiments are substantially the same as the embodiment shown in FIG. 2B, but differ in that the annular member 91 of the casing 70 extends into the annular downcomer passage 81 below the heat exchanger 74. In FIG. 3, the annular member 91 terminates at the bottom of the heat exchanger 74, in FIG. 4, the annular member 91 terminates between the heat exchanger 74 and the thermal shields 60, and in FIG. 5, the annular member 91 terminates below the thermal shields 60. A further difference, in FIGS. 3 and 4 is that the vents 108,110 and 112 are longer to allow for the greater elevation H.sub.v of the standing water columns in the flooded vents 110 and 112 and the pressure vessel 52 is increased in height to accommodate the longer vents 108,110 and 112. In FIG. 5, the surge ports 98 at the downstream end of the annular member 91 enter the primary water coolant circuit downstream of the pumps 78. The separation between the water levels in the pressuriser 94B and primary water coolant circuit is given by: ##EQU3## where H.sub.P is the pump head. The water level in the flooded vents 108,110 and 112 differs from the pressuriser water level 106 by an amount H.sub.V =H.sub.D -H.sub.P. Thus the water in the vents 108,110 and 112 is drawn below the pressuriser water level. A further embodiment of a water cooled nuclear reactor 50F with integral pressuriser 94F according to the present invention is shown in FIG. 6. This is an indirect cycle boiling water reactor although it is applicable to a pressurised water type. The water cooled nuclear reactor 50F again comprises a pressure vessel 52 within which is positioned a reactor core 54. The reactor core 54 is positioned at the lower region of the pressure vessel 52. The reactor core 54 includes a system of movable neutron absorbing control rods linked to drive mechanisms (not shown). The reactor core 54 is surrounded by a neutron reflector 56. A thermal shield 58 is positioned below the reactor core 54 and thermal shields 60 are positioned so as to surround the neutron reflector 56. The thermal shields 58 and 60 protect the pressure vessel 52 from radiation emanating from the reactor core 54. A primary water coolant circuit is used to cool the reactor core 54, and the primary water coolant circuit uses a natural circulating arrangement. Pumps (not shown) may be provided in the downcomer 81 or beneath the thermal shields 60 to enhance the flow of coolant through the reactor core 54. The primary water coolant circuit comprises a hollow cylindrical member 62 which is aligned with and positioned vertically above the reactor core 54 to define a riser passage 64 therein for the natural vertically upward flow of relatively hot primary coolant from the reactor core 54, and an annular downcomer passage 81 is defined with the pressure vessel 52 for the natural vertically downward return flow of relatively cool primary coolant to the reactor core 54. A casing 70 is positioned in the pressure vessel 52, and divides the pressure vessel 52 into a first vertically upper chamber 69 and a second vertically lower chamber 71. The reactor core 54 and the primary coolant circuit are arranged in the lower chamber 71. The cylindrical member 62 extends towards but is spaced from the top of the lower chamber 71 defined by the casing 70, and the upper region of the cylindrical member 62 is provided with apertures 66 for the distribution of flow of the primary water coolant from the riser passage 64 to a heat exchanger 74 in the annular downcomer passage 81. The upper end of the cylindrical member 62 has a large aperture 68 which connects the riser passage 64 to a steam space 82 formed in the upper region of the lower chamber 71 defined by the casing 70 and the pressure vessel 52. A secondary coolant circuit takes heat from the primary water coolant circuit. The secondary coolant circuit comprises a heat exchanger 74 i.e. a steam generator 74 which is annular and positioned coaxially in the upper region of the annular downcomer passage 81. The steam generator 74 comprises one or more steam tubes, which are arranged in the annular cavity 76, and which receive water from a supply of water via a supply pipe and inlet header, and which supply steam to a steam turbine via an outlet header and a supply pipe (not shown). A steam space 82 is formed above the water level 86 and a water space 84 is formed below the water level 86 in the steam generator 74. The casing 70 has one or more vents 90 for incondensibles at its highest point, and the casing 70 has an annular member 91 which extends vertically downwards from the peripheral region of the casing 70. The bottom region of the annular member 91 is secured and sealed to the pressure vessel 52. The annular member 91 extends downwards to a position below the normal water level 86 in the steam generator 74 region. The pressuriser 94F is positioned within the pressure vessel 52 in the vertically upper chamber 69 formed between the casing 70 and the pressure vessel 52. The pressuriser or surge tank which contains water and steam is defined by the pressure vessel 52, the casing 70 and the annular member 91, and a water space 104 is formed below the water level 106 and a steam space 102 is formed above the water level 106 of the pressuriser. The bottom region of the pressuriser is provided with a plurality of circumferentially arranged surge ports 154 which are formed in the annular member 91 of the casing 70. The surge ports 154 fluidly communicate between the pressuriser water space 104 and the annular downcomer passage 81 of the primary coolant circuit, and as shown the surge ports 154 extend into the steam generator 74 region of the primary coolant circuit. The surge ports 154 have low flow resistance for water from the surge tank to the primary coolant circuit; and have high flow resistance for water from the primary water coolant circuit to the surge tank. The surge ports 154 as shown are re-entrant nozzles, but suitable hydraulic diodes or valves could be used to perform this task. A number of vent pipes 162 interconnect the steam space 102 of the pressuriser 94F with an upper portion of the primary coolant circuit, as shown the vent pipes 162 connect to ports 166 formed in the annular member 91 of the casing 70. The ports are circumferentially arranged and are positioned at the water level in the primary coolant circuit as shown. They determine this water level. The water cooled nuclear reactor 50F operates substantially the same as that in FIG. 2B. This arrangement also can be used as in the embodiment of FIG. 2A as an integral pressurised water reactor. The embodiment shown in FIG. 7 is a water cooled PWR nuclear reactor 50G with integral pressuriser and is substantially the same as the embodiment in FIG. 2A, but the bottom region of the annular member 91 is sealingly secured to the pressure vessel 52 and surge ports 254 are provided. The surge ports 254 extend into the downcomer passage 81 below the heat exchanger 74. The surge ports 254 are reentrant nozzles, but other suitable hydraulic diodes could be used. A single vent pipe 93 is provided which again interconnects the steam space 102 of the pressuriser 94G with the upper portion of the primary coolant circuit. The vent pipe 93 is of increased length for the same reason as the vent pipes in FIGS. 3 and 4. This arrangement may also be used as an embodiment of an integral BWR with vents as in FIG. 5. A further embodiment of a water cooled nuclear reactor 50H with integral pressuriser 94H according to the present invention is shown in FIG. 8. This is an indirect cycle BWR nuclear reactor and is substantially the same as the embodiments in FIGS. 2B,3,4,5 and 6. In this example a casing 264 is secured to and seals with the pressure vessel 52 to divide the pressure vessel 52 into an upper chamber 69 and a lower chamber 71. The casing 264 has an annular member 265 which extends downwards therefrom about the axis of the pressure vessel 52, and the member 265 is closed at its bottom end. The annular member 265 extends coaxially into the cylindrical member 62 and is spaced therefrom. The pressuriser 94H is positioned within the vertically upper chamber 69, formed between the casing 264 and the pressure vessel 52. A pressuriser or surge tank which contains water and steam is defined by the pressure vessel 52, the casing 264 and the annular member 265 and a water space 104 is formed below the water level 106 and a steam space 102 is formed above the water level 106 of the pressuriser or surge tank. The bottom region of the annular member 265 is provided with a plurality of surge ports 454 which fluidly communicate between the surge tank water space 104 and the annular downcomer passage 81 of the primary coolant circuit. The surge ports 454 extend through the cylindrical member 62, and into the steam generator 74. The surge ports 454 have low flow resistance for water from the surge tank water space to the primary coolant circuit, but have high flow resistance for water from the primary coolant circuit to the surge tank water space. A number of vent pipes 262 interconnect the steam space 256 of the pressuriser 94H with an upper portion of the primary coolant circuit, as shown the vent pipes 262 extend through annular member 265 and connect to ports 266 formed in the cylindrical member 62. The ports 266 are circumferentially arranged and are positioned at the effective water level in the primary coolant circuit. They determine this water level. In the arrangement the water flowing from the pressuriser water space to the steam generator cavity of primary water coolant circuit during a negative value surge is relatively hot and this enhances the thermal inertia of the plant during power demand transients cushioning the transient steam conditions experienced by the second coolant circuit. This arrangement could also be used in an embodiment of an integral PWR with vents as in FIG. 2A. The embodiment of water cooled nuclear reactor 50J with integral pressuriser 94J shown in FIG. 9, is an indirect cycle BWR nuclear reactor, and is substantially the same as the embodiment in FIG. 5, the annular member 91 of the casing 70 extends downwards below the reactor core 54 or the thermal shield 58. The casing 70 also comprises a bottom member 354 which is sealingly secured to or formed integrally with the bottom end of the annular member 91, and a surge port 98 is formed in the bottom member 354 beneath the reactor core 54. A plurality of vent pipes 362 interconnect the steam space 102 of the pressuriser 94J with the upper region of the primary water coolant circuit below the water level 86. Thus the casing 70 is completely enclosed by the pressure vessel 52, and the casing 70 divides the pressure vessel 52 into a first outer chamber 69 formed between the casing 70 and the pressure vessel 52, and a second inner chamber 71 formed within the casing 70. The downcomer passage 81 is formed between the annular member 91 of the casing 70 and the cylindrical member 62. The pressure vessel 52 is spaced from the annular member 91 to form an annular passage 92 which forms a lower portion of the water space 104 of the pressuriser 94J. The upper portion of the water space 104 of the pressuriser or surge tank 94J is above the casing 70. The annular passage 92 and surge ports 98 have a low resistance to flow of water from the pressuriser water space 104 to the primary water coolant circuit. This facilitates augmentation of the primary water coolant inventory during negative volume surges accompanying transient reductions in primary water coolant circuit temperature and transient increases in primary water coolant circuit effective water level. A low resistance also facilitates gravity draining of water coolant from the water space 104 of the pressuriser into the primary water coolant circuit during accident conditions. A major function of the vents is to prevent steam blanketing of the primary circuit and reactor core under all circumstances and to facilitate gravity drain of the water in the pressuriser into the primary circuit and core under accident conditions resulting in severe reductions in primary coolant pressure. To facilitate this latter function the pressuriser water space may be sized to carry a large reserve of coolant which is readily and continuously available to maintain the reactor core submersed during the development of severe accidents resulting in the loss of normal cooling or a severe reduction in primary system pressure. To enhance this function the surge ports may be located beneath the reactor core as in FIG. 10. In this embodiment the primary water coolant circuit is effectively submerged in the pressuriser water space. The annular passage of the water space may contain coolers to maintain a stratified temperature distribution in the water space of the pressuriser cum surge tank. Hot water above and cooler water below. This arrangement facilitates a rapid core response to increases in steam demand. The resultant negative volume surge draws in cool water from the pressuriser 94J leading to a rapid core response through the negative temperature coefficient of reactivity. The arrangement also enhances the gravity flow of water from the pressuriser to the core during accident conditions which may lead to void formation in the primary circuit. This arrangement could also be used in an embodiment of the integral PWR type with vents as in FIG. 2A. FIGS. 10A and 10B show further alternative embodiments of water cooled nuclear reactors, FIG. 10A shows an integral pressurised water reactor 50K and FIG. 10B shows an integral indirect cycle boiling water reactor 50L. In these two embodiments the pressure vessel 52 is divided into an upper chamber 69 and a lower chamber 71 by a casing 270 which is secured to and seals with the pressure vessel 52. In the integral pressurised water reactor variant in FIG. 10A, a single vent pipe 93 extends upwards from the casing 270 and interconnects the steam space 102 of the pressuriser 94K with the upper portion of the primary coolant circuit, and one or more surge pipes 272 extend into the downcomer passage 81 below the heat exchanger 74. In the integral boiling water reactor variant in FIG. 10B vent pipes 90,108 and 110 also protrude into the stream space 102 of the pressuriser 94L at their upper ends and also protrude by differing distances into the primary coolant circuit to facilitate the primary coolant circuit level regulation in the condensing steam generator, and one or more surge pipes 272 extend into the downcomer passage 81 below the heat exchanger 74. In both these variants the surge pipes may be fitted with hydraulic diodes to facilitate a desuperheating spray of primary coolant through the vent pipes into the pressuriser steam space during a load reducing transient, under the driving action of the primary coolant circuit volume surge caused by the transient imbalance of power between the reactor and steam generator which occurs during such transients. In the integral BWR variant, the surge pipes 272 must extend downwards to an elevation lower than the normal operating effective primary coolant circuit water level. But in the integral PWR variant, such a requirement does not apply, as the casing 270 normally defines the water level, however it may be desirable to allow the vent pipes to extend a small distance below the casing 270 to prevent any tendency for steam from the pressuriser 94K steam space 102 to be drawn down the vent pipes 93 into the primary water coolant circuit by the negative volume surge accompanying a power increasing transient. An essential difference between the integral PWR and BWR variant is that the former has electric immersion heating capacity in the pressuriser water space 104 which maintains saturation conditions in the pressuriser substantially higher than that corresponding to the reactor core outlet temperature, while in the latter the converse applies. The vent pipes 93,108,110 perform the multiple functions of pressuriser spray, pressuriser stabilisation, i.e. vapour venting of the primary coolant circuit during accidents and primary coolant circuit water level control, i.e. for the BWR variant only. The pressuriser spray function may be provided separately with separate vents to provide the other functions. FIG. 10C illustrates the pressuriser spray function of an integral PWR 50K The surge pipes 272 are provided with a hydraulic diode 274 to interconnect the water space 104 of the pressuriser 94K and the riser 64 of the primary coolant circuit. One or more of the vent pipes 93 are provided with a spray nozzle 276, and an auxiliary conventional spray nozzle 278 is also provided in the pressuriser 94K. During a positive volume surge, occasioned by load rejection say, primary coolant water from the primary coolant circuit is forced into the pressuriser 94K through the surge pipe 272, control rod guide tubes and the vent pipes 93. Flow restrictions in the control rod guide tubes and the hydraulic diode 274 in the surge pipe 272 limit the flow through these paths, the bulk of the volume surge thus flows through the vent pipes 93 and spray nozzle 276 to facilitate the spray function. During a negative volume surge occasioned by a power increasing transient, water flows from the pressuriser 94K water space 104 into the primary coolant circuit through the surge pipes 272 and hydraulic diodes 274 which offer a low resistance to flow in this direction. Some water may also flow down the control rod guide tubes. Any tendency for steam to be drawn down the vent pipe 93 is offset by allowing the vent pipe 93 to protrude some distance below the casing 270 into the primary coolant circuit. In the case of an integral PWR having reactor primary coolant pumps to provide forced circulation of the primary coolant circuit it may be necessary to adjust the relative positions of the surge pipes and vent pipes in the primary coolant circuit. This is because some arrangements may result in a component, i.e. pump, steam generator or core, pressure drop between the surge pipe and vent pipe connections in the primary coolant circuit and a resulting tendency for an unwanted primary coolant flow through the vent pipes/spray pipes, pressuriser and surge pipes during normal steady state operation of the plant. In FIG. 10C such a flow is prevented by locating both the surge pipe and vent pipe/spray pipe connections to the primary coolant circuit in the riser. However the spray function may also be provided by locating the surge pipes and vent pipes/spray pipes connections to the primary coolant circuit in the downcomer above or within the steam generator. FIG. 11A,11B,11C, and 11D show further embodiments of an integral PWR and integral indirect cycle BWR. In FIGS. 11A and 11B a number of circumferentially spaced once through steam generator modules 74 are located in an annulus formed by the hollow cylindrical member 62 and the pressure vessel 52. The steam generator modules 74 comprise a number of steam generator tubes which extend through an annular shroud 276. These shrouds 276 are interconnected at their upper ends, and are secured and sealed to the pressure vessel 52 and the hollow cylindrical member 62 to prevent the flow of primary coolant through the interstitial regions of the downcomer 81 between the steam generator 74 shrouds 276. Thus the primary coolant flows through the shrouds 276 to facilitate effective primary coolant flow distribution through the steam generator modules 74. Reactor coolant pumps 80 are located at the top of the downcomer 81 above, upstream, of the steam generators. A number of surge pipes 272 with hydraulic diodes 274 are provided to interconnect the water space 104 of the pressurizer 94M with the primary coolant circuit. One or more spray pipes 278 communicate between the riser 64 of the primary coolant circuit and the steam space 102 of the pressurizer 94M and one or more vent pipes 93 communicate between the riser 64 and the steam space 102 of the pressuriser 94M. FIGS. 11C and 11D illustrate the integral indirect cycle BWR variant of the embodiment in FIGS. 11A and 11B. This also has the steam generator molecules 74 arranged in shrouds 276 between the pressure vessel 52 and the hollow cylindrical member 62. Reactor coolant pumps are dispensed with in this variant, primarily primary coolant circuit flow being induced entirely by natural convection. One or more surge pipes 272 communicate between the water space 104 of the pressuriser 94N and the interstitial regions of the downcomer between the steam generator modules 74, the interstitial region effectively forming part of the surge pipe. A number of primary coolant circuit water level control vent pipes 108,110,112 are provided, and an incondensible vent 90. The vent pipes 110 descend through the riser 64 and penetrates through the hollow cylindrical member 62 and the shrouds 276 to effect primary coolant circuit water level control at this elevation during say high power operation. The vents 112 descend through the interstitial region and penetrate the shrouds 276. The vent 108 terminates in the common plenum region above the shrouds 276 to control the primary coolant circuit water level at a higher elevation at low power operation. An advantage of an integral arrangement of water cooled nuclear reactor is that a propensity exists for the coolant to circulate around the primary circuit and through the core by natural convection. This is a desirable safety characteristic which may be used during power operation in some circumstances. The presence of voidage in the riser due to condensing heat transfer in the steam generator particularly enhances the propensity for natural circulation in the integral indirect cycle BWR variant. Thus reactor circulation pumps may not be needed for some designs of this variant. However, provision for reactor circulation pumps in the downcomer or beneath the thermal shields may be provided in both the BWR and PWR variants. The main features of the present invention are the casing separating the pressuriser and the primary circuit within the same pressure vessel and the system of vents from the upper region of the primary water coolant circuit to the pressuriser steam space. The venting concept extends to the case of a separate pressuriser also. FIGS. 12 and 13 show integral water cooled nuclear reactors with separate pressurisers. FIG. 12 shows an integral indirect cycle BWR nuclear reactor 150 and FIG. 13 shows an integral PWR nuclear reactor 250. In FIG. 12 an indirect cycle BWR nuclear reactor 150 is shown, and a separate external pressuriser 494. A number of vent pipes 508,510 and 512 interconnect a steam space 502 in the pressuriser 494 with an upper portion of the primary water coolant circuit in the region of the heat exchanger 74. The vent pipes extend to different depths into the primary water coolant circuit, but all extend to the same elevation in the pressuriser 494. A vent 490 interconnects the steam space 82 of the BWR nuclear reactor with the steam space 502 of the pressuriser 494 for the flow of incondensibles. A pipe 514 interconnects a water space 504 of the pressuriser 494 with a downcomer passage 81 of the primary water coolant circuit via a surge port 516. In FIG. 13 an integral PWR nuclear reactor 250 is shown, and a separate external pressuriser 594. A single vent pipe 593 interconnects a steam space 602,608 in the pressuriser 594 with an upper portion of the primary water coolant circuit, and a pipe 614 interconnects a water space 604 of the pressuriser 594 with a downcomer passage 81 of the primary water coolant circuit via a surge port 616. The pressuriser 594 shown, is for use when the head of water HV=HD is higher than the pressuriser 594. The pressuriser 594 comprises a main vessel 596 and a secondary vessel 598. The secondary vessel 598 is provided at the upper end of the vent pipe 593, and the secondary vessel 598 has a steam space 608 and a water space 610 separated by a water level 612. The secondary vessel 598 is interconnected with the main vessel 596 by a pipe 611. The pipe 611 connects with the secondary vessel 598 at the water level 612, and connects with the main vessel 596 above the water level 606. Steam condensing in the secondary vessel 598 drain from the water space 610 through pipe 611 into the water space 604 of the main vessel 604. In FIG. 14, the present invention is shown applied to a single loop dispersed PWR nuclear reactor 350, although the invention is applicable to arrangements with two, three or four loops, which are more typical. The dispersed PWR nuclear reactor 350 comprises a pressure vessel 752 within which is positioned a reactor core 754. The reactor core includes a system of neutron absorbing movable control rods linked to drive mechanisms (not shown). A primary water coolant circuit is provided to cool the reactor core 754. The primary water coolant circuit comprises a riser plenum or reactor core exit plenum 756, a first pipe 760 which conveys relatively hot water to an inlet header 762 in a heat exchanger 774 i.e. a steam generator. The inlet header 762 supplies the hot water through a bank of steam generator tubes 764 to an outlet header 766. The relatively cool water is returned through a second pipe 768 to a downcomer 781 which returns the now cool water via a reactor core inlet plenum 755 to the reactor core 754. A pump 770 is provided to drive the water through the primary water coolant circuit and the pump 770 is driven by a motor 772. The heat exchanger or steam generator 774 in this example is positioned outside of the pressure vessel 752 which contains the reactor core 754, and only a portion of the primary water coolant circuit is contained within the pressure vessel 752. Although only one heat exchanger or steam generator 774 is shown in FIG. 14, two, three or four heat exchangers or steam generators may be provided together with respective pipes 760 and 768. A separate pressuriser 694 is provided to maintain the primary water coolant in the primary water coolant circuit at a high pressure so that high primary water coolant temperatures can be achieved without the primary water coolant boiling. The pressuriser 694 comprises a separate pressure vessel 696 containing a steam space 702 and a water space 704 separated by a water level 706. A surge pipe 710 interconnects the water space 704 of the pressuriser 694 with the downcomer 781 of the primary water coolant circuit or the reactor core inlet plenum 755. The surge pipe 710 has a surge port 712. A vent pipe 708 interconnects the riser plenum or reactor core exit plenum 756 with the steam space 702 of the pressuriser 694. The surge port 712 may incorporate a hydraulic diode to give a low resistance to surge flow from the pressuriser water space 704 to the primary water coolant circuit, and a high resistance to surge flow from the primary water coolant circuit to the water space 704 of the pressuriser, so that a substantial surge flow passes through the vent pipe 708 from the primary water coolant circuit to the pressuriser steam space 702 to effect desuperheating of the pressuriser steam space 702 during positive volume surges. A major function of the integral or separate pressuriser of this invention is that the generation of vapour in the reactor core or reactor pressure vessel cannot flood the pressuriser and steam blanket the primary circuit and reactor core. A further advantage is that a lower system pressure can be used than with the simple pressuriser/surgeline arrangement of the prior art. The pressure need only be high enough to suppress, limit or control boiling at the prevailing reactor core outlet temperature, to suit the design of the reactor core installed. Conversely, for a given system pressure the reactor core outlet temperature and secondary steam conditions can be maximised. In the case of the Integral boiling water reactor core outlet temperatures of 300.degree. C. could be obtained with 86 bars pressure i.e. 8.times.10.sup.6 Nm.sup.-2 compared to pressure in excess of 150 bars i.e. 15.times.10.sup.6 Nm.sup.-2 for a dispersed pressurised water reactor. The arrangements according to the present invention are unconditionally stable with respect to large and small perturbations in the primary circuit. They facilitate both integral PWR and integral indirect cycle BWR variants. They prevent vapour locking of the reactor core and primary water coolant circuit. They facilitate natural convection cooling of the reactor core during normal and accident conditions. They facilitate effective pressure control and regulation of integral water cooled nuclear reactors. |
summary | ||
048448397 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The apparatus A used in carrying out the insitu method of treating hazardous waste impoundment includes a power operated vehicle B that is preferably mounted on caterpillar track C to permit the vehicle to travel over soft ground. A boom D is pivotally supported from the vehicle B and extends upwardly and outwardly therefrom as shown in FIG. 1. The boom D is angularly adjustable relative to the vehicle by conventional means (not shown). The boom D has an outer end 10 that rotatably supports a pulley 12 over which a winch operated first cable 14 passes to extend downwardly to a support 16. An elongate vertical framework E is provided that has an upper end portion 18 and a lower end portion 20. A number of second cables 22 extend downwardly from the support 16 and are secured to the upper end portion 18 of the framework. A horizontal vertically movable platform F is disposed within the framework E and is moved upwardly and downwardly by a number of spaced hoist and crown chain belts 24 that engage upper and lower sprockets 26 mounted on the framework E, but with only the upper sprockets being shown. The platform F is secured to one of the reaches of the belts 24 by conventional means 25. The sprockets 26 are secured to shafts (not shown) that are driven by motors 28 as may be seen in FIG. 1. A number of inverted cup shaped housings G are supported on the platform F as may be seen in FIG. 2 and are arranged in four rows, each of which rows includes four housings. Each housing G includes an upper end 30 from which a hook 32 extends downwardly to support a conventional oil well swivel H. A number of hollow kelly stems J are provided and have their upper ends rotatably supported by the swivels H as shown in FIG. 3. Each swivel H has a first hose 34 connected thereto as shown in FIG. 3, with the hose being in communication with a passage 36 that extends downwardly in one of the kelly stems J. Each of the hoses 34 is connected to a tubular member assembly 38 that is in communication with an elongate manifold 40. A second hose 42 is connected to a centered opening 44 in manifold 40, with the second hose extending to a reel 46. Reel 46 is supported on the upper portion of framework E. A third hose 47 extends from reel 46 downwardly alongside framework E to a facility K that serves to store chemicals, chemical blending and proportioning apparatus, a compressor, and pump for discharging dry chemicals and chemical solutions into the third hose 47. This facility is not shown in detail as all of the equipment therein is conventional and may be purchased in the present day commercial equipment market. The vehicle B has stabilizing members 48 and 50 extending therefrom to the framework E. A rotary table L is mounted on the lower portion of the framework E and as may be seen in FIG. 5 is defined by two parallel, vertically spaced plates 52 that are secured together in fixed relationship by conventional means (not shown). One half of the rotary table is shown in FIG. 4, with the other half being of the same structure. A number of ball bearings 54 engage grooves 56a in a pair of ring gears 56. The ball bearings 54 rotatably support the pair of ring gears 56 between the plates. Ring gears 56 have interior and exterior teeth 56a and 56b. Exterior teeth 56b are in engagement as shown in FIG. 4. Eight sprockets 58 are disposed between the plates 52 and are arranged in two rows of four sprockets each. Each sprocket 58 includes a hub 58a that is rotatably supported in a bearing 60 that engages the pair of plates 52. The bearings 60 are held in place on the rotary table L by retaining members 62 that are secured to plates 52 by bolts 64. Each of the hubs 52a has a sleeve 66 extending therethrough, with the sleeve having a passage of square transverse cross section therein that is slidably engaged by one of the hollow kelly stems J that is also of square transverse cross section. A driving gear 68 engages the external threads 56b of one of the ring gears 56 as shown in FIG. 4, with the gears being rotated by a motor 70 shown in FIG. 1. The half of the rotary table L (not shown) in FIG. 4 is of the same structure as that illustrated and is also driven by a second motor (not shown). The framework E has pads 72 secured to the lower portion thereof to permit the framework to be rested and supported on solid ground. A shroud 74 of pliable material extends downwardly from the periphery of the rotary table L as shown in FIG. 1 and is preferably of a length to extend below the pads 74 when the shroud is fully stretched out. Each kelly stem J supports a cutter-injector M on the lower end thereof, the detail structure of one of the cutter-injectors being shown in FIG. 6. Each cutter-injector M includes a rigid elongate vertically disposable member 76 that has the upper end secured to a tubular collar 70a by welding beads 80, and the collar in turn secures the lower end of a kelly stem J by second welding beads 82. Two first straight straps 84 of opposite pitch are axially aligned and extend outwardly in opposite directions from the member 76. The straps 84 on the outer ends develop into second straps 86 of arcuate shape that extend downwardly and inwardly to be secured to the member 76. The second straps 86 are also of opposite pitch. The hollow tubular member 76 has a bladed auger 88 secured to the lower end thereof. A tubular member 90 extends outwardly from the hollow vertical member 76 adjacent the first straps, and serves to have hazardous waste treatment chemicals discharged outwardly therethrough. Treatment of an impoundment P containing hazardous waste R is carried out by the apparatus A, which apparatus is illustrated as a crane in FIG. 1, in the following manner. The apparatus A is moved to land S adjacent an impoundment P as shown in FIG. 9 and the boom D extending outwardly over the impoundment to position the framework E thereover as illustrated in FIG. 10. The kelly stems J are concurrently rotated and the platform F lowered to allow the cutter-injectors M to move downwardly through the hazardous waste R at a station T, a series of which adjacent stations are shown in FIG. 10. Each of the cutter-injectors M as it rotates cuts and intermixes the waste R in a circular downwardly extending zone R-1 as shown in phantom line in FIG. 8. The adjacently disposed cutter-injectors M are of such transverse dimensions that the zones of one R-1 overlap one another as illustrated by phantom line in FIG. 8, and it is to these zones that the treatment chemicals are discharged. The assembly K is used to discharge appropriate chemicals either dry, wet, or gaseous at a desired rate, to the cutter-injectors M, which discharge if desired may take place as the zones R-1 are formed to a desired depth. The augers 88 allow a hard material to be penetrated as the cutter-injectors move downwardly through the waste R. The waste R that is mixed to a uniform consistancy in the zones R-1 remains in place therein, and due to the pitch of the straps 84 and 86 and the opposite rotation of adjacent cutter-injectors M the waste in one zone R-1 will be subjected to an upward force and the waste in an adjacent zone a downward force to obtain optimum intermixing of the waste. The discharging chemicals flow from the tubular member 90 as the cutter-injectors M rotate. After the zones R-1 have been formed the platform F is moved upwardly to cause the cutter-injectors M to rise through the zones R-1 with continued rotation of the cutter-injectors. If chemicals have not been injected into the waste R as the zones R-1 are formed, the chemicals are injected on the upward movement of the cutter-injectors M through the zones R-1. On occasion it may be desirable to inject chemicals into zones R-1 both as they are formed, and as the cutter-injectors are moved upwardly therethrough. Vapors, odors or omissions from the waste R that are not chemically destroyed during the formation of the zones R-1 and the injection of chemicals therein, are collected in the confined space within the end shroud 74 and discharged through a conventional scrubber U to be removed, after which air free of the objectionable omissions is caused to flow through a conduit 92 for discharge to the ambient atmosphere of a desired location. A second form of apparatus U is shown in FIGS. 11-13 that may be used in the detoxifying of a hazardous waste impoundment Y and is the best mode for accomplishing this result. Apparatus U includes a power operated vehicle V which in FIGS. 10-13 is illustrated as a caterpillar type tractor that movably supports a vertically extending frame W and an instrumentation and control cab 100. The frame W extends vertically and is of an open elongate shape. The frame W as shown in FIG. 11 is defined by four elongate corner members 102 between which cross pieces 104 and reinforcing member 106 extend. The frame W includes an upper platform 108 and lower platform 110 rigidly secured thereto. An intermediate platform 112 shown in FIG. 12 is situated within the frame W and is vertically movable relative thereto. The frame W is vertically movable relative to a support assembly 114. The support assembly 114 has a number of elongate support members 116 extending therefrom to the vehicle V as shown in FIG. 13. The support members 116 at their outer ends are secured to support assembly 114 by pivotal connections 116a and to the vehicle V by pivotal connections 116b. A counterweight 118 by a conventional linkage assembly 120 is movably supported from vehicle V on the side thereof opposite that from which frame W is supported. A first hydraulic cylinder assembly 122 is pivotally connected to the vehicle V and linkage assembly 120 to permit lateral movement of frame W and support assembly 114 relative to vehicle V when the hydraulic cylinder assembly is activated. A second hydraulic cylinder assembly 124 is pivotally connected to support assembly 114 and frame W to permit vertical movement of frame W relative to the support assembly 114 and vehicle V when the second hydraulic cylinder assembly is activated. A confined space defining shroud X extends downwardly from lower platform 110 into which a power driven rotatable cutting blade Z is vertically movable. A laterally spaced pair of motors 126 are mounted on lower platform 110 and rotate drive sprockets 128. Each drive sprocket 128 engages an upwardly extending endless link belt 130 that rotatably engages a pair of sprockets 132 rotatably supported from upper platform 108. Intermediate platform 112 is secured by conventional fastening means 112a to a vertical reach 130a of belt 130. The lower platform 110 has two pairs of electric motors 134 mounted thereon that rotate driving sprockets 136 as shown in FIGS. 14 and 20 that are in toothed engagement with a pair of driven gears 138, which gears are also in toothed engagement. The lower platform 110 as may be seen in FIG. 15 is defined by an upper horizontal plate 110a and lower plate 110b. Each gear 138 is ring shaped and is rotatably supported by a sequence of ball bearings 140 from a ring shaped mounting assembly 142 that is secured to lower plate 110b by bolts 144 as shown in FIG. 15. Each driven gear 138 has a flat rigid ring shaped member 146 secured to the upper surface thereof by bolts 148 as shown in FIG. 15. Each member 146 has a cylindrical sleeve 150 projecting upwardly therefrom and passing through an opening 152 in upper plate 110a. In FIG. 15 it will be seen that each sleeve 150 has a flange 154 projecting outwardly therefrom that supports a seal 156 in sliding contact with the upper surface of upper plate 110a. In FIGS. 14 and 16 it will be seen that two spaces pairs of rollers 158 are rotatably supported above upper plate 110a from lugs 160 that are secured to members 146. The intermediate platform 112 as may be seen in FIG. 15 is defined by upper and lower vertically spaced rigid ring shaped horizontal plates 112a and 112b that are joined by connectors 162. Two tubular Kellys 164 used in driving blades Z have upper end portions 164a disposed within intermediate platform 112. Each end portion 164a has an outwardly extending flange 166 secured thereto, which flange has an externally grooved ring shaped member 168 secured thereto that rotatably engages a sequence of ball bearings 170 that engage an internally grooved ring shaped member 172 secured to the upper plate 112a of intermediate platform 112. Kelly 164 has two oppositely disposed vertically extending ribs 174 projecting outwardly from the external surface thereof as shown in FIG. 16, which ribs are rotatably engaged by the two pairs of rollers 158. Kelly 164 has a horizontal member 164a secured to the lower end thereof that supports a centrally disposed tubular member 176 of substantially smaller diameter than that of Kelly 164. Tubular member 176 serves as a mounting for a tube 178 that extends upwardly in Kelly 164, which tube has an outwardly extending seal 180 on the upper end thereof. The lower end of tubular member 176 develops into an outwardly extending flange 182. Tubular member 176 rotatably supports a bushing 184 between member 164a and flange 182. The two bushings 184 rotatably engage cylindrical shells 186 that are connected by arms 188 of an open rectangular frame 190. The frame 190 on the periphery thereof supports a conduit 192 that has spray heads 194 mounted thereon, the purpose of which will later be explained. Cutting blade Z illustrated in FIG. 15 includes an outer tube 196 that has a pointed lower end 196a and the upper end of the tube being secured to a circular plate 198 that has a centered opening 198a therein. An inner tube 200 is secured to plate 198 and is in communication with opening 198. Inner tube 200 on the lower end develops into a discharge nozzle 202 that extends through outer tube 196. Two oppositely disposed cutting blades 204 extend outwardly from the lower end of outer tube 196 and support a number of spaced teeth 206. Two arcuate cutting members 208 extend upwardly from the outer ends of blades 204 to outer tube 196 as shown in FIG. 15. Circular plate 198 is secured to flange 182 by conventional means such as bolts 210 or the like. The seal 180 engages the interior surface of an intermediately positional tube 212 that extends downwardly between Kelly 164 and inner tube 178. A tube extension 214 projects upwardly from tube 212 and is secured thereto by a ring shaped end piece 216 as shown in FIG. 15. In FIG. 16 it will be seen that Kelly 164 has two groove defining ribs 218 on the interior thereof that slidably engage to the exterior surface of tube 212. Two inverted U-shaped tubular fittings 222 are mounted on upper platform 108 and are supplied air under pressure from two pipes 224 that are in communication with an air blower assembly 226 mounted on vehicle V as illustrated in FIG. 12. Two pipes 228 extend downwardly from fittings 222 to two tubular swivels 230, with the lower ends of the swivels connected to the tube extensions 212 as shown in FIG. 13. In FIGS. 15 and 18 it will be seen that a tubular rectangular frame 232 is supported from the underside of lower plate 110b within shroud X and has spray nozzles 234 extending outwardly therefrom. Circular tubes 236 are supported from lower plate 110b and extend around Kellys 164 and support nozzles 238. Liquid under pressure is supplied to tubular frame 232 by a pipe 240 and to circular tubes 236 by a pipe 242. The liquid supplied to tubular frame 232 and circular tubes may be water to not only form sprays to scrub gases from the air in shroud X, but also to wash toxic material from Kellys as the detoxification of impoundment Y proceeds. Toxic gases that arise during the detoxificaton of impoundment Y are prevented from escaping upwardly around Kellys 164 by tubular bellows 244 that envelop the Kellys. The lower end of the bellows 244 are secured to lugs 160 by conventional means and the upper ends of the bellows to the lower surface of intermediate platform 112. Prior to using the apparatus U it is desirable that an underground radar scan be made of the hazardous waste impoundment to locate buried drums, tanks, barrels, and the like that may contain extremely dangerous materials. Suitable precautions must be taken when detoxifying the portions of the impoundment Y adjacent thereto. After obtaining the above information, as well as an analysis of a sample of the hazardous waste impoundment Y to obtain the composition thereof, the apparatus U is moved to a first station as shown in FIG. 25 adjacent the impoundment and the frame W moved to dispose the shroud X in sealing contact with the upper surface of impoundment Y. The motors 134 are now caused to drive the members 146 with the rollers 158 exerting a rotational force on the ribs 174 to rotate Kellys 164 and the cutting blade Z. Motors 126 are now energized to drive belts 130 to move intermediate platform 112 downwardly to exert a downward force on Kellys 164. Rotation of Kelly 164 is accompanied by the concurrent rotation of tubes 176, 178, 212, and 200, and pressurized air may now be discharged downwardly there through from blower assembly 226 to exit through nozzle 202. Operation of the apparatus U results in the forming of a downwardly extending zone A-1 of particled hazardous waste impoundment material as shown in FIG. 26. If the detoxifying agent is a dry powdered material it is introduced into the air stream from blower assembly 226 to discharge from nozzle 202. As the forming of zone A-1 takes place a pressurized liquid is discharged from the nozzles 194 to assist cutting blade Z in forming zone A-1 and reducing the size of the particles. Discharge of liquid from nozzles 194 causes the formation of a layer of turbulent liquid and particles above the blade Z which acts as a vertically movable seal to minimize the upward flow of toxic gases in zone A-1 into the interior of shroud X, and toxic gases below the seal being detoxified by the detoxifying agent. Toxic gases that flow upwardly into the shroud X are scrubbed therefrom by a series of liquid spray from nozzles 234 and 238, prior to air from shroud X being discharged to the ambient atmosphere. The liquid serving as the scrubbing agent flows downwardly into zone A-1 and is detoxified therein. Toxic gases from zone A-1 are prevented from flowing upwardly around Kelly 164 to the ambient atmosphere, due to the portion of the Kelly above the lower platform being encased in the longitudinally movable bellows 224. After the detoxifying method has been performed at a first station the apparatus U is returned to its initial position and subsequently moved to a sequence of second stations where the above described method is repeated. Although the method has been described with the use of pressurized air to displace toxic gases from the particled material in zone A-1., steam may be used for this purpose. Use of steam is desirable when the hazardous waste contains substantial quantities of volatile organic components. The pair of concurrently rotating blades Z do not interfere with one another due to the gears 138 being in toothed engagement as shown n FIG. 20. The rotating blades Z particle the hazardous material in zone A-1 without the particled material being appreciably discharged upwardly therefrom. First, second and third alternate forms of blades Z-1, Z-2 and Z-3 are shown in FIGS. 22, 23 and 24, each of which includes a pair of oppositely disposed arms 244 secured to outer tube 196 and have arcuate cutting blades 244 extending downwardly therefrom to the outer tube. The third alternate form Z-3 includes a special cutting member 248 and teeth 250 secured to outer tube 196. Instead of using a chemical detoxifying agent, the apparatus U may be used to introduce microorganisms into the zone A-1 to detoxify the latter. The microorganisms are either those already present in the impoundment Y or microorganisms that have been genetically engineered to biodegrade the hazardous material. The introduction of the microorganisms is accompanied with a liquid nutrient therefor. An assembly is shown diagrammatically in FIG. 28 that permits the composition of the hazardous waste impoundment to be determined as the zones A-1 are formed and the amount of detoxifying agent necessary to treat the same being determined by a computer system. A frame 248 is supported from tubes 196 above cutting blades Z, which frame supports a jetting assembly 250. A liquid wetting reagent or dionized water from a storage tank 252 is fed by pump 254 through line 256 into flushing jet assembly 250 as shown in FIG. 28. The jetting assembly 256 erodes or displaces or washes the contaminated waste, causing the wash water to surround the sampling device and probes later to be described mounted on frame 244. The wash water containing waste contaminants can be sampled or be in contact with the probes at any preprogrammed depth of the zone A-1. The sampling device shown as 258 picks up the flushed water and removing such water through line 260, the water pick-up is achieved by vacuum pump 262 and routed to the receiving chambers of an ICP Spectrometer or such suitable equipment shown as 262, for the screening of such toxic elements as heavy metals; to a radiation detector or such suitable equipment shown as 264 for the screening of radioactive substances; to a reactivity and conductivity analyser shown as 266 for screening the sampled water for such properties; to a biological analyser or such suitable equipment shown as 268 to characterize the biological properties therein or to preprep such samples for traditional laboratory analysis. The pH and Oxidation Reduction Potential (ORP) probe shown as 270 signals the pH and ORP of the wash water and transmits such signals to the pH and ORP meter shown as 272. The temperature and moisture content probe shown as 272 transmits signals through cable 274 to temperature and moisture meter shown as 276. Gases or vapors that may be released from the subsurface contents during mixing and homogenization are collected in shroud X, such liberated gases or vapors are collected by sensor 278 mounted on the shroud X. Such gases are routed to the photoionization detector or similar equipment and are screened for a wide range of chemical organic compounds, volatiles, and explosive vapors. The photoionization detector or suitable similar equipment is shown as 280. Sensor 278 also directs gases and vapors from zone A-1 to Sulfur Dioxide and Hydrogen Sulfide Detector shown as 282 measuring the concentration levels of those elements. The data acquired from the ICP Spectrometer 262; The Radiation Detector 264; the Reactivity and conductivity analyzer 266; the Biological Analyzer 268; the pH and ORP meter 272; the Temperature and moisture meter 274; the Photoionization detector 280 and the Sulfur Dioxide and Hydrogen Sulfide Detector 282 are signaled to the data scan and interface analyzer and controller 284, then routed to the treatment menu programmer interface system 286 which determines the specific treatment parameter and treatment media dosage rate trigerring feeder shown as 288, for the programmed feeding of the treatment media from the pneumatic chemical tanks shown as 290, such treatment media may include chemical reagents, bacteria, bacteria nutrients and oxygen generating chemicals. The selected treatment media is then fed into the Kellys 164 through top thereof as shown by line 224, for the integration with the subsurface waste. During the treatment stages all data aquisition systems earlier described are used where specifically needed for the characterization of the chemically or biologically improved subsurface contents. Gases and emissions are released from shroud X to scrubber 294 through a line 296. The data acquisition and analyzer equipment are not limited to those described above, and equipment or analyzers similar in function or purpose may be incorporated, since contaminants present in hazardous waste sites are not typical but in general can be found to be highly variable and complex. During the scrubbing of liberated gases or vapors by scrubber 294, the released and scrubbed emissions are routed through bypass shown at 296 to the photoionization detector 280 or Sulfur Dioxide and Hydrogen Sulfide detector shown as 282 to determine released compliance, or by pass 296 may be connected to an emission analyzer and the results of such data signalled to the data scan shown as 284. The plasticity meter shown as 298 acquires placticity or density of the contents of zone A-1 from the alternating power-load of the Kelly drive motors 134. Such data characterizes the completion of the solidification of the subsurface contents if the preferred treatment of the waste contents require solidification thereof. The RPM meter shown as 300 acquires such data from the Kelly drive motors 134. The vertical travel depth and speed of blades Z is screened by scanner 302, acquiring such data from a vertical travel monitor shown at 304. All acquired data from the Plasticity Meter 298, RPM meter 300 and Vertical Travel devices 302 and 304 are signalled to the Treatment Program Interface System 286 for incorporation into the preferred treatment of the hazardous material in zone A-1. In addition to the detoxifications previously described, the apparatus U may be used to vitrify zone A-1 if the same is of a sandy or clay composition. Such vitrification is accomplished by the use of plasma torches 350 held in the lower ends of tubes 196 by supports 352 as shown in FIG. 29. After the zone A-1 has been particled by use of blades Z, the blades are moved upwardly therein and the material therebelow is subjected to plasma arcs to melt and subsequently cool to a vitrified, nonsoluble, rigid mass. In the event the hazardous waste in zone A-1 does not contain sufficient sand or clay to vitrify, sand, clay or other vitrifiable material is added thereto through Kellys 164 by an air stream during the forming of the hazardous waste into particles. The use and operation of the invention has been described previously in detail and need not be repeated. Persons skilled in the art will readily appreciate that various modifications can be made from the preferred embodiment thus the scope of protection is intended to be defined only by the limitations of the appended claims. |
claims | 1. A packaging device for the transport and/or storage of a radioactive medium generating flammable gases and/or explosives via radiolysis, the said device comprising at least one canister intended to contain the radioactive medium, the said canister defining an inner storage space accessible via an opening for filling of the medium, on which plug-forming means are mounted,characterized in that:the said device also comprises a structure forming a chamber, and where said canister is arranged outside of said structure forming the chamber, and said canister is removably mounted on an exterior of said structure,and means for placing in communication allowing a first fluid communication to be set up between the said inner storage space and the said chamber. 2. The packaging device according to claim 1, characterized in that:the canister further comprises a first orifice opening into the inner storage space,in that the said chamber-forming structure comprises a second orifice opening into the said chamber,and in that the said first and second orifices form the two opposite ends of the said first fluid communication. 3. The packaging device according to claim 2, characterized in that the said means for placing in communication comprise a first member mobile between an open position in which it sets up the said first fluid communication, and a second closed position in which it shuts the said second orifice, the said first mobile member being mounted on the said chamber-forming structure. 4. The packaging device according to claim 2, characterized in that:the canister further comprises a third opening into the inner storage space,in that the said chamber-forming structure comprises a fourth orifice opening into the said chamber,and in that the said means for placing in communication allow a second fluid communication to be set up between the said inner storage space and the said chamber, the said third and fourth orifices forming the two opposite ends of the said second fluid communication. 5. The packaging device according to claim 2, characterized in that:the canister further comprises a third orifice opening into the inner storage space,in that the said chamber-forming structure comprises a fourth orifice opening into the said chamber,in that the said means for placing in communication allow a second fluid communication to be set up between the said inner storage space and the said chamber, the said third and fourth orifices forming the two opposite ends of the said second fluid communication,and in that a fifth and a sixth orifice are provided in the chamber-forming structure, and communicate with each other via a connecting duct forming an integral part of the said means for placing in communication. 6. The packaging device according to claim 5, characterized in that the said means for placing in communication comprise a first mobile member and a second mobile member each mounted on the said structure forming a chamber and able to be moved between an open position and a closed position, the said first mobile member being designed so that:in open position, firstly it places in communication the said first orifice with the said fifth orifice, and secondly it sets up the said second fluid communication by placing in communication the said third orifice with the said fourth orifice;in closed position, firstly it ensures the placing in communication of the fourth orifice with the said fifth orifice, and secondly it prohibits the communication of each of the fourth and fifth orifices with the outside of the said chamber,and in that the said second mobile member is designed so that:in open position, firstly it places in communication the said sixth orifice with the outside of the said chamber, and secondly it places the said second orifice in communication with the outside of the said chamber;in closed position, firstly it ensures the placing in communication of the sixth orifice with the said second orifice, and secondly it prohibits the communication of each of the second and sixth orifices with the outside of the said chamber. 7. The packaging device according to claim 3, characterized in that the said canister comprises an additional first mobile member, mobile between an open position in which it sets up the said first fluid communication, and a closed position in which it shuts the said first orifice, either one of the first mobile member and additional first mobile member being a leading member and the other a follower member of the actuating member, so that the movement of the actuating member from its closed position to its open position leads the said follower member also to move from its closed position to its open position, and conversely. 8. The packaging device according to claim 7, characterized in that the said actuating member also forms a mechanical connection member for the said canister on the chamber-forming structure. 9. The packaging device according to claim 8, characterized in that the said actuating member is designed so that its movement from its closed position to its open position, with the said canister bearing upon the first mobile member, ensures a mechanical connection of the canister, and so that the movement from its open position to its closed position ensures mechanical disconnection of this canister. 10. The packaging device according to claim 9, characterized in that the said actuating member forms a male or female part of a bayonet mechanical connection. 11. The packaging device according to claim 1, characterized in that it comprises a plurality of canisters each associated with means for placing in communication allowing a first fluid communication to be set up between its inner space and the said chamber. 12. The packaging device according to claim 11, wherein the means for placing in communication of all the canisters share the same second orifice,and a second mobile member is designed so that:in open position, firstly it places in communication each of sixth orifices with the outside of the said chamber, and secondly it places the said single second orifice in communication with the outside of the said chamber;in closed position, firstly it ensures the placing in communication of each of the sixth orifices with the said single second orifice, and secondly it prohibits the communication of the single second orifice and of each of the sixth orifices with the outside of the said chamber. 13. An assembly comprising the said packaging device according to claim 1, each canister housing in its inner storage space a given volume of radioactive medium, defining a level forming a horizontal boundary line with a gaseous headspace completing this inner storage space, the said means for placing in communication associated with the said canister having a first orifice opening into the said inner storage space, and arranged so that at all times it is in communication with the gaseous headspace, irrespective of the spatial orientation of the said canister integrating the said given volume of medium. 14. The assembly according to claim 13, characterized in that the said first orifice is provided at least in part in a duct projecting inside the said inner storage space. 15. The assembly according to claim 14, characterized in that the said first orifice opens in the vicinity of a baric centre of the said inner storage space. 16. A package for the transport and/or storage of a radioactive medium, characterized in that it comprises an overpack forming a cavity inside which an assembly according to claim 13. 17. Container for the transport and/or storage of a radioactive medium, characterized in that it comprises an overpack forming a cavity inside which a packaging device according to claim 1 is housed. 18. A method for packaging a radioactive medium in a device according to claim 1, wherein:the radioactive medium is introduced into the inner storage space of the canister;the canister is sealed using plug-forming means; andthe said first fluid communication is set up between the said inner storage space and the said chamber. |
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039430375 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Turning now to FIGS. 1 and 2, in a round reactor building 1 having a generally circularly extending vertical wall 1a, there is eccentrically situated a reactor pressure vessel 2, the upper terminal portion of which projects upwardly into a reactor pool 7. The latter is surrounded in a semicircle by a fuel element storage pool 3 which is of arcuate outline and which is filled with water to a predetermined level. Adjacent the two ends of the semicircular storage pool and in alignment therewith, there are provided openings 12 and 13 which are needed during the transport and insertion of new fuel elements. A connection between the reactor pool 7 and the fuel element storage pool 3 is provided by a channel 4 controlled by a watertight gate. In the fuel element storage pool 3, directly across from the channel 4, there are situated box-stripping machines 5 and, according to a further development of the invention, laterally of the box-stripping machines 5, there are provided tool storing devices 6 for the handling rods and the like. If now a fuel element exchange operation is to be performed, first the pressure vessel 2 is opened by removing the pressure vessel lid 19 and the cover 20. The reactor pool 7, which is situated above the reactor pressure vessel 2, may then be flooded and the gate of the channel 4 may be opened. After these preparatory steps, a main fuel element exchange gantry 8 is put to work. The gantry 8 has a bridge 8a, one end of which is supported for rotation at 14a by a vertical column 14 disposed substantially in the center of the reactor building 1. The bridge 8a is thus horizontally rotatable about a vertical axis 14b which passes through one end of the bridge 8a and which extends in the center of the building 1. The other end of the gantry bridge 8a is supported on a rail 9 which is situated above the level of an operating platform 10 (approximately at a vertical distance of 5 meters therefrom), and extends in an arc along the wall 1a of the reactor building 1. By means of a non-illustrated grasping device arranged on the hoist trolley of the main fuel element exchange gantry 8, a fuel element situated in the reactor pressure vessel 2 may be grasped, withdrawn from the pressure vessel 2 and carried through the channel 4 to one of the box-stripping machines 5. The arrangement of the rail 9 of the main fuel element exchange gantry 8 above the level of the operating platform 10 makes possible the provision of a second, more simple, fuel element exchange gantry 11 also having a fuel element grasping device mounted thereon. The gantry 11 operates simultaneously with the main fuel element exchange gantry 8 and transports the fuel elements from the storage pool 3 to the box-stripping machines 5 and conversely. Further, by means of the fuel element exchange gantry 11, the fuel elements and the fuel element storage racks 18 may be rearranged in the storage pool 3. The fuel element exchange gantry 11 has a cantilever bridge 11a and is supported and travels on two vertically spaced, parallel extending rails 21 and 22, both disposed beneath the rail 9 for the gantry 8. The support structure for the rail 9 may simultaneously serve -- as it is seen in FIG. 1 -- as the support for the upper rail 21 of the gantry 11. As it is also observable in FIG. 1, the gantry 11 may pass freely under the gantry 8. According to another embodiment of the invention, as illustrated in FIG. 3, there is provided a central column 17 which is a modification of the column 14 of the FIG. 1 embodiment. The column 17, in addition to supporting the gantry 8, also supports the inner rail of a building gantry 16 situated above the gantry 8 and further serves as a support member for a building roof 15. The box-stripping machines 5 are not illustrated in detail, because the usual box-stripping machines may be used. As already mentioned the fuel elements (fuel-rod assembly) are disposed in sheath (boxes). The box-stripping machines 5 serve to pull the fuel-rod assembly out of its box. After being pulled out of its box the fuel-rod assembly mas be transported directly to the fuel element storage pool 3 by the gantry 11. On the way back to the box-stripping machines the gantry 11 transports a new fuel element (fuel-rod assembly) which then is inserted in the box by the box stripping machine 5. Also the tool storing devices 6 are not illustrated in detail because the shape of the devices is determined by the tools. These tools may be for example; handling rods, screw-gripper, underwater-screwdrivers, television equipment and so on. It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptations, and the same are intended to be comprehended within the meaning and range of equivalents of the appended claims. |
054266793 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT In FIG. 1, a strainer housing of basically cylindrical form is generally designated by reference numeral 1 and comprises a tube-formed strainer wall 2 which has a multitude of small holes or apertures, e.g. in the shape of perforated holes in a sheet-metal. At one of its ends, in this case the upper one, the strainer housing is closed by a gable plate or wall 3. At the opposite end, the strainer housing is connected to a tube-formed suction conduit 4 which in its turn is connected to a suction pump (not shown). The latter is normally located outside the reactor containment, and water can be sucked in from a pool of water through strainer wall 2, and then be fed to the emergency cooling system. On the outside of strainer wall 2 are provided a number of wings, in this case four equidistantly separated wings 5a, 5b, 5c and 5d. As may be clearly seen in FIG. 1, these wings 5 extend the whole way along the axial length of strainer wall 2. According to the present invention a supple or flexible shield means unit is provided within strainer housing 1, the shield means unit being generally designated by reference numeral 6. It is shown in an unfolded or untightened condition in FIG. 2. In practice, the shield means unit 6 may advantageously be made of a piece of rubber or plastic of suitable stiffness and thickness (for instance 3 to 6 mm). As may be seen in FIG. 2, shield means unit 6 has four (in the present case) distinctly separate edges, viz. a first edge 7, second and third edges 8,9 and a fourth edge 10. In the shown example, the shield means unit is slightly sector-shaped in its unfolded, untightened state, since the two longitudinal edges 8,9 diverge from each other, long side edge 10 attains a longer length than opposite short side edge 7. More precisely, long side edge 10 has substantially the same length as half the circumference of the cylindrical strainer wall 2 of strainer housing 1, while the length of short side edge 7 substantially corresponds to the diameter of cylinder tube 2. In the area of the longer curved side edge 10, a mechanical spring 11 is inserted into the rubber or plastic cloth, the spring being flexible in a direction across the plane of the cloth. Spring 11 may be in different forms. However, a spring with the shape of wave or sinusoidal-formed spring wire is preferred, which is capable of being shortened in connection with bending across the plane of the spring and the cloth. When the thus formed rubber cloth 6, which serves as a shield means unit, is mounted within strainer housing 1, three of the four edges of the cloth are tightened, viz. edges 7, 8, 9, while the fourth edge 10 remains freely movable. The tightening of edges 7, 8, 9 is schematically illustrated in FIG. 1 using two clamping strips 12 which are located on the lower side of gable plate 3, thus clamping short side edge 7. Corresponding clamping strips also clamp longitudinal side edges 8, 9, although this is not shown in FIG. 1, for the sake of clarity. In practice, the three fastened edges 7, 8, 9 may also be fastened in another way, e.g., between flanges on two semi-cylindrical housing halves together forming a cylindrical housing when joined and mounted. It will be evident that the fourth, free edge 10 of the rubber cloth may be bent between two opposition adjustment positions in which the edge in question abuts against the inside of strainer wall 2. Reference is now made to FIGS. 4 through 6 which illustrate the function of the invention. FIG. 5 shows how the upper edge 7 of rubber cloth 6 extends between wings 5a and 5c which are located on the outside of the strainer housing. Thus, longitudinal side edges 8,9 extend parallel to each other, immediately adjacent to the inner edges of wings 5a,5c. In this way, strainer wall 2 is divided into two separate, first and second part surfaces which are designated 2a and 2b, respectively. In FIGS. 4 and 5 rubber cloth 6 is shown as being adjusted in a first position or condition in which the free edges 10 is held in abutment or engagement against the inside of part surface 2b by spring 11, more particularly in the area of the lower edge of the part surface. In this condition, rubber cloth 6 interrupts the contact between strainer wall surface 2b and suction conduit 4. Therefore, when water is sucked into conduit 4 via strainer housing 1, it is accomplished solely via the first strainer wall surface 2a, as illustrated by the arrows in FIG. 4. As water is sucked through the strainer housing, fibres may build up on the outside of strainer wall surface 2a and form a semicircular mat, as shown in FIG. 5. When this mat has "grown" sufficiently thick, it will exert such a large resistance against the suction of water that an area of low-pressure will arise in the space between strainer wall surface 2a and rubber cloth 6 (i.e., a pressure lower than the surrounding water pressure). At a sufficiently low-pressure in said space, this suction effect becomes so strong that spring 11 will no longer be able to retain its initial position. This means that the spring will revert, over its own inherent spring action, to a diametrically opposite position, as shown in FIG. 6, i.e., into a second working position in which the lower edge of the cloth abuts against the inside of the first strainer wall surface 2a while interrupting the communication between this strainer wall surface and suction conduit 4. This means that the connection between strainer wall surface 2b and suction conduit 4 is opened so that water can be sucked or drawn in via strainer wall surface 2b. When the spring and the rubber cloth converts over from the first to the second working position, the previous suction flow in the space inside strainer wall 2a ceases, whereby the fibre mat on the outside expands and detaches from the wall. Moreover, a light pressure pulse arises in said space, contributing to the removal of the mat, as indicated to the left in FIG. 6. This removal is further enhanced by the fact that wing 5b divides the curved mat 13 into two quarter-circular halves. When a sufficiently thick fibre mat is also built up on strainer wall surface 2b, the above described phenomenon is repeated, but in the opposition direction. The advantages of the invention will thus be evident. Independently of which of its two opposite working positions the rubber cloth occupies, one of the two part surfaces 2a,2b of the strainer wall is always kept open for the suction of water. Furthermore, the rubber cloth or shield means unit, which in practice functions as switch-over valve, also contributes to an automatic removal of a fibre mat by giving rise to at least a light pressure gust in the space inside the part surface being covered by fibres, in connection with the switching-over from one working position to the other. Besides, the valve function obtained by the rubber cloth works automatically, i.e., without any outer control, in that the cloth is bent from one position to the other as soon as a sufficiently thick fibre mat has built-up on the outside of the strainer wall surface in question. It should also be pointed out that the simple and inexpensive rubber cloth is included into the strainer housing in its entirety, implying that external washwater conduits and pumps (together with the required control equipment) may be entirely eliminated. In FIG. 7, an alternative embodiment is shown, according to which the strainer device comprises two separate strainer housings 1' and 1" respectively, which are connected to a common suction conduit 4' via branch conduits 14,15. In this case, a rubber cloth or another flexible shield means unit 6' is mounted adjacent to the fork point between the main suction conduit 4' and the branch conduits 14,15. In the working position shown in FIG. 1, rubber cloth 6' interrupts the connection between strainer housing 1' and suction conduit 4', at the same time as the connection between strainer housing 1" and conduit 4' is kept open. In this condition, the suction of water takes place via strainer housing 1", while strainer housing 1' is inactive. However, as soon as a fibre mat of sufficient thickness has built-up on the outside of strainer housing 1", a low pressure area eventually results in a switch-over of rubber cloth 6' to the working position indicated by the dash-dotted lines, at which point the connection between strainer housing 1' and suction conduit 4' is opened. At the switching-over of rubber cloth 6' a light pressure pulse is also created in branch conduit 15, this pressure pulse being sufficient to blow away the fibre mat on the outside of strainer housing 1". Similarly, to part surfaces 2a,2b in the arrangement according to FIG. 1, the two strainer housing 1' and 1" will also work alternately between the two working positions. Alternate Embodiments of the Invention It is evident that the invention is not restricted solely to the specific embodiments described above and shown in the drawings. Thus, other means than just a rubber cloth may be used for forming the flexible shield means unit. Thus, it is feasible to use thin, plied sheetings or other plates of a flexible nature. Further, the geometrical form of the shield means unit may vary within wide limits. Thus, the two longitudinal edges of the shield means unit may transpose into a third, clampable portion or into each other via rounded material portions instead of via sharp corners, as exemplified in the drawings. It should further be pointed out that the supple edge portion with the spring does not necessarily have to be in tight abutment against the inside of the strainer wall. Thus, a certain leakage flow between this edge portion and the strainer wall is acceptable. It should also be pointed out that the shield means unit may be mounted into housings or tubes with another cross-sectional form other than circular, e.g., oval or polygonal. |
041982711 | abstract | The reactor core is carried from the base region of the primary vessel by a strongback. An annular foot is attached to the inside surface of the leak jacket leaving a clearance of 6 mm between the foot and the outer surface of the base region of the primary vessel. In the event of deflection of the primary vessel, downward displacement of the core relative to the control rods is limited thereby avoiding serious reactivity instability. |
summary | ||
047120142 | summary | BACKGROUND OF THE INVENTION The invention relates to a radiation lamp unit with a housing containing a number of highly-polished concave reflectors, with light-orange radiation lamps arranged in their focal point areas, and at least one UV lamp which is secured on a base of a lamp unit. Such radiation lamp units, developed and constructed by the applicant, are well known. It has been determined, however, that previously known radiation lamp units are not completely satisfactory for the treatment of a large body area, e.g. Psoriasis covering a large part of the body, simply because uniform treatment is not possible when the distance between the radiation sources and the affected parts of the body increases. It is particularly evident, even if the UV lamps are swivelled to and fro, if one UV-B and one UV-C lamp are fitted in the lamp base, since in one case one lamp is closer to the part of the body to be treated than the other. SUMMARY OF THE INVENTION To solve this problem, the invention uses two UV lamp units instead of one, preferrably arranged symmetrical to the center axis, and also in front of the focal point area of a light-orange radiation lamp in each case. It may also be favorable in certain cases to arrange each of the UV lamp units between two concave reflectors. The lamp units preferably have two UV lamps secured to a common base. |
abstract | A light delivery device (10) having a light source (12) and a variable aperture unit (18) is temporarily connected by a light guide (24; 24, 26) to a radiometer (38) for detecting irradiance of the delivered light. The light delivery device has a memory (30) for storing irradiance levels. The light delivery device is calibrated by adjusting the aperture to each of a series of predetermined settings, obtaining from the radiometer a corresponding series of delivered light irradiance levels measured thereby, storing the irradiance levels and aperture settings in memory, and applying a best fit algorithm to the irradiance measurements and aperture settings. Thereafter, a desired irradiance level can be set by selecting the best fit aperture setting. Output intensity levels may be measured at the same time as the irradiance levels and used to compensate for light source degradation when setting a desired irradiance level. |
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053176065 | abstract | This invention relates to an automation system for nuclear power plants. This system comprises operation plan making means, supervisory control means, and control means for controlling controlled objects. The operation plan making means makes a new operation plan to restore a normal operating condition which existed before the plant condition deviated from the normal operation range. The supervisory control means outputs control commands according to an operation plan under abnormal condition when the plant operating condition deviates from the normal operation range. Furthermore, the supervisory control means outputs control commands according a new operation plan prepared by the operation plan making means after an operation according to an operation plan under abnormal condition has been finished. The control means for controlling controlled objects controls the controlled objects according to control commands output from the supervisory control means. |
claims | 1. A radiation monitor comprising:a gas flow path in which a sampling solenoid valve, a purge solenoid valve, a pump, and an exhaust solenoid valve, each being operated by an AC electric power, are provided, and a sampling gas suctioned from a sampling point is circulated;a flow sensor installed in said gas flow path;a pressure sensor installed in said gas flow path;an AC control section configured to supply the AC electric power to said sampling solenoid valve, said purge solenoid valve, said pump, and said exhaust solenoid valve, and when a flow path abnormality signal is received, the AC control section is configured to control said sampling solenoid valve, said purge solenoid valve, and said exhaust solenoid valve to be closed and is configured to control said pump to be stopped;a DC control section operated by a DC electric power in which the AC electric power supplied from said AC control section is converted, the DC control section configured to output the flow path abnormality signal in the case where a measured value of said flow sensor or a measured value of said pressure sensor is lower than a set value;a detection unit which detects radiation to be released from the sampling gas suctioned to said gas flow path, and configured to output a detection signal; anda measurement unit which measures the detection signal outputted from said detection unit, and configured to output radioactivity concentration,wherein said AC control section is configured to output an AC power source instantaneous power failure detection signal to said DC control section when a decrease in AC voltage is detected;said DC control section is configured to measure a duration time of the AC power source instantaneous power failure detection signal when the AC power source instantaneous power failure detection signal is received from said AC control section, and output an instantaneous power failure restart signal to said AC control section if the AC voltage is restored within a time shorter than the set value; andsaid AC control section is configured to perform switching control from close to open of said sampling solenoid valve, said purge solenoid valve, and said exhaust solenoid valve, and restart said pump after a constant time when the instantaneous power failure restart signal is received from said DC control section. 2. The radiation monitor according to claim 1,wherein said exhaust solenoid valve is installed on the lower stream side of said pump and in a common piping portion of said gas flow path. 3. The radiation monitor according to claim 1,wherein said exhaust solenoid valve is installed on the lower stream side of said pump and in a branch piping portion of said gas flow path. 4. The radiation monitor according to claim 2,wherein said sampling solenoid valve and said exhaust solenoid valve are opened in non-excitation; and said purge solenoid valve is closed in non-excitation. 5. The radiation monitor according to claim 3, whereinsaid sampling solenoid valve and said exhaust solenoid valve are opened in non-excitation; and said purge solenoid valve is closed in non-excitation. 6. The radiation monitor according to claim 2,wherein, in the case where a decrease time of the AC voltage is longer than a predetermined time, said AC control section is configured to restart the pump after a set time after the switching control from close to open of said sampling solenoid valve, said purge solenoid valve, and said exhaust solenoid valve is performed. 7. The radiation monitor according to claim 3,wherein, in the case where a decrease time of the AC voltage is longer than a predetermined time, said AC control section is configured to restart the pump after a set time after the switching control from close to open of said sampling solenoid valve, said purge solenoid valve, and said exhaust solenoid valve is performed. 8. The radiation monitor according to claim 2,wherein said sampling solenoid valve and said purge solenoid valve are mounted in said gas flow path in a state where the direction of an arrow showing a mounting direction is pointed to the upper stream side. 9. The radiation monitor according to claim 3,wherein said exhaust solenoid valve is mounted in said gas flow path in a state where the direction of an arrow showing a mounting direction is pointed to the upper stream side. |
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051704188 | abstract | An X-ray exposure method and apparatus wherein an exposure chamber coupled with a synchrotron radiation device through a beam line receives synchrotron radiation through a window material provided on the beam line to execute an exposure process, a pressure sensor detects pressure in the exposure chamber, a cutoff valve is provided in a portion of the beam line between the window material and the synchrotron radiation device, and a bypass has a communication valve for communicating a portion of the beam line between the window material and the cutoff valve with a portion between the window material and the exposure chamber. In addition, a vacuum evacuating device effects evacuation of a portion of the beam line between the window material and the cutoff vlave, a pump valve is disposed in a conduit, coupling the beam line with the vacuum evacuating device, and a controller responds to a pressure detected by the pressure sensor so that, when the detected pressure represents a steady state lower than a predetermined pressure, the controller operates to open the cutoff valve and the pump valve and to close the communication valve and, when the detected pressure is higher than the predetermined pressure, the controller operates to close the cutoff valve and the pump valve and thereafter to open the communication valve. |
description | The present application is related to and/or claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Priority Applications”), if any, listed below (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Priority Application(s)). In addition, the present application is related to the “Related Applications,” if any, listed below. For purposes of the USPTO extra-statutory requirements, the present application constitutes a divisional of U.S. Pat. No. 8,529,713, entitled SYSTEM AND METHOD FOR ANNEALING NUCLEAR FISSION REACTOR MATERIALS, naming Charles E. Ahlfeld, John Rogers Gilleland, Roderick A. Hyde, David G. McAlees, Jon David McWhirter, Ashok Odedra, Clarence T. Tegreene, Joshua C. Walter, Kevan D. Weaver, Charles Whitmer, Lowell L. Wood, Jr., and George B. Zimmerman as inventors, granted 10 Sep. 2013, which is currently co-pending or is an application of which a currently co-pending application is entitled to the benefit of the filing date. U.S. Pat. No. 8,721,810, entitled SYSTEM AND METHOD FOR ANNEALING NUCLEAR FISSION REACTOR MATERIALS, naming Charles E. Ahlfeld, John Rogers Gilleland, Roderick A. Hyde, David G. McAlees, Jon David McWhirter, Ashok Odedra, Clarence T. Tegreene, Joshua C. Walter, Kevan D. Weaver, Charles Whitmer, Lowell L. Wood, Jr., and George B. Zimmerman as inventors, granted 13 May 2014, is related to the present application. U.S. Pat. No. 8,784,726, entitled SYSTEM AND METHOD FOR ANNEALING NUCLEAR FISSION REACTOR MATERIALS, naming Charles E. Ahlfeld, John Rogers Gilleland, Roderick A. Hyde, David G. McAlees, Jon David McWhirter, Ashok Odedra, Clarence T. Tegreene, Joshua C. Walter, Kevan D. Weaver, Charles Whitmer, Lowell L. Wood, Jr., and George B. Zimmerman as inventors, granted 22 Jul. 2014, is related to the present application. U.S. Pat. No. 9,011,613, entitled SYSTEM AND METHOD FOR ANNEALING NUCLEAR FISSION REACTOR MATERIALS, naming Charles E. Ahlfeld, John Rogers Gilleland, Roderick A. Hyde, David G. McAlees, Jon David McWhirter, Ashok Odedra, Clarence T. Tegreene, Joshua C. Walter, Kevan D. Weaver, Charles Whitmer, Lowell L. Wood, Jr., and George B. Zimmerman as inventors, granted 21 Apr. 2015, is related to the present application. The United States Patent Office (USPTO) has published a notice to the effect that the USPTO's computer programs require that patent applicants reference both a serial number and indicate whether an application is a continuation, continuation-in-part, or divisional of a parent application. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette Mar. 18, 2003. The USPTO further has provided forms for the Application Data Sheet which allow automatic loading of bibliographic data but which require identification of each application as a continuation, continuation-in-part, or divisional of a parent application. The present Applicant Entity (hereinafter “Applicant”) has provided above a specific reference to the application(s) from which priority is being claimed as recited by statute. Applicant understands that the statute is unambiguous in its specific reference language and does not require either a serial number or any characterization, such as “continuation” or “continuation-in-part,” for claiming priority to U.S. patent applications. Notwithstanding the foregoing, Applicant understands that the USPTO's computer programs have certain data entry requirements, and hence Applicant has provided designation(s) of a relationship between the present application and its parent application(s) as set forth above and in any ADS filed in this application, but expressly points out that such designation(s) are not to be construed in any way as any type of commentary and/or admission as to whether or not the present application contains any new matter in addition to the matter of its parent application(s). If the listings of applications provided above are inconsistent with the listings provided via an ADS, it is the intent of the Applicant to claim priority to each application that appears in the Priority Applications section of the ADS and to each application that appears in the Priority Applications section of this application. All subject matter of the Priority Applications and the Related Applications and of any and all parent, grandparent, great-grandparent, etc. applications of the Priority Applications and the Related Applications, including any priority claims, is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. The present application relates to nuclear fission materials, and systems, methods, apparatuses, and applications related thereto. If an Application Data Sheet (ADS) has been filed on the filing date of this application, it is incorporated by reference herein. Any applications claimed on the ADS for priority under 35 U.S.C. §§119, 120, 121, or 365(c), and any and all parent, grandparent, great-grandparent, etc. applications of such applications, are also incorporated by reference, including any priority claims made in those applications and any material by reference, to the extent such subject matter is not inconsistent herewith. Illustrative embodiments provide systems, methods, apparatuses, and applications related to annealing nuclear fission reactor materials. The foregoing summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. First, an overview will be set forth regarding illustrative embodiments, non-limiting examples of components that may be annealed, and annealing effects on components of nuclear fission reactors. Next, illustrative methods will be explained. Then, illustrative apparatuses will be explained. Overview Illustrative embodiments provide systems, methods, apparatuses, and applications related to annealing nuclear fission reactor materials. In some embodiments, illustrative methods are provided for annealing nuclear fission reactor materials, such as without limitation a nuclear fission reactor core or fuel assembly or components thereof. For example, referring to FIG. 1A an illustrative method 100 is provided for annealing at least a portion of at least one metallic component of a nuclear fission fuel assembly of a nuclear fission reactor. Referring to FIG. 2A, an illustrative method 200 is provided for annealing at least a portion of at least one component of a reactor core of a nuclear fission reactor. Referring to FIG. 3A, an illustrative method 300 is provided for treating at least a portion of at least one component of a reactor core of a nuclear fission reactor. Referring to FIG. 4A, a method 400 is provided for producing an annealing effect. Referring now to FIG. 5A, a method 500 is provided annealing at least a portion of at least one component of a nuclear fission reactor core. Details will be set forth further below. The illustrative methods, systems, and apparatuses described herein may be used for annealing any irradiated component of a core of any type of nuclear fission reactor as desired and without limitation. A brief overview of illustrative reactor core components that may be annealed will now be set forth by way of non-limiting examples. It will be understood that the following examples of components that may be annealed are described by way of illustration only and not limitation. For example, components of a reactor core assembly of a pressurized water reactor may be annealed. Referring now to FIG. 6A, an illustrative pressurized water reactor 600, given by way of non-limiting example, includes a reactor pressure vessel 602 that contains a reactor core assembly 604 in which nuclear fission occurs within the thermal spectrum. Each primary reactor coolant loop 606 includes its own heat exchanger 608, such as a steam generator, and reactor coolant pump 610. A pressurizer 612 is connected to one of the primary reactor coolant loops 610 and controls reactor coolant pressure, typically through use of heaters (not shown) that control temperature of the reactor coolant in the pressurizer 612. The pressurizer 612 helps maintain pressure of the reactor coolant sufficiently high, such as around 2250 psig or so, to help prevent formation of steam in the primary system. The reactor coolant pumps 610 pump reactor coolant through cold legs 614 into the reactor pressure vessel 602. The reactor coolant is heated by heat from nuclear fission occurring in the reactor core assembly 604. Reactor coolant exits the reactor pressure vessel 602 through hot legs 616 and enters the steam generators 608. Heat is transferred from the reactor coolant to secondary coolant in U-tubes (not shown) in the steam generators, thereby generating steam that can be used to drive turbines (not shown), such as electrical turbine generators, engines, or the like. Referring additionally to FIG. 6B, a basic unit of the reactor core assembly 604 is a nuclear fission fuel element 618, such as a fuel rod or fuel pin. Nuclear fission fuel material, such as uranium dioxide, is pressed into cylindrical pellets 620 that are sintered, ground to desired dimensions, and sealed, such as by welding shut, in cladding 622, such as an alloy of zirconium (like zircalloy). Flow of reactor coolant, at typical operating temperatures of around 600° F., through the fuel assemblies 626 helps maintain temperature of the zircalloy cladding 622 nominally below around 700° F. An annular space 624 typically is provided between the pellets 620 and the cladding 622. Referring additionally to FIG. 6C, nuclear fission fuel elements 618 are assembled into a fuel assembly 626. In a typical fuel assembly 626, the nuclear fission fuel elements 618 are assembled into a square array that is held together by spring clip grid assemblies 628 and by nozzles 630 and 632 at the top and bottom, respectively, of the nuclear fission fuel assembly 626. An open structure of the nuclear fission fuel assembly 626 defines reactor coolant channels that permit flow of reactor coolant (vertically and horizontally). The nuclear fission fuel assembly 626 may also include provision for passage of one or more control rods 634 that contain neutron absorbing material. Referring additionally to FIG. 6D, the reactor core assembly 604 includes several fuel assemblies 626. The reactor core assembly 604 also includes cooling components, such as baffles 636 and the reactor coolant channels that direct reactor coolant to, through, and from the fuel assemblies 626. The reactor core assembly 604 also includes structural members that form the fuel assemblies 626 into the reactor core assembly 604, such as core support columns 638, an upper core plate 640, a lower core plate 642, a core barrel 644, and the like. By way of further examples, components of a reactor core assembly of a fast breeder reactor may be annealed. Referring now to FIG. 7A, a liquid metal fast breeder reactor 700 uses a liquid metallic reactor coolant, such as sodium, lead, lead-bismuth, or the like, to cool a reactor core assembly 702. The reactor coolant is pumped by a reactor coolant pump 704 in a primary reactor coolant loop 706. A heat exchanger 708 transfers heat from the reactor coolant to intermediate loop coolant (which may be the same fluid as the reactor coolant in the primary reactor coolant loop 706) that is pumped by an intermediate coolant pump 710 in an intermediate coolant loop 712. A heat exchanger 714, such as a steam generator, generates steam that can be used to drive one or more turbines 716, such as electrical turbine generators, engines, or the like. A condenser 718 condenses steam that is exhausted by the turbine 716. Condensate from the condenser 718 is pumped by a feedwater pump 720 to the heat exchanger 714. Referring additionally to FIG. 7B, a basic unit of the reactor core assembly 702 is a nuclear fission fuel element 722, such as a fuel rod or fuel pin. A portion of the nuclear fission fuel element 722 includes fissile material 724, such as 239Pu, 233U, or 235U. Because the liquid metal fast breeder reactor 700 is a breeder reactor, the reactor core assembly 702 typically produces as much or more fissile material than it consumes. To that end, the nuclear fission fuel element 722 also includes portions of fertile material 726, such as 238U or 232Th. In one approach, the fissile material 724 and the fertile material 726 typically are pressed into oxide pellets that are sealed, such as by welding shut, in cladding 728, such as stainless steel. Referring additionally to FIG. 7C, nuclear fission fuel elements 722 are assembled into a fuel assembly 730. In a typical fuel assembly 730, the nuclear fission fuel elements 722 are assembled into an assembly that is held together by a handling fixture 732 and by a grid plate 734. An open structure of the nuclear fission fuel assembly 730 defines reactor coolant channels that permit flow of reactor coolant. Referring additionally to FIG. 7D, the reactor core assembly 702 includes several fuel assemblies 730. The reactor core assembly 702 also includes cooling components, such as throttling inserts that can throttle reactor coolant to the fuel assemblies 730. The reactor core assembly 702 also includes structural members that form the fuel assemblies 730 into the reactor core assembly 702, such as an upper core support plate 738, a lower core support plate 740, a core barrel 742, and the like. The reactor core assembly 702 is contained within a reactor pressure vessel 744. In some other arrangements and referring additionally to FIG. 7E, a pool-type liquid metal fast breeder reactor 700A uses a pool of liquid metallic reactor coolant, such as sodium, lead, lead-bismuth, or the like, in a reactor pressure vessel 744A to cool a reactor core assembly 702A. The reactor pressure vessel 744A contains the pool of reactor coolant, the reactor core assembly 702A, the reactor coolant pump 704, and the heat exchanger 708. Another example of a fast breeder reactor is a gas cooled fast breeder reactor. Referring now to FIG. 7F, a gas cooled fast breeder reactor 750 includes a reactor pressure vessel 744B that contains a reactor core assembly 702B that is cooled by a gaseous reactor coolant, such as helium, that is circulated by a gaseous coolant circulator 752. The gaseous reactor coolant is circulated through the reactor core assembly 702B and is heated, and heat is transferred from the gaseous reactor coolant in a heat exchanger 754, such as a steam generator. Referring additionally to FIG. 7G, the reactor core assembly 702B includes nuclear fission fuel elements that are assembled into fuel assemblies 730B by structural components, such as a grid plate 734B and a grid support structure. The nuclear fission fuel elements and the fuel assemblies 730B are generally similar to the nuclear fission fuel elements 722 (FIGS. 7B and 7C) and the fuel assemblies 730 (FIG. 7C), with the difference that the nuclear fission fuel elements of the gas cooled fast breeder reactor 750 have surfaces that are roughened to provide increased surface area for heat transfer to the gaseous reactor coolant (that is, a thermally conductive member). Referring now to FIGS. 7A-7G, in some arrangements the liquid metal fast breeder reactors 700 (FIGS. 7A-7D) and 700A (FIG. 7E) and the gas cooled fast breeder reactor 750 (FIGS. 7F-7G) may entail conventional nucleonics that involve reprocessing of breeder blankets. In some other arrangements, liquid metal fast breeder reactors and gas cooled fast breeder reactors may entail nucleonics in which a nuclear fission deflagration wave is initiated and propagated. Initiation and propagation of a nuclear fission deflagration wave is discussed in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the contents of which are hereby incorporated by reference. Reactor materials, such as without limitation components, like metallic components, of reactor cores discussed in the illustrative non-limiting examples set forth above, can experience exposure to neutrons with energy sufficient to create degradation, such as defects, in the material on the atomic and molecular level. Radiation damage to structural materials (measured in dislocations per atom (dpa)) is primarily reflective of exposure to neutrons with energies greater than 1 MeV. Damage from neutron exposure tends to cause radiation hardening, such that the ductile-to-brittle transition temperature of the material increases. Moreover, in a nuclear fission deflagration wave fast breeder reactor, reactor core materials may experience a high level of fluence due to exposure to high energy (that is, fast spectrum) neutrons over a prolonged time (due to slow propagation velocity of the nuclear fission deflagration wave). For some classes of structural materials (such as ferritic/martensitic steels), it is known that some radiation damage can be removed by heating the material to greater than around 40% or so of its melting point and holding the material at that temperature for a pre-determined amount of time—that is, annealing the material. This removal of radiation damage results from relieving stress by, primarily, thermally inducing migration of crystalline defects to grain boundaries. When these defects are in the form of dislocations, these dislocation points act as localized stress risers within the crystal. Increasing the temperature of the material increases the mobility of the dislocations, thereby enabling the dislocation to migrate to a grain boundary where the stress is relieved. Subsequent cooling (e.g., quenching) for a predetermined amount of time followed by an increase in temperature can temper the material, thereby “locking in” its desired metallurgical qualities. Counter to this effect is creep (that is, physical geometry change of the bulk material due to applied stresses such as fuel element internal pressure from fission products). The rate of creep increases with increasing temperature for a given stress. The creep rate in conjunction with internal vs. external pressures on the fuel element and/or fuel assemblies may limit annealing temperatures and annealing times. The illustrative methods, systems, and apparatuses described herein can be used to treat or anneal components of reactor core assemblies or fuel assemblies, as desired for a particular application. To that end, it will be appreciated that the discussion set forth above regarding components of reactor core assemblies and components of fuel assemblies (that may be annealed by illustrative embodiments disclosed herein) is provided by way of non-limiting examples. That is, the components of reactor core assemblies and the components of fuel assemblies that may be treated or annealed by illustrative embodiments disclosed herein is not limited to those components of reactor core assemblies and components of fuel assemblies discussed above. To that end, any irradiated component of any reactor core assembly or any fuel assembly can be treated or annealed by illustrative embodiments disclosed herein. Illustrative Methods Now that an overview of illustrative methods and non-limiting examples of illustrative components that may be treated or annealed has been set forth, illustrative details of methods will now be discussed. Following are a series of flowcharts depicting implementations of processes. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an overall “big picture” viewpoint and thereafter the following flowcharts present alternate implementations and/or expansions of the “big picture” flowcharts as either sub-steps or additional steps building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an overall view and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular design paradigms. Referring now to FIG. 1A, the illustrative method 100 for annealing at least a portion of at least one metallic component of a nuclear fission fuel assembly of a nuclear fission reactor begins at a block 102. At a block 104 an annealing temperature range for at least a portion of at least one metallic component of a nuclear fission fuel assembly of a nuclear fission reactor is determined. At a block 106 at least the portion of the at least one metallic component of the nuclear fission fuel assembly is annealed within the annealing temperature range. The method 100 stops at a block 108. Illustrative details will be set forth below. It will be appreciated that any metallic component of any fuel assembly can be annealed by the method 100. For example, in some embodiments the at least one metallic component can include cladding, a cooling component, a structural member, a thermally conductive member, and/or nuclear fission fuel material. As discussed above, metals such as zircalloy and stainless steel also serve as the fuel element enclosure (that is, cladding). However, it will be appreciated that treatment by annealing as described herein can expand the types of materials that may be used for reactor core materials. To that end and given by way of non-limiting examples, metal from which the metallic component is made can include without limitation steel, oxide dispersion strengthened (ODS) steels, austenitic steels (304, 316), ferritic/martensitic steels refractory metal, a refractory metal alloy, a non-ferrous metal, a non-ferrous metal alloy, and/or a superalloy (such as Inconels, Zircaloys, and/or Hastelloys). In some embodiments, the annealing temperature range determined at the block 104 may be greater than a predetermined operating temperature range of the nuclear fission fuel assembly. For example, some illustrative pressurized water reactor fuel assemblies may have an operating temperature range between cold leg temperature TC of around 550° F. and hot leg temperature TH of around 650° F. (at a nominal coolant pressure of around 2,250 psig); an illustrative loop type liquid metal fast breeder reactor fuel assembly may have an operating temperature range between TC of around 700° F. and TH of around 1000° F.; and an illustrative gas cooled fast breeder reactor fuel assembly may have an operating temperature range between TC of around 600° F. and TH of around 1000° F. However, as will be described below, the annealing temperature range may be greater than the predetermined operating temperature range of the nuclear fission fuel assembly. In some embodiments, the annealing temperature range may be determined based upon radiation exposure of the at least one metallic component of the nuclear fission fuel assembly. For example, annealing temperature range may be based upon factors such as energy of the neutron spectrum to which the metallic component has been exposed. For example, for a given exposure time (such as may be measured in effective full power hours), exposure to a fast neutron spectrum (like in a fast breeder reactor) may result in more radiation damage than would exposure of the metallic component to a thermal neutron spectrum (like in a pressurized water reactor). As another example, for exposure to a given neutron spectrum (such as a thermal neutron spectrum or a fast neutron spectrum), exposure for a longer time (such as may be measured in effective full power hours) may result in more radiation damage than would exposure of the metallic component exposure for a shorter time. In such a case, a higher annealing temperature range (for a given annealing processing time) may be entailed for the case of longer exposure to the given neutron spectrum than would be entailed for the case of shorter exposure to the given neutron spectrum. Moreover, in some cases a portion of some components, such as without limitation, a middle of a fuel assembly or fuel element, may have a radiation exposure history that is different from a radiation exposure history of another portion of the component, such as without limitation, an edge region of the nuclear fission fuel assembly or fuel element. In such a case, a radiation damage gradient may exist along the component. Thus, an annealing temperature range may be different for one portion of the component to be annealed than for other portions of the component. In some other embodiments, the annealing temperature range may be determined based upon an operating temperature history during which the radiation occurred. It will be appreciated that lower temperature regions of a fuel assembly may suffer more radiation damage effects than higher temperature regions in the same fuel assembly. Moreover, in some cases a portion of some components, such as without limitation a middle of a fuel assembly or fuel element, may have an operating temperature history that is different from an operating temperature history of another portion of the component, such as without limitation an edge region of the nuclear fission fuel assembly or fuel element. In such a case, a radiation damage gradient may exist along the component. Thus, an annealing temperature range may be different for one portion of the component to be annealed than for other portions of the component. In some other embodiments, the annealing temperature range may be determined based upon an annealing history of the component to be annealed. That is, in some embodiments historical data regarding annealing temperature of past annealing operations for a metallic component may be used to determine future annealing temperature ranges for the metallic component. In some other embodiments, the annealing temperature range may be determined based upon material properties of the at least one metallic component of the nuclear fission fuel assembly. For example, in some embodiments a minimum temperature of the annealing temperature range may be at least around thirty percent of a melting point of the at least one metallic component of the nuclear fission fuel assembly. In one of the non-limiting examples discussed above, for stainless steel with a melting point of around 2,732° F., such a minimum temperature of the annealing temperature range can be around 820° F. In another non-limiting example discussed above, for Zircaloy with a melting point of around 3,362° F., such a minimum temperature of the annealing temperature range can be around 1,009° F. As another example, in some other embodiments, an annealing temperature within the annealing temperature range may be around forty percent of a melting point of the at least one metallic component of the nuclear fission fuel assembly. In one of the non-limiting examples discussed above, for stainless steel with a melting point of around 2,732° F., such an annealing temperature within the annealing temperature range can be around 1,093° F. In another non-limiting example discussed above, for Zircaloy with a melting point of around 3,362° F., such an annealing temperature within the annealing temperature range can be around 1,345° F. As another example, for some metallic components an annealing temperature within the annealing temperature range may be selected up to around 122° F. above an austenic temperature of the metal (as determined by the metal's percentage composition of carbon). Given by way of non-limiting examples, based upon such a material property an annealing temperature range could be between around 1360° F. and around 1482° F. for carbon compositions above around 0.8 percent. As a further non-limiting example, based upon such a material property an annealing temperature range could range between around 1360° F. and around 1482° F. for carbon compositions around 0.8 percent and vary substantially linearly up to an annealing temperature range between around 1657° F. and around 1774° F. for carbon compositions around 0 percent. However, in some embodiments, a maximum temperature of the annealing temperature range may be selected as desired to provide a predetermined safety margin below a melting point of at least one component of the nuclear fission fuel assembly. In some other embodiments, a maximum temperature of the annealing temperature range may be selected as desired to provide a predetermined safety margin below structural degradation of at least one component of the nuclear fission fuel assembly. As also discussed above, the creep rate in conjunction with internal vs. external pressures on the fuel element and/or fuel assemblies may affect annealing temperatures (and also annealing times). Annealing at least the portion of the at least one metallic component at the block 106 can be performed in various locations, as desired. For example, in some embodiments annealing at least the portion of the at least one metallic component can be performed in-place. However, the at least one metallic component need not be annealed in-place. For example and referring now to FIG. 1B, in some other embodiments the nuclear fission fuel assembly (that includes the component to be annealed) may be moved at a block 110 from an in-place location prior to annealing at the block 106. In one arrangement, annealing may be performed within a reactor core of the nuclear fission reactor. For example, the nuclear fission fuel assembly may be moved from its in-place location to another location within the reactor core where the annealing is to take place. In another arrangement, annealing may be performed external of a reactor core of the nuclear fission reactor. For example, the nuclear fission fuel assembly may be moved from its in-place location to a location external of the reactor core but still internal to the reactor pressure vessel where the annealing is to take place. As another example, the nuclear fission fuel assembly may be moved from its in-place location to a location external of the reactor pressure vessel where the annealing is to take place. In such a case, annealing may be performed on-site of the nuclear fission reactor or off-site from the nuclear fission reactor, as desired. In some embodiments and referring now to FIG. 1C, the nuclear fission fuel assembly may be moved to a location within a reactor core of the nuclear fission reactor after annealing. For example, the annealed fuel assembly may be moved to its in-place location or any other in-core location as desired after having been annealed in a location other than its in-place location. In some embodiments and referring to FIG. 1D, at a block 113 the nuclear fission fuel assembly may be re-oriented. Given by way of non-limiting example, the nuclear fission fuel assembly may be rotated 180 degrees for replacement in the reactor core. In such an arrangement, an end of the nuclear fission fuel assembly that was adjacent a cold leg inlet can be re-oriented for replacement in the reactor core adjacent a hot let outlet, and vice versa. That is, at the block 106 the nuclear fission fuel assembly can be turned “upside down” for replacement in the reactor core. In some embodiments, an entire nuclear fission fuel assembly need not be re-oriented at the block 113 in order for the nuclear fission fuel assembly to be considered re-oriented. For example, one or more fuel elements may be re-oriented within the nuclear fission fuel assembly in the same manner as described above (that is, rotated 180 degrees or turned “upside down”). It will be appreciated that the nuclear fission fuel assembly may be re-oriented before annealing or after annealing, as desired. In some other embodiments and referring to FIG. 1E, at a block 115 the nuclear fission fuel assembly may be reconfigured. Given by way of non-limiting example, components (or portions of components) of the nuclear fission fuel assembly, such as fuel elements (or portions of fuel elements), may be removed from their original position in the nuclear fission fuel assembly and replaced in a different position in the nuclear fission fuel assembly. For example, a portion of a fuel element that was located away from an end (such as toward a middle) of the fuel element can be removed and swapped with a portion of the fuel element that was located toward an end of the fuel element, thereby reconfiguring the fuel element and, as a result, the nuclear fission fuel assembly. It will be appreciated that the nuclear fission fuel assembly may be reconfigured before annealing or after annealing, as desired. As another example and referring to FIG. 1F, at a block 114 the nuclear fission fuel assembly may be moved from an in-place location after annealing. In such an arrangement, the annealed fuel assembly may be moved to another location other than its original in-place location after having been annealed in-place. Such relocations as described above may be performed as part of a fuel assembly utilization plan, if desired. Annealing at least the portion of the at least one metallic component of the nuclear fission fuel assembly within the annealing temperature range at the block 106 can be performed in various manners as desired for a particular application. For example and referring to FIG. 1G, in some embodiments annealing at least the portion of the at least one metallic component of the nuclear fission fuel assembly within the annealing temperature range at the block 106 can include adjusting operational parameters of the nuclear fission reactor to establish operating conditions of a region of the nuclear fission reactor containing the at least one metallic component within the determined annealing temperature range for a period of time selected to produce annealing of at least the portion of the at least one metallic component at a block 116. It will be appreciated that in some arrangements one or more portions of a component (such as portions that generate heat from nuclear fission during power range operations or that generate decay heat) that are hotter than other portions of the component may experience more annealing effect than the other portions of the component. In some embodiments and referring to FIG. 1H, adjusting operational parameters at the block 116 can include raising temperature of the region of the nuclear fission reactor containing the at least one metallic component from a predetermined operating temperature range of the reactor core toward the annealing temperature range at a block 118. Referring to FIG. 1I, adjusting operational parameters at the block 116 can include maintaining temperature of the region of the nuclear fission reactor containing the at least one metallic component substantially within the annealing temperature range at a block 120. Illustrative details regarding adjusting operational parameters to raise and/or maintain temperature and regarding selecting a period of time to produce annealing will be discussed below. Referring now to FIG. 1J, adjusting operational parameters at the block 116 can include providing heat from an external heat source at a block 122. It will be appreciated that an external heat source can be placed in thermal communication with a portion of the component or all or substantially all of the component, as desired for a particular application. In some arrangements, placing an external heat source in thermal communication with a portion of the component can help permit annealing one portion of the component. In such an arrangement, other portions of the component may experience less annealing effect than the portion in thermal communication with the external heat source. In some arrangements (such as in a reactor core that is shut down), the portion in thermal communication with the external heat source may experience an annealing effect and the other portions may experience little or no annealing effect. In some embodiments, the external heat source can include at least one electrical heat source. In some other embodiments, the external heat source can include at least one source of residual heat. For example, the residual heat can include decay heat. Given by way of non-limiting example, the decay heat may be generated by nuclear fission fuel material of one or more nuclear fission fuel elements of a fuel assembly that contains the metallic component being annealed and/or by nuclear fission fuel material of one or more nuclear fission fuel elements of one or more fuel assemblies that do not contain the metallic component being annealed. In some cases, such as when the metallic component to be annealed is cladding or metallic nuclear fission fuel material, the decay heat may be generated by nuclear fission fuel material of the nuclear fission fuel element that contains the metallic component being annealed. In some other embodiments, the external heat source can include a heating fluid. For example, a heating fluid can be placed in thermal communication with the metallic component to be annealed. In such an arrangement the temperature of the heating fluid can be established around a predetermined temperature to produce a desired annealing effect. The heating fluid, by way of non-limiting example, may include the reactor coolant as a major component of the heating fluid. In this example, the temperature of the heating fluid is brought to a desired temperature by any one or more of the methods discussed above and placed in thermal communication with the metallic component to be annealed. By way of another non-limiting example, the fluid my be substantially different from the reactor coolant and may include any non-reactive fluid, such as nitrogen, argon, helium, and/or combinations of these fluids, with the reactor coolant. The non-reactive fluid temperature may also be controlled by any one or more of the methods discussed above. In some other embodiments and referring to FIG. 1K, adjusting operational parameters at the block 116 can include substantially maintaining coolant flow rate at a block 124 and reducing an amount of heat transferred from the coolant at a block 126. In such an arrangement, in some embodiments the heat transfer that is reduced is the heat transfer from the reactor coolant to a heat exchanger. Given by way of non-limiting example, an amount of heat transferred from the coolant can be reduced by reducing an amount of fluid that exits a secondary side of a heat exchanger through which reactor coolant flows on a primary side. For example, a valve can be throttled toward a shut position on a secondary side of a primary-to-secondary heat exchanger in a pressurized water reactor, a pool-type liquid metal fast breeder reactor, or a gas-cooled fast breeder reactor. As a further example, a valve can be throttled toward a shut position on an intermediate side of an intermediate heat exchanger in a loop-type liquid metal fast breeder reactor. Given by way of further example, a heat load presented to any of the heat exchangers described above can be reduced. Similarly and referring to FIG. 1L, in some embodiments adjusting operational parameters at the block 116 can include substantially maintaining coolant flow rate at the block 124 and reducing an amount of heat transferred to the coolant at a block 128. In such an arrangement, in some embodiments the heat transfer that is reduced is the heat transfer from the nuclear fission fuel assembly containing the metallic component to be annealed to the reactor coolant. For example, if the heat transferred to a heat sink, such as a heat exchanger like a steam generator or the like, is reduced then the primary coolant temperature increases. This temperature increase of the reactor coolant in turn causes the temperature of the nuclear fission fuel assembly containing the metallic component to be annealed to rise for a given heat flux (that is, if decay heat is used as a heat source then the heat generation rate in the fuel will be roughly constant on short time scales). The nuclear fission fuel assembly then reaches a new temperature based on the new rate of heat rejection at the secondary loop or intermediate loop, as the case may be for a particular reactor application. To that end and given by way of non-limiting example, an amount of heat transferred to the coolant can be reduced by reducing an amount of fluid that exits a secondary side of a heat exchanger through which reactor coolant flows on a primary side. For example, a valve can be throttled toward a shut position on a secondary side of a primary-to-secondary heat exchanger in a pressurized water reactor, a pool-type liquid metal fast breeder reactor, or a gas-cooled fast breeder reactor. As a further example, a valve can be throttled toward a shut position on an intermediate side of an intermediate heat exchanger in a loop-type liquid metal fast breeder reactor. Given by way of further example, a heat load presented to any of the heat exchangers described above can be reduced. In other embodiments, referring to FIG. 1M adjusting operational parameters at the block 116 can include lowering, from a predetermined coolant flow rate, coolant flow rate into the region of the nuclear fission reactor containing the at least one metallic component at a block 130. For example, coolant flow rate can be lowered by throttling down a flow adjustment device, such as a valve. As another example, coolant flow rate can be lowered by shifting reactor coolant pump speed downward, such as from fast speed to slow speed, or by reducing the number of operating reactor coolant pumps. In other embodiments, referring to FIG. 1N adjusting operational parameters at the block 116 can include reversing direction of reactor coolant flow into the region of the nuclear fission reactor containing the at least one metallic component at a block 131. For example, in some arrangements reactor coolant flow can be reversed by appropriate positioning of cutoff valves and check valves. In some other arrangements, such as when the reactor coolant includes an electrically-conductive liquid reactor coolant, such as liquid metals, coolant flow can be reversed by an appropriate electrical device that can electrically control flow of electrically-conductive liquids. In other embodiments, referring to FIG. 1O adjusting operational parameters at the block 116 can include raising temperature of coolant entering the region of the nuclear fission reactor containing the at least one metallic component at a block 132. For example, reactivity level can be raised (such as, without limitation, by withdrawing control rods or otherwise removing neutron absorbing material) thereby increasing the amount of heat transferred from the nuclear fission fuel elements to the reactor coolant and, thus, raising temperature of reactor coolant for a given coolant flow rate. It will be appreciated that a negative temperature coefficient of reactivity (that is, negative αT) can help maintain inherent stability of the nuclear fission reactor in such cases. In some embodiments, referring to FIG. 1P adjusting operational parameters at the block 116 can include replacing at least a portion of a first reactor coolant having first heat transfer characteristics with second coolant having second heat transfer characteristics at a block 134. For example, in a liquid metal fast breeder reactor some or all of sodium reactor coolant may be replaced with lead reactor coolant or lead-bismuth reactor coolant; some or all of lead reactor coolant may be replaced with sodium reactor coolant or lead-bismuth reactor coolant; and some or all of lead-bismuth reactor coolant may be replaced with sodium reactor coolant or lead reactor coolant. Similarly, in a gas-cooled fast breeder reactor gaseous helium reactor coolant may be replaced with, given by way of non-limiting examples, gaseous argon, nitrogen, or supercritical carbon dioxide reactor coolant. In a pressurized water reactor, the liquid water reactor coolant may be replaced with, given by way of non-limiting examples, steam, inert gas, or the like. Referring now to FIG. 1Q, in some embodiments adjusting operational parameters at the block 116 can include raising pressure in the region of the nuclear fission reactor containing the at least one metallic component at a block 136. For example, pressure can be raised by a pressurizer (such as by energizing additional heaters in the pressurizer). Raising pressure can raise the temperature at which reactor coolant boils, thereby permitting raising temperature of the reactor coolant (and thus temperature of the metallic component to be annealed) without inducing local boiling in the reactor coolant in the region of the nuclear fission reactor containing the at least one metallic component. Referring now to FIG. 1R, in some other embodiments adjusting operational parameters at the block 116 can include lowering pressure in the region of the nuclear fission reactor containing the at least one metallic component. For example, pressure can be lowered by a pressurizer (such as by de-energizing heaters in the pressurizer). In some cases it may be desirable for local boiling to occur in the region of the nuclear fission reactor containing the at least one metallic component. Lowering pressure can lower the temperature at which reactor coolant boils, thereby permitting inducing local boiling in the reactor coolant in the region of the nuclear fission reactor containing the at least one metallic component. Because boiling is an isothermal process, substantially even temperature distribution can be maintained within the nuclear fission fuel assembly to be annealed. A substantially even temperature distribution may be maintained even in the event of increased heat generation rate of the heat generating material within the nuclear fission fuel assembly Referring now to FIG. 1S, a determination may be made at a block 140 regarding when the at least one metallic component of the nuclear fission fuel assembly is to be annealed. The determination of when the at least one metallic component of the nuclear fission fuel assembly is to be annealed may be made at the block 140 in a variety of manners, as desired for a particular application. For example, in some embodiments and referring to FIG. 1T, determining when the at least one metallic component of the nuclear fission fuel assembly is to be annealed at the block 140 can include scheduling a predetermined time for annealing the at least one metallic component of the nuclear fission fuel assembly at a block 142. In some embodiments, the predetermined time may be scheduled during design of the reactor core assembly. In such a case, annealing may be considered to be part of reactor operation. As such, annealing may be performed for the reactor core assembly in bulk, if desired. Moreover, if applicable, bulk annealing of the reactor core assembly may be performed on a periodic schedule. In some other embodiments, determining when the at least one metallic component of the nuclear fission fuel assembly is to be annealed at the block 140 may be based upon history of the at least one metallic component. For example, determining when the at least one metallic component of the nuclear fission fuel assembly is to be annealed at the block 140 may be based upon an annealing history of the at least one metallic component. That is, in some embodiments historical data regarding time between annealing operations for a metallic component may be used to predict and schedule future annealing operations for the metallic component. As another example, determining when the at least one metallic component of the nuclear fission fuel assembly is to be annealed at the block 140 may be based upon an operational history of the nuclear fission fuel assembly. Given by way of non-limiting example, the operational history of the nuclear fission fuel assembly may include temperature history and/or radiation exposure or the like. In some embodiments, it may be known that materials typically are brought to annealing conditions at a certain operational time (such as may be measured in effective full power hours) or at a specific location within a reactor core assembly. In such a case, determining when to anneal the metallic component may be based on input from fluence history and temperature history. This fluence and temperature input may then be input into a calculation that can estimate (i) extent of radiation damage, if any; (ii) if annealing is needed; and (iii) in cases where annealing is needed, which annealing parameters are to be used. In some other embodiments and referring now to FIG. 1U, determining when the at least one metallic component of the nuclear fission fuel assembly is to be annealed at the block 140 may include testing materials that are indicative of the at least one metallic component of the nuclear fission fuel assembly at a block 144. In some embodiments and referring to FIG. 1V, testing materials that are indicative of the at least one metallic component of the nuclear fission fuel assembly at the block 144 can include testing at least a portion of the at least one metallic component of the nuclear fission fuel assembly at a block 146. Given by way of non-limiting example, referring to FIG. 1W testing materials that are indicative of the at least one metallic component of the nuclear fission fuel assembly at the block 146 may include testing for changes in material properties indicative of radiation damage at a block 148. For example, some illustrative material properties indicative of radiation damage may include electrical resistivity, physical dimensions, displacement response to physical stress, response to stimulus, speed of sound within material, ductile-to-brittle transition temperature, and/or radiation emission. In some embodiments and referring to FIG. 1X, annealing at least the portion of the at least one metallic component of the nuclear fission fuel assembly within the annealing temperature range is stopped at a block 150. That is, in some embodiments temperature may be returned from the annealing temperature range toward the predetermined operating temperature range. In some cases, temperature may be reduced to ambient (such as when a reactor is shut down, cooled down, and depressurized for maintenance or any other application as desired). A determination of when to stop annealing at least the portion of the at least one metallic component of the nuclear fission fuel assembly within the annealing temperature range at the block 150 may be made in any manner as desired for a particular application. For example, in some embodiments annealing may be stopped at the block 150 after a predetermined time period. Given by way of non-limiting example, the predetermined time period may be a function of temperature. For example, the predetermined time period may have an inverse relationship to the annealing temperature (that is, the lower the annealing temperature the longer the predetermined time period, and vice versa). In some other embodiments, the predetermined time period may be a function of changes in material properties indicative of radiation damage. For example, the predetermined time period may be directly (as opposed to inversely) proportional to changes in material properties indicative of radiation damage. In some cases, for a given annealing temperature the predetermined time period may be proportional to an amount or extent of radiation damage throughout the at least one metallic component. In some other cases, for a given annealing temperature the predetermined time period may be proportional to severity of radiation damage regardless of amount or extent of radiation damage throughout the at least one metallic component. In some embodiments the predetermined time period may be a function of radiation exposure. In such an arrangement, radiation damage to the at least one metallic component need not be determined. In some cases, for a given annealing temperature the predetermined time period may be proportional to energy of the neutron spectrum to which the at least one metallic component has been exposed. For example, a predetermined time period associated with exposure to a fast neutron spectrum (such as in a fast breeder reactor) may be longer than a predetermined time period associated with exposure to a thermal fission spectrum (such as in a pressurized water reactor). In some other cases, for a given annealing temperature the predetermined time period may be proportional to length of time of exposure. For example, longer exposure of a metallic component may entail a longer predetermined time period of annealing before stopping the annealing operation. However, it will be appreciated that exposure in a thermal reactor may entail additional exposure time to result in equivalent exposure time in a fast reactor. Referring now to FIG. 1Y, in some embodiments material properties of at least a portion of the at least one metallic component of the nuclear fission fuel assembly may be tested at a block 152 during annealing at the block 106. In such an arrangement, annealing at the block 106 is stopped at a block 150A responsive to testing material properties of at least a portion of the at least one metallic component of the nuclear fission fuel assembly. For example, results of testing of material properties can be monitored during annealing. When monitored results of a desired parameter have returned within desired levels, then annealing may be stopped. It will also be appreciated that, as discussed above, the creep rate in conjunction with internal vs. external pressures on the fuel element and/or fuel assemblies may limit annealing times (as well as temperatures). After annealing has been stopped at the block 150, it may be desirable in some embodiments to further treat that which has been annealed. To that end and referring now to FIG. 1Z, in some embodiments at a block 156 at least the portion of the at least one metallic component of the nuclear fission fuel assembly can be treated with post-annealing treatment. In some embodiments, post-annealing treatment can include quenching. Quenching can produce a phase of crystal types in the material of the metallic component, thereby hardening the material. To that end and referring to FIG. 1AA, in some embodiments post-anneal treating at least the portion of the at least one metallic component of the nuclear fission fuel assembly at the block 156 can include lowering temperature from the annealing temperature range to a quenching temperature range at a block 158. The quenching temperature range suitably is sufficiently low enough to cool the material that has been annealed. Given by way of non-limiting example, in some embodiments a suitable quenching temperature range can be around 200° C.-300° C. (392° F.-572° F.). However, any suitable quenching temperature range may be selected as desired for a particular application. For example, in some embodiments in which the reactor coolant is a liquid metal, it will be appreciated that the quenching temperature range should be sufficiently high enough for a liquid metal reactor coolant to remain in liquid phase. Given by way of non-limiting examples, sodium has a melting point of 207.9° F., lead-bismuth eutectic has a melting point of 254.3° F., and lead has a melting point of 327.5° F. In such arrangements, the quenching temperature range may be selected to be as low as desired to cool the material to perform quenching yet be high enough to keep the liquid metal reactor coolant in liquid phase. It will be noted that it may be desirable to lower temperature at the block 158 at a rate sufficient to achieve a quenching effect. To that end and referring to FIG. 1AB, in some embodiments lowering temperature to a quenching temperature range at the block 158 can include lowering temperature at a predetermined rate at a block 160. It will be appreciated that such a predetermined rate of lowering temperature may be selected as desired for a particular application and may depend on various factors, such as without limitation material to be quenched, amount of hardening desired, limitations on rate of lowering temperature due to reactor plant construction characteristics, and the like. If desired, in some embodiments a reactor plant may be shut down and cooled down and/or depressurized to help lower temperature toward the quenching temperature. In some other embodiments, replacement reactor coolant (for example, at a lower temperature than existing reactor coolant) may be introduced into the reactor core to help lower temperature toward the quenching temperature. In some other embodiments, post-annealing treatment can also include tempering after quenching. While quenching can produce a phase, tempering can grow the produced phase to any gaps in a grain boundary, thereby helping to relax grain boundary stress that may have developed during annealing and, as a result, toughening the material. To that end and referring to FIG. 1AC, in some embodiments post-anneal treating at least the portion of the at least one metallic component of the nuclear fission fuel assembly at the block 156 can also include raising temperature from the quenching temperature range to a tempering temperature range at a block 162. The quenching temperature range suitably is any temperature range as desired that is between the quenching temperature range and the annealing temperature range. In some embodiments the tempering temperature range may be higher than the operating temperature range. In some other embodiments the tempering temperature range may be lower than the operating temperature range. Referring now to FIG. 1AD, in some embodiments after annealing has stopped at the block 150 temperature may be established at an operational temperature range at a block 166, if desired. Referring now to FIG. 1AE, in some embodiments annealing at the block 106 can be performed after commencement of transition of reactivity condition of at least a portion of the nuclear fission reactor from a first state to a second state at a block 154. Given by way of non-limiting example, the first state can include power range operation and the second state can include a shut-down state. It will be appreciated that any number of metallic components of any number of fuel assemblies may be annealed, as desired for a particular application. For example, in some embodiments fewer than all nuclear fission fuel assemblies of a reactor core of the nuclear fission reactor can be annealed. In some other embodiments, substantially all nuclear fission fuel assemblies of a reactor core of the nuclear fission reactor can be annealed, as desired. Other illustrative methods will be described below. Referring now to FIG. 2A, the illustrative method 200 for annealing at least a portion of at least one component of a reactor core of a nuclear fission reactor begins at a block 202. At a block 204 an annealing temperature range (that is higher than a predetermined operating temperature range of the reactor core) for at least a portion of at least one component of the reactor core of a nuclear fission reactor is determined. At a block 206 at least the portion of the at least one component is annealed within the annealing temperature range. The method 200 stops at a block 208. Illustrative aspects will be described briefly below. While the method 100 (FIG. 1A) discloses annealing at least a portion of at least one metallic component of a nuclear fission fuel assembly of a nuclear fission reactor, the method 200 discloses annealing at least a portion of any component or components of a reactor core of a nuclear fission reactor. Thus, the method 200 can be used for annealing at least a portion of one or more reactor core components such as without limitation a nuclear fission fuel assembly, a reactor core cooling component, and/or a reactor core structural member, non-limiting examples of which are discussed above. In some arrangements, the method 200 can be used to anneal at least a portion of one or more components of a nuclear fission fuel assembly, such as without limitation cladding, a cooling component, a structural member, a thermally conductive member, and/or nuclear fission fuel material, non-limiting examples of which are discussed above. Moreover, the method 200 can be performed on any component of a reactor core—regardless of whether the component is metallic or not. Thus, annealing as disclosed by the method 200 can permit use in reactor cores of advanced materials, such as composite materials like SiC/SiC or the like. However, it will be appreciated that the method 200 may also be used for a one or more reactor core components that are made of metals—such as without limitation steel, oxide dispersion strengthened (ODS) steels, austenitic steels (304, 316), ferritic/martensitic steels refractory metal, a refractory metal alloy, a non-ferrous metal, a non-ferrous metal alloy, and/or a superalloy (such as Inconels, Zircaloys, and/or Hastelloys). Further, while the annealing temperature range for the method 100 (FIG. 1A) need not be higher than an operating temperature range, it will also be noted that the annealing temperature range determined at the block 204 is higher than the predetermined operating temperature range of the reactor core. With the exception of the differences noted directly above, other aspects of the method 200 are similar to aspects of the method 100 (FIG. 1A). To that end and for sake of brevity, aspects of the method 200 will be described briefly. As noted above, the annealing temperature range determined at the block 204 is higher than a predetermined operating temperature range of the reactor core. The discussion of optional arrangements of the method 100 (FIG. 1A) in which annealing temperature range is higher than a predetermined operating temperature range is applicable to the method 200. Thus, details of the annealing temperature range being higher than the operating temperature range need not be repeated for an understanding. However, aspects of the method 200 will be noted below for completeness. For example, in some embodiments, the annealing temperature range may be determined based upon any one or more factors such as radiation exposure of the at least one component, an operating temperature history during which the radiation occurred, and/or an annealing history of the component to be annealed. In some other embodiments, the annealing temperature range may be determined based upon material properties of the at least one component. For example, in some embodiments a minimum temperature of the annealing temperature range may be at least around thirty percent of a melting point of the at least one component. As another example, in some other embodiments, an annealing temperature within the annealing temperature range may be around forty percent of a melting point of the at least one component. In some embodiments, a maximum temperature of the annealing temperature range may be selected as desired to provide a predetermined safety margin below a melting point of at least one component. In some other embodiments, a maximum temperature of the annealing temperature range may be selected as desired to provide a predetermined safety margin below structural degradation of at least one component. Annealing at least the portion of the at least one component at the block 206 can be performed in various locations, as desired. For example, in some embodiments annealing at least the portion of the at least one component can be performed in-place. However, the at least one component need not be annealed in-place. For example and referring now to FIG. 2B, in some other embodiments the at least one component to be annealed may be moved at a block 210 from an in-place location prior to annealing at the block 206. In one arrangement, annealing may be performed within a reactor core of the nuclear fission reactor. For example, the at least one component may be moved from its in-place location to another location within the reactor core where the annealing is to take place. In another arrangement, annealing may be performed external of a reactor core of the nuclear fission reactor. For example, the at least one component may be moved from its in-place location to a location external of the reactor core but still internal to the reactor pressure vessel where the annealing is to take place. As another example, the at least one component may be moved from its in-place location to a location external of the reactor pressure vessel where the annealing is to take place. In such a case, annealing may be performed on-site of the nuclear fission reactor or off-site from the nuclear fission reactor, as desired. In some embodiments and referring now to FIG. 2C, the at least one component may be moved to a location within a reactor core of the nuclear fission reactor after annealing. In some embodiments and referring to FIG. 2D, at a block 213 the at least one component may be re-oriented. In some other embodiments and referring to FIG. 2E, at a block 215 the at least one component may be reconfigured. As another example and referring to FIG. 2F, at a block 214 the at least one component may be moved from an in-place location after annealing. Annealing at least the portion of the at least one component of the reactor core within the annealing temperature range at the block 206 can be performed in various manners as desired for a particular application. For example and referring to FIG. 2G, in some embodiments annealing at least the portion of the at least one component within the annealing temperature range at the block 206 can include adjusting operational parameters of the nuclear fission reactor to establish operating conditions of a region of the nuclear fission reactor containing the at least one component within the determined annealing temperature range for a period of time selected to produce annealing of the at least one metallic component at a block 216. In some embodiments and referring to FIG. 2H, adjusting operational parameters at the block 216 can include raising temperature of the region of the nuclear fission reactor containing the at least one component from a predetermined operating temperature range of the reactor core toward the annealing temperature range at a block 218. Referring to FIG. 2I, adjusting operational parameters at the block 216 can include maintaining temperature of the region of the nuclear fission reactor containing the at least one component substantially within the annealing temperature range at a block 220. Illustrative details regarding adjusting operational parameters to raise and/or maintain temperature and regarding selecting a period of time to produce annealing will be discussed below. Referring now to FIG. 2J, adjusting operational parameters at the block 216 can include providing heat from an external heat source at a block 222. In some embodiments, the external heat source can include at least one electrical heat source. In some other embodiments, the external heat source can include at least one source of residual heat. For example, the residual heat can include decay heat. In some other embodiments, the external heat source can include a heating fluid. In some other embodiments and referring to FIG. 2K, adjusting operational parameters at the block 216 can include substantially maintaining coolant flow rate at a block 224 and reducing an amount of heat transferred from the coolant at a block 226. Similarly and referring to FIG. 2L, in some embodiments adjusting operational parameters at the block 216 can include substantially maintaining coolant flow rate at the block 224 and reducing an amount of heat transferred to the coolant at a block 228. In other embodiments, referring to FIG. 2M adjusting operational parameters at the block 216 can include lowering, from a predetermined coolant flow rate, coolant flow rate into the region of the nuclear fission reactor containing the at least one component at a block 230. In other embodiments, referring to FIG. 2N adjusting operational parameters at the block 216 can include reversing direction of reactor coolant flow into the region of the nuclear fission reactor containing the at least one component at a block 231. In other embodiments, referring to FIG. 2O adjusting operational parameters at the block 216 can include raising temperature of coolant entering the region of the nuclear fission reactor containing the at least one component at a block 232. In some embodiments, referring to FIG. 2P adjusting operational parameters at the block 216 can include replacing at least a portion of a first reactor coolant having first heat transfer characteristics with second coolant having second heat transfer characteristics at a block 234. Referring now to FIG. 2Q, in some embodiments adjusting operational parameters at the block 216 can include raising pressure in the region of the nuclear fission reactor containing the at least one component at a block 236. Referring now to FIG. 2R, in some other embodiments adjusting operational parameters at the block 216 can include lowering pressure in the region of the nuclear fission reactor containing the at least one component. Referring now to FIG. 2S, a determination may be made at a block 240 regarding when the at least one component is to be annealed. The determination of when the at least one component is to be annealed may be made at the block 240 in a variety of manners, as desired for a particular application. For example, in some embodiments and referring to FIG. 2T, determining when the at least one component is to be annealed at the block 240 can include scheduling a predetermined time for annealing the at least one component at a block 242. In some other embodiments, determining when the at least one component is to be annealed at the block 240 may be based upon history of the at least one component. For example, determining when the at least one component is to be annealed at the block 240 may be based upon an annealing history of the at least one component. As another example, determining when the at least one component is to be annealed at the block 240 may be based upon an operational history of the at least one component. Given by way of non-limiting example, the operational history of the nuclear fission fuel assembly may include temperature history and/or radiation exposure or the like. In some other embodiments and referring now to FIG. 2U, determining when the at least one component is to be annealed at the block 240 may include testing materials that are indicative of the at least one component at a block 244. In some embodiments and referring to FIG. 2V, testing materials that are indicative of the at least one component at the block 244 can include testing at least a portion of the at least one component at a block 246. Given by way of non-limiting example, referring to FIG. 2W testing materials that are indicative of the at least one component at the block 246 may include testing for changes in material properties indicative of radiation damage at a block 248. For example, some illustrative material properties indicative of radiation damage may include electrical resistivity, physical dimensions, displacement response to physical stress, response to stimulus, speed of sound within material, ductile-to-brittle transition temperature, and/or radiation emission. In some embodiments and referring to FIG. 2X, annealing at least the portion of the at least one component within the annealing temperature range is stopped at a block 250. A determination of when to stop annealing at least the portion of the at least one component within the annealing temperature range at the block 250 may be made in any manner as desired for a particular application. For example, in some embodiments annealing may be stopped at the block 250 after a predetermined time period. Given by way of non-limiting example, the predetermined time period may be a function of temperature. In some other embodiments, the predetermined time period may be a function of changes in material properties indicative of radiation damage. In some embodiments the predetermined time period may be a function of radiation exposure. Referring now to FIG. 2Y, in some embodiments material properties of at least a portion of the at least one component may be tested at a block 252 during annealing at the block 206. In such an arrangement, annealing at the block 206 is stopped at a block 250A responsive to testing material properties of at least a portion of the at least one component. After annealing has been stopped at the block 250, it may be desirable in some embodiments to further treat that which has been annealed. To that end and referring now to FIG. 2Z, in some embodiments at a block 256 at least the portion of the at least one component can be treated with post-annealing treatment. In some embodiments, post-annealing treatment can include quenching. To that end and referring to FIG. 2AA, in some embodiments post-anneal treating at least the portion of the at least one component at the block 256 can include lowering temperature from the annealing temperature range to a quenching temperature range at a block 258. Referring to FIG. 2AB, in some embodiments lowering temperature to a quenching temperature range at the block 258 can include lowering temperature at a predetermined rate at a block 260. In some other embodiments, post-anneal treating can also include tempering after quenching. To that end and referring to FIG. 2AC, in some embodiments post-anneal treating at least the portion of the at least one component at the block 256 can also include raising temperature from the quenching temperature range to a tempering temperature range at a block 262. Referring now to FIG. 2AD, in some embodiments after annealing has stopped at the block 250 temperature may be established at an operational temperature range at a block 266, if desired. Referring now to FIG. 2AE, in some embodiments annealing at the block 206 can be performed after commencement of transition of reactivity condition of at least a portion of the nuclear fission reactor from a first state to a second state at a block 254. Given by way of non-limiting example, the first state can include power range operation and the second state can include a shut-down state. It will be appreciated that any number of components and any number of fuel assemblies and their components may be annealed, as desired for a particular application. For example, in some embodiments fewer than all nuclear fission fuel assemblies of a reactor core of the nuclear fission reactor can be annealed. In some other embodiments, substantially all nuclear fission fuel assemblies of a reactor core of the nuclear fission reactor can be annealed, as desired. Referring now to FIG. 3A, the illustrative method 300 for treating at least a portion of at least one component of a reactor core of a nuclear fission reactor begins at a block 302. At a block 304 a temperature of a region of a reactor core of a nuclear fission reactor is elevated, from a predetermined operating temperature range to an annealing temperature range, for a time period sufficient to produce annealing of at least a portion of at least one selected component of the region of the reactor core without removing the at least one selected component from the reactor core. The method 300 stops at a block 306. Illustrative aspects will be described briefly below. While the method 100 (FIG. 1A) discloses annealing at least a portion of at least one metallic component of a nuclear fission fuel assembly of a nuclear fission reactor, the method 300 discloses annealing at least a portion of any component or components of a reactor core of a nuclear fission reactor. Thus, similar to the method 200 (FIG. 2A), the method 300 can be used for annealing at least a portion of one or more reactor core components such as without limitation a nuclear fission fuel assembly, a reactor core cooling component, and/or a reactor core structural member, non-limiting examples of which are discussed above. In some arrangements, the method 300 can be used to anneal at least a portion of one or more components of a nuclear fission fuel assembly, such as without limitation cladding, a cooling component, a structural member, a thermally conductive member, and/or nuclear fission fuel material, non-limiting examples of which are discussed above. Moreover (and also similar to the method 200 (FIG. 2A)), the method 300 can be performed on any component or components of a reactor core—regardless of whether the component is metallic or not. Thus, annealing as disclosed by the method 300 can permit use in reactor cores of advanced materials, such as composite materials like SiC/SiC or the like. However, it will be appreciated that the method 300 may also be used for a one or more reactor core components that are made of metals—such as without limitation steel, oxide dispersion strengthened (ODS) steels, austenitic steels (304, 316), ferritic/martensitic steels refractory metal, a refractory metal alloy, a non-ferrous metal, a non-ferrous metal alloy, and/or a superalloy (such as Inconels, Zircaloys, and/or Hastelloys). Further, while the annealing temperature range for the method 100 (FIG. 1A) need not be higher than an operating temperature range, it will be noted that at the block 304 temperature of a region of a reactor core of a nuclear fission reactor is elevated from a predetermined operating temperature range to an annealing temperature range. Lastly, while annealing performed by either the method 100 (FIG. 1A) or the method 200 (FIG. 2A) need not occur within a reactor core, it will also be noted that the method 300 can produce annealing of at least a portion of at least one selected component of the region of the reactor core without removing the at least one selected component from the reactor core. With the exception of the differences noted directly above, other aspects of the method 300 are similar to aspects of the method 100 (FIG. 1A). To that end and for sake of brevity, aspects of the method 300 will be described briefly. As noted above, at the block 304 temperature of a region of a reactor core of a nuclear fission reactor is elevated from a predetermined operating temperature range to an annealing temperature range. The discussion of optional arrangements of the method 100 (FIG. 1A) in which annealing temperature range is higher than a predetermined operating temperature range is applicable to the method 300. Thus, details of the annealing temperature range being higher than the operating temperature range need not be repeated for an understanding. However, aspects of the method 300 will be noted below for completeness. For example, in some embodiments, the annealing temperature range may be determined based upon any one or more factors such as radiation exposure of the at least one component, an operating temperature history during which the radiation occurred, and/or an annealing history of the component to be annealed. In some other embodiments, the annealing temperature range may be determined based upon material properties of the at least one selected component. For example, in some embodiments a minimum temperature of the annealing temperature range may be at least around thirty percent of a melting point of the at least one selected component. As another example, in some other embodiments, an annealing temperature within the annealing temperature range may be around forty percent of a melting point of the at least one selected component. In some embodiments, a maximum temperature of the annealing temperature range may be selected as desired to provide a predetermined safety margin below a melting point of at least one selected component. In some other embodiments, a maximum temperature of the annealing temperature range may be selected as desired to provide a predetermined safety margin below structural degradation of at least one selected component. Elevating the temperature to perform annealing at the block 304 can be performed in various locations of a reactor core, as desired. For example, in some embodiments elevating the temperature to perform annealing of at least the portion of the at least one selected component can be performed in-place. However, the at least one selected component need not be annealed in-place. For example and referring now to FIG. 3B, in some other embodiments the at least one selected component to be annealed may be moved at a block 310 from an in-place location prior to elevating the temperature to perform annealing at the block 304. It will be noted that, as discussed above, elevating the temperature to perform annealing is performed within a reactor core of the nuclear fission reactor. Thus, in some arrangements, the at least one selected component may be moved from its in-place location to another location within the reactor core where the annealing is to take place. In some embodiments and referring now to FIG. 3C, the at least one selected component may be moved to a location within a reactor core of the nuclear fission reactor after elevating the temperature to perform annealing. In some embodiments and referring to FIG. 3D, at a block 313 the at least one selected component may be re-oriented annealing. In some other embodiments and referring to FIG. 3E, at a block 315 the at least one selected component may be reconfigured. As another example and referring to FIG. 3F, at a block 314 the at least one selected component may be moved from an in-place location after elevating the temperature to perform annealing. Elevating the temperature to perform annealing at the block 304 can be performed in various manners as desired for a particular application. For example and referring to FIG. 3G, in some embodiments elevating the temperature to perform annealing at the block 304 can include adjusting operational parameters of the nuclear fission reactor to establish operating conditions of the region of the nuclear fission reactor containing the at least one selected component within the annealing temperature range for a period of time selected to produce annealing of at least the portion of the at least one metallic component at a block 316. In some embodiments and referring to FIG. 3H, adjusting operational parameters at the block 316 can include raising (that is, changing) temperature of the region of the nuclear fission reactor containing the at least one selected component from a predetermined operating temperature range of the reactor core toward the annealing temperature range at a block 318. Referring to FIG. 3I, adjusting operational parameters at the block 316 can include maintaining temperature of the region of the nuclear fission reactor containing the at least one selected component substantially within the annealing temperature range at a block 320. Illustrative details regarding adjusting operational parameters to raise and/or maintain temperature and regarding selecting a period of time to produce annealing will be discussed below. Referring now to FIG. 3J, adjusting operational parameters at the block 316 can include providing heat from an external heat source at a block 322. In some embodiments, the external heat source can include at least one electrical heat source. In some other embodiments, the external heat source can include at least one source of residual heat. For example, the residual heat can include decay heat. In some other embodiments, the external heat source can include a heating fluid. In some other embodiments and referring to FIG. 3K, adjusting operational parameters at the block 316 can include substantially maintaining coolant flow rate at a block 324 and reducing an amount of heat transferred from the coolant at a block 326. Similarly and referring to FIG. 3L, in some embodiments adjusting operational parameters at the block 316 can include substantially maintaining coolant flow rate at the block 324 and reducing an amount of heat transferred to the coolant at a block 328. In other embodiments, referring to FIG. 3M adjusting operational parameters at the block 316 can include lowering, from a predetermined coolant flow rate, coolant flow rate into the region of the nuclear fission reactor containing the at least one selected component at a block 330. In other embodiments, referring to FIG. 3N adjusting operational parameters at the block 316 can include reversing direction of reactor coolant flow into the region of the nuclear fission reactor containing the at least one selected component at a block 331. In other embodiments, referring to FIG. 3O adjusting operational parameters at the block 316 can include raising temperature of coolant entering the region of the nuclear fission reactor containing the at least one selected component at a block 332. In some embodiments, referring to FIG. 3P adjusting operational parameters at the block 316 can include replacing at least a portion of a first reactor coolant having first heat transfer characteristics with second coolant having second heat transfer characteristics at a block 334. Referring now to FIG. 3Q, in some embodiments adjusting operational parameters at the block 316 can include raising pressure in the region of the nuclear fission reactor containing the at least one selected component at a block 336. Referring now to FIG. 3R, in some other embodiments adjusting operational parameters at the block 316 can include lowering pressure in the region of the nuclear fission reactor containing the at least one selected component. Referring now to FIG. 3S, a determination may be made at a block 340 regarding when the at least one selected component is to be annealed. The determination of when the at least one selected component is to be annealed may be made at the block 340 in a variety of manners, as desired for a particular application. For example, in some embodiments and referring to FIG. 3T, determining when the at least one component is to be annealed at the block 340 can include scheduling a predetermined time for annealing the at least one selected component at a block 342. In some other embodiments, determining when the at least one selected component is to be annealed at the block 340 may be based upon history of the at least one selected component. For example, determining when the at least one selected component is to be annealed at the block 340 may be based upon an annealing history of the at least one selected component. As another example, determining when the at least one selected component is to be annealed at the block 340 may be based upon an operational history of the at least one selected component. Given by way of non-limiting example, the operational history of the nuclear fission fuel assembly may include temperature history and/or radiation exposure or the like. In some other embodiments and referring now to FIG. 3U, determining when the at least one selected component is to be annealed at the block 340 may include testing materials that are indicative of the at least one selected component at a block 344. In some embodiments and referring to FIG. 3V, testing materials that are indicative of the at least one selected component at the block 344 can include testing at least a portion of the at least one selected component at a block 346. Given by way of non-limiting example, referring to FIG. 3W testing materials that are indicative of the at least one selected component at the block 346 may include testing for changes in material properties indicative of radiation damage at a block 348. For example, some illustrative material properties indicative of radiation damage may include electrical resistivity, physical dimensions, displacement response to physical stress, response to stimulus, speed of sound within material, ductile-to-brittle transition temperature, and/or radiation emission. In some embodiments and referring to FIG. 3X, elevating the temperature to perform annealing is stopped at a block 350. A determination of when to stop elevating the temperature to perform annealing at the block 350 may be made in any manner as desired for a particular application. For example, in some embodiments elevating the temperature to perform annealing may be stopped at the block 350 after a predetermined time period. Given by way of non-limiting example, the predetermined time period may be a function of temperature. In some other embodiments, the predetermined time period may be a function of changes in material properties indicative of radiation damage. In some embodiments the predetermined time period may be a function of radiation exposure. Referring now to FIG. 3Y, in some embodiments material properties of at least a portion of the at least one selected component may be tested at a block 352 during elevating the temperature to perform annealing at the block 304. In such an arrangement, elevating the temperature to perform annealing at the block 304 is stopped at a block 350A responsive to testing material properties of at least a portion of the at least one selected component. After annealing has been stopped at the block 350, it may be desirable in some embodiments to further treat that which has been annealed. To that end and referring now to FIG. 3Z, in some embodiments at a block 356 at least the portion of the at least one selected component can be treated with post-annealing treatment. In some embodiments, post-annealing treatment can include quenching. To that end and referring to FIG. 3AA, in some embodiments post-anneal treating at least the portion of the at least one selected component at the block 356 can include lowering temperature from the annealing temperature range to a quenching temperature range at a block 358. Referring to FIG. 3AB, in some embodiments lowering temperature to a quenching temperature range at the block 358 can include lowering temperature at a predetermined rate at a block 360. In some other embodiments, post-annealing treatment can also include tempering after quenching. To that end and referring to FIG. 3AC, in some embodiments post-anneal treating at least the portion of the at least one selected component at the block 356 can also include raising temperature from the quenching temperature range to a tempering temperature range at a block 362. Referring now to FIG. 3AD, in some embodiments after annealing has stopped at the block 350 temperature may be established at an operational temperature range at a block 366, if desired. Referring now to FIG. 3AE, in some embodiments elevating the temperature to perform annealing at the block 304 can be performed after commencement of transition of reactivity condition of at least a portion of the nuclear fission reactor from a first state to a second state at a block 354. Given by way of non-limiting example, the first state can include power range operation and the second state can include a shut-down state. It will be appreciated that any number of components and any number of fuel assemblies and their components may be annealed, as desired for a particular application. For example, in some embodiments fewer than all nuclear fission fuel assemblies of a reactor core of the nuclear fission reactor can be annealed. In some other embodiments, substantially all nuclear fission fuel assemblies of a reactor core of the nuclear fission reactor can be annealed, as desired. Referring now to FIG. 4A, the illustrative method 400 for producing an annealing effect begins at a block 402. At a block 404 a reactor coolant system is adjusted to produce a temperature deviation from a nominal operating temperature range. At a block 406 the temperature deviation from the nominal operating temperature is maintained for a period selected to produce a selected annealing effect. At a block 408, after the period selected to produce the selected annealing effect, the reactor coolant system is adjusted to return to the nominal operating temperature range. The method 400 stops at a block 410. Illustrative details will be set forth below. In some embodiments, the selected annealing effect can anneal at least a portion of at least one reactor core component such as at least one nuclear fission fuel assembly, reactor core cooling component, and/or reactor core structural member. When at least one nuclear fission fuel assembly is annealed, the annealed component can include cladding, a cooling component, a structural member, a thermally conductive member, and/or nuclear fission fuel material. In some embodiments, the selected annealing effect can include a predicted annealing effect. That is, a desired extent of annealing to be performed can be predicted. The desired extent of annealing can be a function of one or more factors, such as annealing temperature, annealing time, material properties of a component to be annealed, exposure of the component to be annealed, operational history of the component to be annealed, and/or annealing history of the component to be annealed, all of which have been discussed above. In some other embodiments the selected annealing effect can include a measured annealing effect. That is, as discussed above material properties of the component can be monitored as desired during annealing. When the monitored material properties return to a desired range of values, the selected annealing effect has been produced, and the reactor coolant system can be adjusted to return to the nominal operating temperature range at the block 410. Referring to FIG. 4B, in some embodiments adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range at the block 404 can include adjusting reactor coolant flow at a block 412. As discussed above, reactor coolant flow can be adjusted by throttling a flow adjustment device, such as a valve. As another example, reactor coolant flow can be adjusted by shifting reactor coolant pump speed, such as between fast speed and slow speed, or by changing the number of operating reactor coolant pumps. In other embodiments, referring to FIG. 4C direction of reactor coolant flow at a block 413. Referring now to FIG. 4D and in some embodiments, in addition to adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range at the block 404, a rate of heat generation can be adjusted at a block 414. Given by way of non-limiting examples and as discussed above, heat generation can be adjusted by providing heat from an external heat source, such as at least one electrical heat source, a heating fluid, and/or at least one source of residual heat, such as decay heat. In addition, heat generation may be adjusted temporarily by adjusting reactivity, such as without limitation by withdrawing or inserting control rods or otherwise adjusting an amount of neutron absorbing material, thereby raising or lowering reactor coolant temperature. It will be appreciated that such an adjustment of heat generation may have a temporary effect on temperature in nuclear fission reactors with a negative temperature coefficient of reactivity αT. Referring to FIG. 4E, in some embodiments adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range at the block 404 can include adjusting a rate of heat transferred from the reactor coolant at a block 416. As discussed above, a rate of heat transferred from the reactor coolant can be adjusted in a number of ways. Given by way of non-limiting examples, an amount of fluid that exits a secondary side of a heat exchanger through which reactor coolant flows on a primary side can be adjusted; a valve can be throttled toward a shut position on a secondary side of a primary-to-secondary heat exchanger in a pressurized water reactor, a pool-type liquid metal fast breeder reactor, or a gas-cooled fast breeder reactor; a valve can be throttled toward a shut position on an intermediate side of an intermediate heat exchanger in a loop-type liquid metal fast breeder reactor; a heat load presented to any of the heat exchangers described above can be reduced; or the like. Referring to FIG. 4F, in some other embodiments adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range at the block 404 can include adjusting a rate of heat transferred to the reactor coolant at a block 418. As discussed above, a rate of heat transferred to the reactor coolant can be adjusted in a number of ways. Given by way of non-limiting examples, the heat transfer that is adjusted is the heat transfer from a fuel assembly containing the component to be annealed to the reactor coolant. Given by way of non-limiting examples, an amount of heat transferred to the coolant can be adjusted by adjusting an amount of fluid that exits a secondary side of a heat exchanger through which reactor coolant flows on a primary side. For example, a valve can be throttled on a secondary side of a primary-to-secondary heat exchanger in a pressurized water reactor, a pool-type liquid metal fast breeder reactor, or a gas-cooled fast breeder reactor. As a further example, a valve can be throttled on an intermediate side of an intermediate heat exchanger in a loop-type liquid metal fast breeder reactor. Given by way of further example, a heat load presented to any of the heat exchangers described above can be adjusted. Referring now to FIG. 4G, in some embodiments adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range at the block 404 can include replacing at least a portion of a first reactor coolant having first heat transfer characteristics with second coolant having second heat transfer characteristics at a block 420. Examples of replacement of at least a portion of reactor coolant have been discussed above. Referring to FIG. 4H, in some embodiments adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range at the block 404 can include adjusting temperature of reactor coolant at a block 422. Given by way of non-limiting examples, temperature of reactor coolant can be adjusted by adding pump heat, such as by increasing the number of operating reactor coolant pumps or by increasing the pump velocity for a given level of heat rejection from either the primary or intermediate cooling loops. It will be appreciated that, when the reactor is shut down, addition of pump heat can raise reactor coolant temperature. For a reactor coolant pump, pump power is proportional to velocity. Moreover, for a typical reactor coolant pump, around 1 MW or so of power is typically lost to inefficiency. This lost power is transferred as heat to reactor coolant in the reactor coolant loop. Referring now to FIG. 5A, the illustrative method 500 for annealing at least a portion of at least one component of a nuclear fission reactor core begins at a block 502. At a block 504 a nuclear fission reactor core is operated within a predetermined operating temperature range. At a block 506 the nuclear fission reactor core is shut down. At a block 508 temperature of at least a portion of the nuclear fission reactor core is raised above the predetermined operating temperature range to an annealing temperature range for at least one component of the nuclear fission reactor core. At a block 510 temperature of at least the portion of the nuclear fission reactor core is maintained within the annealing temperature range for a time period selected to perform annealing of at least a portion of the at least one component of the nuclear fission reactor core. The method 500 stops at a block 512. Illustrative details will be set forth below. After temperature of at least the portion of the nuclear fission reactor core was maintained within the annealing temperature range for the time period at the block 510, it may be desirable in some embodiments to further treat at least a part of that which has been annealed. To that end and referring now to FIG. 5B, in some embodiments at a block 556 at least the portion of the at least one component can be treated with post-annealing treatment. In some embodiments, post-annealing treatment can include quenching. To that end and referring to FIG. 5C, in some embodiments post-anneal treating at least the portion of the at least one component at the block 556 can include lowering temperature from the annealing temperature range to a quenching temperature range at a block 558. Referring to FIG. 5D, in some embodiments lowering temperature to a quenching temperature range at the block 558 can include lowering temperature at a predetermined rate at a block 560. In some other embodiments, post-annealing treatment can also include tempering after quenching. To that end and referring to FIG. 5E, in some embodiments post-anneal treating at least the portion of the at least one component at the block 556 can also include raising temperature from the quenching temperature range to a tempering temperature range at a block 562. Referring to FIG. 5F, in some embodiments, at a block 514 temperature of at least the portion of the nuclear fission reactor core can be lowered from the annealing temperature range toward the predetermined operating temperature range after temperature of at least the portion of the nuclear fission reactor core was maintained within the annealing temperature range for the time period. Referring now to FIG. 5G, after temperature of at least the portion of the nuclear fission reactor core has been lowered from the annealing temperature range at the block 514, the nuclear fission reactor core can be re-started at a block 516, as desired. It will be appreciated that applicable initial conditions for re-starting the reactor core should be met when re-starting the reactor core at the block 516. Referring now to FIG. 5H, substantially constant reactor coolant flow can be maintained through at least a portion of the nuclear fission reactor core at a block 518. It will be appreciated that reactor coolant flow can be adjusted as desired, if at all, during operation at the block 504 and shut down at the block 506. Thus, in some embodiments the reactor coolant flow may be maintained substantially constant while raising the temperature at the block 508 and maintaining temperature at the block 510. Referring to FIG. 5I, in some embodiments raising temperature of at least a portion of the nuclear fission reactor core above the predetermined operating temperature range to an annealing temperature range at the block 508 can include adding heat to fluid in thermal communication with at least a portion of the nuclear fission reactor core at a block 520. As discussed above, heat can be added by providing heat from an external heat source, such as at least one electrical heat source or a heating fluid. In addition, pump heat may be added to reactor coolant by operating reactor coolant pumps. Referring to FIG. 5J, in some embodiments raising temperature of at least a portion of the nuclear fission reactor core above the predetermined operating temperature range to an annealing temperature range at the block 508 can include generating decay heat at a block 522. Generation of decay heat has been discussed above. Referring now to FIG. 5K, in some embodiments raising temperature of at least a portion of the nuclear fission reactor core above the predetermined operating temperature range to an annealing temperature range at the block 508 can include reducing an amount of heat transferred to reactor coolant at a block 524. As discussed above, in some embodiments the heat transfer that can reduced is the heat transfer, such as decay heat, from a fuel assembly containing the component to be annealed to the reactor coolant. Given by way of non-limiting example, an amount of heat transferred to the coolant can be reduced by reducing an amount of fluid that exits a secondary side of a heat exchanger through which reactor coolant flows on a primary side. For example, a valve can be throttled toward a shut position on a secondary side of a primary-to-secondary heat exchanger in a pressurized water reactor, a pool-type liquid metal fast breeder reactor, or a gas-cooled fast breeder reactor. As a further example, a valve can be throttled toward a shut position on an intermediate side of an intermediate heat exchanger in a loop-type liquid metal fast breeder reactor. Given by way of further example, a heat load presented to any of the heat exchangers described above can be reduced. Referring to FIG. 5L, in some other embodiments raising temperature of at least a portion of the nuclear fission reactor core above the predetermined operating temperature range to an annealing temperature range at the block 508 can include reducing an amount of heat transferred from reactor coolant at a block 526. As discussed above, in some embodiments the heat transfer that is reduced is the heat transfer from the reactor coolant to a heat exchanger. Given by way of non-limiting example, an amount of heat transferred from the coolant can be reduced by reducing an amount of fluid that exits a secondary side of a heat exchanger through which reactor coolant flows on a primary side. For example, a valve can be throttled toward a shut position on a secondary side of a primary-to-secondary heat exchanger in a pressurized water reactor, a pool-type liquid metal fast breeder reactor, or a gas-cooled fast breeder reactor. As a further example, a valve can be throttled toward a shut position on an intermediate side of an intermediate heat exchanger in a loop-type liquid metal fast breeder reactor. Given by way of further example, a heat load presented to any of the heat exchangers described above can be reduced. Referring now to FIG. 5M, in some embodiments reducing an amount of heat transferred from reactor coolant at the block 526 can include maintaining reactor coolant flow rate substantially constant at a block 528. However, in some other embodiments and referring to FIG. 5N, reducing an amount of heat transferred from reactor coolant at the block 526 can include reducing reactor coolant flow rate at a block 530. Reducing reactor coolant flow rate has been discussed above. Referring to FIG. 5O, in some other embodiments reducing an amount of heat transferred from reactor coolant at the block 526 can include substantially stopping transfer of heat from reactor coolant at a block 532. Given by way of non-limiting example, heat transfer from the reactor coolant may be substantially stopped, if desired, by performing any one or more of the techniques discussed above for the block 526 in conjunction with reducing reactor coolant flow rate, as desired for a particular application. Referring to FIG. 5P, in some embodiments raising temperature of at least a portion of the nuclear fission reactor core above the predetermined operating temperature range to an annealing temperature range at the block 508 can include raising temperature of reactor coolant entering at least the portion of the nuclear fission reactor core at a block 534. Given by way of non-limiting example, temperature of reactor coolant can be raised by providing heat from an external heat source, such as at least one electrical heat source or a heating fluid. In addition, pump heat may be added to reactor coolant by operating reactor coolant pumps. Moreover, decay heat may also raise temperature of reactor coolant. Referring to FIG. 5Q, maintaining temperature of at least the portion of the nuclear fission reactor core within the annealing temperature range for a time period at the block 510 can include establishing substantially isothermal conditions within at least the portion of the nuclear fission reactor core within the annealing temperature range at a block 536. For example and referring now to FIG. 5R, in some embodiments establishing substantially isothermal conditions within at least the portion of the nuclear fission reactor core within the annealing temperature range at the block 536 can include transferring a reduced amount of heat by reactor coolant that is less than a predetermined amount of heat transferred by reactor coolant during reactor operation at a block 538. Referring now to FIG. 5S, in some embodiments reducing an amount of heat transferred to reactor coolant at the block 524 can include replacing at least a portion of a first reactor coolant having first heat transfer characteristics with second coolant having second heat transfer characteristics at a block 540. Replacing a portion of reactor coolant has been discussed above. Referring to FIG. 5T, in some embodiments raising temperature of at least a portion of the nuclear fission reactor core above the predetermined operating temperature range to an annealing temperature range for at least one component of the nuclear fission reactor core at the block 508 can include raising pressure in at least the portion of the nuclear fission reactor core at a block 542. Raising pressure has been discussed above. It will be appreciated that any portion of the reactor core may be annealed, as desired for a particular application. For example, in some embodiments less than all of the reactor core can be annealed. In some other embodiments, substantially all of the reactor core of the nuclear fission reactor can be annealed, as desired. Illustrative Systems and Apparatuses Illustrative systems and apparatuses will now be described. The illustrative systems and apparatuses can provide host environments for performance of any of the methods described herein. It will be appreciated that the illustrative systems and apparatuses shown in the accompanying FIGS. 8A-8K and described below are illustrated in functional block diagram form. As such, the block diagrams of FIGS. 8A-8K show illustrative functions and are not intended to convey limitations on locations of all components that may perform the illustrated functions. In addition, any type of nuclear fission reactor whatsoever may serve as a host environment for the systems and apparatuses shown in FIGS. 8A-8K. Referring to FIG. 8A, a functional relationship is illustrated in which at least a portion of at least one component 810 may be annealed by heat transfer, indicated by an arrow 812, from a heat source 814 that is in thermal communication (as indicated by the arrow 812) with at least the portion of the component 810. In the relationship shown in FIG. 8A, annealing can occur within a reactor pressure vessel 816. The component 810 may include any of the components discussed above. In some embodiments and given by way of non-limiting example, the component 810 can include at least one reactor core component such as at least one nuclear fission fuel assembly, reactor core cooling component, and/or reactor core structural member. When at least one nuclear fission fuel assembly is annealed, the component 810 can include cladding, a cooling component, a structural member, a thermally conductive member, and/or nuclear fission fuel material. The heat source 814 may include any of the heat sources discussed above. In some embodiments in which the heat source 814 is located within the reactor pressure vessel 816, the heart source 814 may include nuclear fission fuel material, such as that contained in nuclear fission fuel elements and/or fuel assemblies, thereby generating heat during power range operations or by generating decay heat after shutdown from power range operations. In some other embodiments, the heat source 814 may include an external heat source (that is, external to a fuel assembly), such as at least one electrical heat source, a heating fluid, and/or at least one source of residual heat, such as decay heat. As shown in FIG. 8A, the heat transfer mechanism of thermal communication between the heat source 814 and the component 810 can include reactor coolant. The reactor coolant can include liquid metal or gaseous reactor coolant, non-limiting examples of which have been described above. Referring to FIG. 8B, in some embodiments the component 810 and the heat source 814 are located in a reactor core assembly 818 within the reactor pressure vessel 816. In such an arrangement, and as indicated by the arrow 812, annealing of at least the portion of the component 810 can occur within the reactor core assembly 818. In some embodiments, annealing of at least the portion of the component 810 can be performed in an in-place location of the component 810. In some other embodiments, the component 810 can be moved, with suitable handling equipment, from its in-place location to another location within the reactor core assembly 818 where annealing can occur. Referring to FIG. 8C, in some embodiments the component 810 can be re-located from the reactor core assembly 818 with suitable handling equipment. In such an arrangement, and as indicated by the arrow 812, annealing of at least the portion of the component 810 can occur exterior of the reactor core assembly 818 but within the reactor pressure vessel 816. Referring to FIG. 8D, in some other embodiments the component 810 can be re-located, as indicated by an arrow 820, from the reactor pressure vessel 816 with suitable handling equipment and placed within suitable nuclear shielding 822 in an annealing facility, as desired. In such an arrangement, and as indicated by the arrow 812, annealing of at least the portion of the component 810 can occur exterior of the reactor pressure vessel 816. Also, in such an arrangement, the heat source 814 can be any of the heat sources described above. However, when the heat source 814 includes nuclear fission fuel material (such as when a nuclear fission fuel element or a fuel assembly is removed from the reactor pressure vessel 816 and relocated to the nuclear shielding 822 within the annealing facility) then the heat is generated via decay heat generation as opposed to power range operations. In some embodiments, annealing can occur on-site of the nuclear fission reactor. In some other embodiments, annealing can occur off-site from the nuclear fission reactor. Referring to FIG. 8E, in some embodiments heat transfer from the heat source 814 to the component 810 can be adjusted, such as with a flow adjust function 824. The flow adjust function 824 can cause reactor coolant flow to be adjusted, thereby adjusting amount of heat transferred from the heat source 814 to the component 810. In some embodiments the flow adjust function can be responsive to a control input 826. In some embodiments the control input 826 can be a mechanical input. In some other embodiments the control input 826 can be a signal input, such as an electrical signal, an optical signal, a radio-frequency signal, or the like. In some embodiments systems and apparatuses are provided for annealing at least a portion of at least one component. Referring to FIG. 8F, in some embodiments an illustrative apparatus 830 includes electrical circuitry 832, such as a control system, configured to determine an annealing temperature range for at least a portion of at least one component 810 of a nuclear fission fuel assembly of a nuclear fission reactor. A subassembly 834 is responsive to the electrical circuitry 832 and is configured to establish at least the portion of the nuclear fission fuel assembly within the annealing temperature range. In some other embodiments, the electrical circuitry 832 may be configured to determine an annealing temperature range for at least the portion of at least one component 810 of the reactor core assembly 818 of a nuclear fission reactor, wherein the annealing temperature range is higher than a predetermined operating temperature range of the reactor core assembly 818. In such an arrangement, the subassembly 834 is responsive to the electrical circuitry 832 and is configured to establish at least the portion of the nuclear fission reactor within the annealing temperature range. It will be appreciated that in some embodiments the electrical circuitry 832 may include a numerical model of material damage and/or annealing/temperature response. In some other embodiments the electrical circuitry 832 may include stored data representing annealing/temperature responses discussed above. The stored data may be determined empirically or analytically, as desired, and may be updated or supplemented with sensor data (e.g. acoustic response of steel showing degradation or restoration, or the like). In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. In a general sense, those skilled in the art will also recognize that in the various embodiments described herein the subassembly 834 can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, and electro-magnetically actuated devices, or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment), and any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, as well as other systems such as motorized transport systems, factory automation systems, security systems, and communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. The apparatus 830 may include a sensing system 836 that provides sensed data to the electrical circuitry 832. In some embodiments the sensing system 836 may be configured to sense conditions, such as temperature, pressure, reactor coolant flow rate, or the like, of the region of the reactor core assembly 818 containing the component 810. As such, the sensing system 836 may include sensors such as temperature sensors, pressure sensors, flow sensors, or the like. In some other embodiments the sensing system 836 may be further configured to test material properties of at least a portion of the component 810 during annealing. As discussed above, in some embodiments the heat source 814 can include an external heat source, such as at least one electrical heat source and/or a heating fluid, and/or at least one source of residual heat, such as decay heat. In some embodiments the subassembly 834 can be further configured to adjust operational parameters of the nuclear fission reactor to establish operating conditions of a region of the nuclear fission reactor containing the at least one component within the determined annealing temperature range for a period of time selected to produce annealing of at least the portion of the at least one component. Referring now to FIG. 8G, in some embodiments the subassembly 834 can include a reactor coolant system. Given by way of non-limiting example and referring to FIG. 8H, the reactor coolant system can include at least one reactor coolant pump 838. In some embodiments the at least one reactor coolant pump 838 can be responsive to a reactor coolant pump controller 840, such as for starting, stopping, and/or changing speeds of the reactor coolant pump 838. In some other embodiments and referring to FIG. 8I, the reactor coolant system can include at least one flow adjustment device 842, such as a valve like an isolation valve, a throttle valve, or the like. The flow adjustment device 842 can be a mechanical device with mechanical actuation, a mechanical device with electrical actuation, or an electrical device that can electrically control flow of electrically-conductive liquid reactor coolant, such as liquid metals. In some embodiments the at least one flow adjustment device 842 can be responsive to a flow adjustment device controller 844. Referring now to FIG. 8J, in some other embodiments the subassembly 834 can include a reactor control system. Given by way of non-limiting example, the reactor control system can control reactivity within the reactor core assembly 818, such as by inserting or withdrawing control rods or otherwise inserting or removing neutron absorbing material or the like. Referring to FIG. 8K, in some embodiments the subassembly 834 can include a pressurizer. Given by way of non-limiting example, the pressurizer can control pressure by turning on or turning off pressurizer heaters, as desired. One skilled in the art will recognize that the herein described components (e.g., blocks), devices, and objects and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are within the skill of those in the art. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar herein is also intended to be representative of its class, and the non-inclusion of such specific components (e.g., blocks), devices, and objects herein should not be taken as indicating that limitation is desired. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. Furthermore, it is to be understood that the invention is defined by the appended claims. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to inventions containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that virtually any disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms. For example, the phrase “A or B” will be understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. With respect to context, even terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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claims | 1. A diagnostic system for a digital rod position indication (DRPI) system of a nuclear power plant to monitor in real time DRPI signals generated by a plurality of detector coils of the DRPI system while the nuclear power plant is operating, the diagnostic system comprising:a digital diagnostic unit connected in parallel between a DRPI display cabinet and redundant pair of DRPI A and DRPI B data cabinets, the digital diagnostic unit having:inputs configured to receive DRPI signals communicated between the DRPI display cabinet and the DRPI A and B data cabinets;a coil diagnostic unit configured to receive voltage signals from each one of the detector coils;a plurality of data acquisition modules configured to receive digital rod position signals for each detector coil from the DRPI A and B data cabinets;at least one address input/output module configured to drive rod addresses to the DRPI A and B data cabinets; anda gate array module configured to acquire the DRPI signals from the data acquisition and address input/output modules, the gate array module having an interface connected with a controller configured to monitor the digital rod position signals from the DRPI A and B data cabinets for each coil and identify mismatches between a DRPI A rod position of the DRPI A data cabinet and a DRPI B rod position of the DRPI B data cabinet for each coil while the nuclear power plant is operating. 2. The system of claim 1, wherein the digital rod position signals include rod address signals and rod position data signals, and wherein the rod position errors are determined based on signal level variation and/or signal timing variation of the rod address signals and the rod position data signals. 3. The diagnostic system of claim 1, wherein the digital diagnostic unit further detects parity bit errors in the digital rod position signals between the DRPI display cabinet and the DRPI data cabinet. 4. The diagnostic system of claim 3, wherein the digital diagnostic unit stores measured voltages of the digital rod position signals when a rod position error or a parity bit error is detected. 5. The diagnostic system of claim 3, wherein the digital diagnostic unit monitors at least one of the signal level variation, the signal timing variation, and the parity bit error of the digital rod position signals to identify the component that is the source of the error. 6. The diagnostic system of claim 5, wherein the digital diagnostic unit monitors variations in the digital rod position signals while the nuclear power plant is operating. 7. The diagnostic system of claim 5, wherein the digital diagnostic unit monitors variations in the digital rod position signals to identify errors of a particular card, cable, or control rod. 8. The diagnostic system of claim 1, wherein the diagnostic unit monitors Gray code rod drop signals of the digital rod position signals. |
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abstract | An anti-proliferation technique is disclosed to reduce the likelihood of nuclear proliferation due to the use fissionable fuel salts. The technique includes doping the fuel salt with one or more elements (referred to herein as activation dopants) that, upon exposure to neutrons such as would occur in the fuel salt when a reactor is in operation, undergo a nuclear reaction to, directly or indirectly, form highly active “protecting isotopes” (of the same element as the activation dopant or a different element). A sufficient mass of activation dopants is used so that the Figure of Merit (FOM) of the fuel salt is decreased to below 1.0 within some target number of days of fission. This allows the FOM of the fuel salt to be controlled so that the fuel becomes too dangerous to handle before to the creation of a significant amount of weaponizable isotopes. |
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claims | 1. A system for monitoring a machine, said system comprising:a) a machine monitoring device connected to said machine, said machine monitoring device comprising:i) input means connected to said machine for receiving inputs from said machine;ii) an engine connected to said input means for performing transformations on said inputs, wherein said transformations apply a mathematical operation or a logical operation on said inputs to generate outputs;iii) a database system connected to said engine to store said outputs;output means connected to said engine for transmitting output signals from said machine monitoring device to said machine and another machine, wherein said engine performs additional transformations on said inputs, said additional transformations apply a mathematical operation or logical operation to said inputs to generate additional outputs, and said engine generates said output signals based on said additional outputs; andiv) report generating means connected to said database system for generating reports based on said outputs; andb) a client computing device connected to said machine monitoring device by a communications network for receiving said reports to allow a user to monitor said machine from said client computing device. 2. The system of claim 1, wherein said report generating means comprises a web server for transmitting said reports in the form of web pages to said client computing device. 3. The system of claim 2, wherein said web server further generates web page user interfaces from which said user can configure said machine monitoring device, said transformations, or said reports from said client computing device. 4. The system of claim 3, wherein said web server comprises a configuration Common Gateway Interface (CGI) module for generating web page user interfaces from which a user may enter or view configuration information from said client computing device. 5. The system of claim 2 wherein said web server comprises a reports Common Gateway Interface (CGI) module for generating web page user interfaces from which said user may request said reports and enter parameters required for said reports from said client computing device. 6. The system of claim 2 wherein said machine monitoring device is a designated machine monitoring device connected over said communications network to a plurality of machine monitoring devices, said web server of said designated machine monitoring device generating a web page user interface comprising a list of said plurality of machine monitoring devices and permitting a user to select reports from one or more of said plurality of machine monitoring devices for viewing from said client computing device. 7. The system of claim 1 wherein said report generating means comprises a reporter module for automatically generating and automatically transmitting said reports to said client computing device. 8. The system of claim 1 wherein said input means comprises a digital input connector for receiving digital inputs from said machine. 9. The system of claim 1 wherein said input means comprises an analog input connector for receiving analog inputs from said machine. 10. The system of claim 1 wherein said output means comprises a digital output connector and said output signals comprise digital output signals. 11. The system of claim 1, further comprising at least one serial port for providing serial communications between said machine and said machine monitoring device. 12. The system of claim 1 further comprising an Ethernet port for providing Ethernet communications between said machine and said machine monitoring device. 13. The system of claim 1, wherein said machine monitoring device further comprises a configuration interface module for reading and writing configuration information, said configuration information being initially entered when said machine monitoring device is configured. 14. The system of claim 13 wherein said configuration interface module maintains usernames, access and modification rights for said configuration information, and passwords for said user as part of said configuration information. 15. The system of claim 1, wherein said machine monitoring device further comprises drivers connected to said input means for converting said inputs into values associated with variables by said engine, said engine performing said transformations on said values to generate additional values for said reports, wherein said additional values are associated with report variables, said outputs are comprised of said additional values, and said reports are generated from said additional values associated with said report variables. 16. A method for monitoring a machine comprising the steps of:a) monitoring inputs from said machine by means of a machine monitoring device connected to said machine;b) performing transformations on said inputs, wherein said transformations apply a mathematical operation or a logical operation on said inputs to generate outputs, wherein said step of performing transformations is performed when a change in said inputs is detected during said monitoring, said transformations being performed by an engine within said machine monitoring device;c) generating reports on said machine monitoring device, wherein said reports are generated by said machine monitoring device from said outputs, said step of performing transformations further comprising the step of storing changes to said outputs for said reports resulting from said transformations in a database system within said machine monitoring device; andd) transmitting said reports from said machine monitoring device to a client computing device on a communications network. 17. The method of claim 16, further comprising the steps of generating e-mail notifications and e-mail notification escalations and transmitting said e-mail notifications and said e-mail notification escalations to said client computing device, said e-mail notifications and said e-mail notification escalations being generated by said machine monitoring device based on said inputs and said outputs. 18. The method of claim 16, wherein said step of generating reports further comprises the steps of:i) automatically generating a query by a reporter module within said machine monitoring device at configured time intervals or shifts;ii) processing said query and transmitting data resulting from said query back to said reporter module, wherein said processing is effected by said machine monitoring device; andiii) generating a report from said data to be transmitted automatically to said client computing device, said generating of said report being effected by said reporter module. 19. The method of claim 16, wherein said step of generating reports further comprises the steps of:i) entering the Internet Protocol address of said machine monitoring device to cause generation of a menu of available reports;ii) selecting a desired report from said menu;iii) generating a query for said desired report;iv) processing said query and transmitting data resulting from said query to a reports Common Gateway Interface (CGI) module within said machine monitoring device, wherein said processing is effected by said machine monitoring device; and generating said desired report from said data, said generating of said desired report being effected by said reports CGI module. 20. A method for configuring a machine monitoring device connected to a communications network for monitoring a machine connected to said machine monitoring device, wherein said configuring comprises the steps of:a) determining desired reports and desired output signals from said machine monitoring device to said machine;b) identifying required inputs from said machine and required outputs for generating said desired reports and said desired output signals;c) defining transformations to be performed on said required inputs, said transformations applying a mathematical or logical operation to said required inputs to generate said required outputs;d) associating said required inputs and said required outputs with variables and report variables, wherein said variables store values for use in said transformations and said required outputs for output signals and said report variables store values for said required outputs for transmission in said desired reports;e) configuring said desired reports using said report variables and said desired output signals using said variables; andf) storing configuration information entered within said machine monitoring device. 21. The method of claim 20, further comprising the step of configuring shifts and time intervals for use in said generating said desired reports. |
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abstract | A system for storing and/or transporting high level radioactive waste, and a method of manufacturing the same. In one aspect, the invention is a ventilated vertical overpack (“VVO”) having specially designed inlet ducts that refract radiation back into the storage cavity. A clear line-of-sight does not exist through the inlet ducts and, thus, the canister can be supported on the floor of the VVO. Also disclosed is a method of manufacturing a variable height VVO that falls within a regulatory license previously obtained for a shorter and taller version of the VVO. |
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summary | ||
claims | 1. A nuclear reactor cooling system, comprising:a cooling water tank disposed above a containment and containing cooling water therein;a spray header connected to the cooling water tank through a first communicating pipe; andan air tank disposed inside the containment; wherein the spray header is disposed outside the containment and used for spraying the cooling water to an outer wall of the containment; the cooling water tank is a closed container, and the air tank is connected to a top portion of the cooling water tank through a second communicating pipe, and the nuclear reactor cooling system further comprises a bell-shaped shield around an exterior of the containment. 2. The nuclear reactor cooling system according to claim 1, wherein the shield is used for covering the containment and setting the containment in its interior, the spray header is disposed in a space between an inner wall of the shield and the outer wall the containment, and the cooling water tank is disposed on a top portion of the shield. 3. The nuclear reactor cooling system according to claim 2, further comprising an exhaust hole disposed on the top portion of the shield. 4. The nuclear reactor cooling system according to claim 3, wherein the cooling water tank has a shape of annulus, the axis of the cooling water tank is collinear with the axis of the shield. 5. The nuclear reactor cooling system according to claim 3, further comprising a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. 6. The nuclear reactor cooling system according to claim 4, further comprising a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. 7. The nuclear reactor cooling system according to claim 2, further comprising a cooling water outlet disposed on a bottom portion of the shield. 8. The nuclear reactor cooling system according to claim 7, further comprising a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. 9. The nuclear reactor cooling system according to claim 2, wherein the spray header is axisymmetrically disposed above the containment. 10. The nuclear reactor cooling system according to claim 9, further comprising a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. 11. The nuclear reactor cooling system according to claim 2, further comprising a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. 12. The nuclear reactor cooling system according to claim 1, further comprising a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. |
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