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	claims	1. A nuclear reactor comprising a reactor vessel, nuclear fuel, a plurality of control rods, reactor coolant, at least one steam generator, a first pipe penetrating a wall of said reactor vessel and connected to a first inlet of said steam generator and at least one reactor coolant pump including a motor, said pump connected to a second inlet of said steam generator, wherein said nuclear fuel, said control rods, said reactor coolant, said steam generator and said reactor coolant pump including said motor are all located inside said reactor vessel, said pump comprising: a housing; an annular stator disposed in said housing and having a generally cylindrical passage extending therethrough and a plurality of stator windings, said stator having energizing means for electrically connecting a source of electrical power to said plurality of stator windings; an axial flow impeller assembly rotatably mounted in said generally cylindrical passage in said stator, said impeller assembly comprising an impeller and a sealed rotor mounted around the perimeter of said impeller and positioned inside said stator to form an electric motor, the operation of which rotates said impeller to produce a pressurized flow of fluid through said generally cylindrical passage; at least one radial bearing mounted between said impeller assembly and said housing; a thrust bearing mounted between said impeller assembly and said housing and located downstream from said rotor; a shaft centrally positioned in said generally cylindrical passage in said housing and secured to said rotor; said impeller assembly rotatably supported by said shaft; and insulation material disposed within said stator, said insulation material comprising a plurality of solid pieces of mica, glass and ceramic insulation tightly packed within said stator. 2. The reactor of claim 1 , further comprising a second pipe penetrating a wall of said reactor vessel, wherein said second pipe is connected to an outlet of said steam generator. claim 1 3. The reactor of claim 2 , wherein six steam generators are located inside said reactor vessel. claim 2 4. The reactor of claim 3 , wherein six reactor coolant pumps are located inside said reactor vessel and are each connected to an inlet of one of said six steam generators. claim 3 5. The reactor of claim 4 , wherein said thrust bearing comprises a bearing runner and a pair of bearing pads. claim 4 6. A nuclear reactor comprising a reactor vessel, nuclear fuel, a plurality of control rods, reactor coolant, at least one steam generator, and at least one reactor coolant pump including a motor, said pump connected to the inlet of said steam generator, wherein said nuclear fuel, said control rods, said reactor coolant, said steam generator and said reactor coolant pump including said motor are all located inside said reactor vessel; said pump comprising: a housing; an annular stator disposed in said housing and having a generally cylindrical passage extending therethrough and a plurality of stator windings, said stator having energizing means for electrically connecting a source of electrical power to said plurality of stator windings; an axial flow impeller assembly rotatably mounted in said generally cylindrical passage in said stator, said impeller assembly comprising an impeller and a sealed rotor mounted around the perimeter of said impeller and positioned inside said stator to form an electric motor, the operation of which rotates said impeller to produce a pressurized flow of fluid through said generally cylindrical passage; at least one radial bearing mounted between said impeller assembly and said housing; a thrust bearing mounted between said impeller assembly and said housing and located downstream from said rotor; a shaft centrally positioned in said generally cylindrical passage in said housing and secured to said rotor; said impeller assembly rotatably supported by said shaft; insulation material disposed within said stator, said insulation material comprising a plurality of solid pieces of insulation tightly packed within said stator; and wherein said energizing means comprises a terminal gland capable of withstanding a pressure of approximately 2500 psi connected to an electrical power source, said terminal gland comprising a stainless steel body, said body attached to an alumina ceramic insulator by a first glass preform, said ceramic insulator attached to a terminal gland stud by a second glass preform.
	claims	1. A measurement device for determining a boron concentration in a nuclear power station, comprising: an emitter to be disposed on a boron-carrying component with a coolable space with a cooling duct through which a coolant flows formed in between said emitter and said boron-carrying component; and a receiver to be disposed on said boron-carrying component with a coolable space with a cooling duct through which said coolant flows, formed in between said receiver and said boron-carrying component; spacers formed of a temperature-resistant and expansion-resistant material, said spacers maintaining a length of a measurement path between said emitter and said receiver substantially constant even in the event of a change in ambient conditions. 2. The measurement device according to claim 1 , wherein said emitter and said receiver are held on said boron-carrying component by at least one fastener so as to require no structural measures or interventions on said boron-carrying component. claim 1 3. The measurement device according to claim 1 , wherein said coolant is air. claim 1 4. The measurement device according to claim 1 , wherein said coolant has a predetermined temperature. claim 1 5. The measurement device according to claim 1 , which further comprises an additional insulation layer in said coolable space. claim 1 6. The measurement device according to claim 1 , wherein each of said emitter and said receiver are disposed in a respective chamber. claim 1 7. The measurement device according to claim 1 , wherein said emitter is a neutron source. claim 1 8. The measurement device according to claim 1 , wherein said receiver is at least one counter tube. claim 1 9. The measurement device according to claim 1 , which further comprises a shield respectively enclosing said emitter and said receiver. claim 1 10. The measurement device according to claim 9 , wherein said shield includes a first layer of an absorbing moderator. claim 9 11. The measurement device according to claim 10 , wherein said shield includes a second layer of neutron-absorbing material. claim 10 12. The measurement device according to claim 11 , wherein said shield includes a third layer of austenitic material. claim 11 13. The measurement device according to claim 9 , wherein said shield includes a layer of austenitic material. claim 9 14. The measurement device according to claim 1 , which further comprises a casing housing said emitter and said receiver. claim 1 15. The measurement device according to claim 14 , wherein said casing is formed in two parts. claim 14 16. The measurement device according to claim 1 , which comprises at least one fastener for fastening said emitter and said receiver to the component. claim 1 17. The measurement device according to claim 1 , which further comprises an evaluation device connected to said receiver. claim 1 18. The measurement device according to claim 1 , wherein said emitter and said receiver are disposed substantially opposite one another on the component. claim 1 19. The measurement device according to claim 1 , wherein said emitter and said receiver are disposed on the component such that a measurement signal received by said receiver is substantially a reflected measurement signal. claim 1
	description	This application claims the benefit under 35 U.S.C. § 119 of Chinese Application No. 201810041590.0, filed Jan. 16, 2018, which hereby incorporated in its entirety. The present invention relates to the technical field of the reactor control rod drive mechanism, and more particularly to an electromagnetic coil assembly for a control rod driving mechanism and a method of manufacturing the same. A reactor is a core portion of the nuclear power station. The control rod driving mechanism (CRDM) on the reactor can raise, insert or maintain the position of the control rod in the reactor core to control the fission rate of the reactor, realize starting and stopping the reactor as well as regulating the reactor power, and rapidly insert the control rod (i.e., scram) under accident conditions, which causes the reactor to be scrammed in a short time to ensure the safe operation of the nuclear power station. CRDM cannot realize aforementioned functions without the electromagnetic coil assembly which must have features such as stability, reliability, heat resistance, good insulation performance, radiation resistance and long service life due to the special working environment and extreme importance of the functions thereof. In the prior art, regarding to the designs of the service life and reliability of the electromagnetic coil, the people skilled in the art made various explorations on the structure of the electromagnetic coil assembly as well as the material of the parts thereof, for example, the invention application No. CN103329209A and the invention application No. 201410500311.4. Further studies on the service life and stability of the electromagnetic coil assembly can definitely further improve the reliability of the control rod driving mechanism. Further studies on the problems proposed above regarding to the service life and stability of the electromagnetic coil assembly can definitely further improve the reliability of the control rod driving mechanism. The present invention provides an electromagnetic coil assembly for a control rod driving mechanism and a method of manufacturing the same. The coil assembly provided in the present invention or the coil assembly obtained by the method can remarkably reduce the temperature inside the coil, thereby improving the reliability of the CRDM electromagnetic coil assembly and prolonging the service life of the CRDM electromagnetic coil assembly. The electromagnetic coil assembly for the control rod driving mechanism provided in the present invention solves the problems by following technical points: an electromagnetic coil assembly for a control rod driving mechanism, comprising one or more coils and a yoke for embedding the one or more coils, wherein one or more damascene holes are disposed on the yoke, the one or more coils are installed in the one or more damascene holes, the yoke comprises at least one first yokes and at least one second yokes, and the one or more damascene holes are disposed on the at least one first yokes; the at least one first yokes are connected with the at least one second yokes, and a through hole for cooperating with a sealing shell assembly is disposed on the at least one second yokes; and among the at least one first yokes and the at least one second yokes, a thermal conductivity of the at least one first yokes is stronger than a thermal conductivity of the at least one second yokes. The designs relating to the reliability and service life of the electromagnetic coil assembly for the control rod driving mechanism, in the prior art, include the technical solutions such as the use of the high-temperature resistant non-metallic material as the coil framework, the use of the metallic material having a better temperature tolerance as the coil framework, and reinforcement of the heat dispersion of the coils. In the present technical solution, the first yokes and the second yokes are respectively used as parts of the yoke. Meanwhile, the coils are all installed on the first yokes through the damascene holes disposed on the first yokes. Therefore, when such assembly is cooperated with the sealing shell assembly, the center holes on the coils and the through holes of the second yokes are served as the passage for the sealing shell assembly to pass through such assembly. At the same time, in the prior art, since the operating temperature of the CRDM pressure resistant sealing shell assembly located in the center hole of such assembly is about 300° C., the heat transferred from the sealing shell assembly to the coils of such assembly by heat conduction is particularly large. The thermal conductivity of the first yokes is stronger than the thermal conductivity of the second yokes, so that the heat transferred from the second yokes to the first yokes and the coils due to the transfer mode of heat conduction can be effectively reduced, and the heat generated during working process of the coils can be dissipated by the first yokes which have the stronger thermal conductivity. Consequently, the design of the yoke structure provided by the present invention can effectively discharge the heat generated by the coils through the first yokes, while reduce the heat transferred from the sealing shell assembly to the coils, which allow the coils to operate at a relatively low temperature, thereby significantly reducing the temperature inside the coils, improving the reliability of CRDM electromagnetic coil assembly, and extending the service life of the CRDM electromagnetic coil assembly. Preferably, as specific implementation forms of the first yokes and the second yokes, the material of the first yokes may be pure iron, electromagnetic pure iron and the like, and a plurality of heat dissipating fins are processed on the outer surface of the first yokes to increase the heat dissipating area of the first yokes; the material of the second yokes may be cast iron, martensitic stainless steel, permeability alloy, soft magnetic ferrite, soft magnetic amorphous alloy and the like. The people skilled in the art should appreciate that the thermal conductivity described above refers to the heat conduction capability of the material, and can be represented by the heat conductivity coefficient of the material. Preferably, the projection of the coils toward the end surface of each of the second yokes falls outside the through hole. The projection of the coils toward the end surface of each of the second yokes is defined to fall outside the through hole, that is to say that the aforementioned projection of the coils is situated outside the through hole, which means, in such assembly, the inner diameter of the passage at the positions of the coils is larger than the inner diameter of the passage at the positions of the second yokes. Therefore, when the assembly is cooperated with the sealing shell assembly, the engagement positions of the assembly and the sealing shell assembly are located at the positions of the second yokes, preventing the coils from directly contacting the sealing shell assembly. Since the structure features provided in the present invention avoids the direct contact between the coils and the sealing shell assembly, there is no direct heat conduction between the coils and the sealing shell assembly. When the sealing shell assembly is not in direct contact with the coils, the transfer mode of the heat from the sealing shell to the coils becomes to thermal radiation. To further reduce the heat transferred from the sealing shell to the coils through thermal radiation, preferably, the inner hole surface of the coils, namely the hole wall of the center hole of the coils is polished, for example, into a mirror surface to improve the brightness of the inner hole surface of the coils, which can effectively reflect the thermal radiation generated by the sealing shell, further reducing the heat of the coils. A further technical solution is that: as specific implementation forms of the damascene holes and the through hole, either the one or more damascene holes or the through hole is a round hole, and an axis of the one or more damascene holes is collinear with an axis of the through hole; and each of the one or more coil comprises an inner frame and a coil winding wound around the inner frame, the inner frame has a cylindrical shape, an axis of the inner frame is collinear with the axis of the one or more damascene holes, and a diameter of a center hole of the inner frame is larger than a diameter of the through hole. An amount of the at least one first yokes exceeds an amount of the at least one second yokes by one. Two adjacent first yokes of the at least one first yokes are connected by one of the at least one second yokes. The one or more damascene holes are disposed on each of the at least one first yokes, and the one or more coils are installed in each of the one or more damascene holes. In such technical solution, the first yokes are in an interval distribution, allowing better external heat dissipation environment of the single first yoke. In such technical solution, the amount of the first yokes exceeds the amount of the second yokes by one, which actually aims to define that one of the second yokes is disposed between two adjacent first yokes. It is understood that the one of the second yokes used herein also refers to a second yoke disposed between two first yokes. If a second yoke is connected to the outer side of the first yoke located at the end portion of the assembly to form an additional connection portion between the assembly and a containment assembly, the additional second yoke should not be counted in the count range of the above number difference. As described above, since the materials of the first yokes and the second yokes are different, the first yokes and the second yokes are connected by the bolt which is a detachable connection form not affecting the properties of the first yokes and second yokes during connection. Also, a sealing ring is disposed on the connecting surface of the first yokes and the second yokes. The sealing ring is used for sealing the inner side and the outer side of the assembly, that is to say that the sealing ring provides a radial seal, having characteristics of improving the moisture resistance of the coils of the assembly. Preferably, as the electromagnetic coil assembly for the control rod driving mechanism generally contains three coils, the end portions of the assembly are both the second yokes. Meanwhile, each of the second yokes at the end portions is also provided with an annular sealing groove for installing the sealing ring to realize the axial sealing of the assembly. Such structure form also facilitates the increase of the contact area of the assembly and the sealing shell assembly, as well as the reliability of the connection of the assembly and the sealing shell assembly. As an implementation that the coils can be completely enclosed in the first yokes, the depth of each of the one or more damascene holes is equal to or greater than a length of each of the one or more coils. A space between a hole wall of each of the one or more damascene holes and each of the one or more coils is filled with a potting layer. Two end faces of each of the one or more coils are both located between two end faces of each of the one or more damascene holes, or the two end faces of each of the one or more coils are respectively coincident with the two end faces of each of the one or more damascene holes. In such technical solution, the first yokes are able to protect the coils and act as the outer frames (i.e. outer shells) of the coils, which facilitate simplifying the structure design of the coils. The above potting layer not only facilitates the structure stability of the assembly, but also makes the seismic performance of the assembly stronger. Meanwhile, the above potting layer is beneficial to heat conduction between the coils and the first yokes. A wire hole is disposed on the at least one first yokes and the at least one second yokes. The wire hole disposed on the at least one first yokes and the one or more damascene holes are holes relatively independent from each other. The wire hole disposed on the at least one second yokes and the through hole are holes relatively independent from each other. Further definition that the wire hole and the damascene holes are relatively independent holes aims to separate the wire hole from the damascene holes, thereby effectively preventing the potting material from blocking the wire hole when the potting layer is disposed, which may affects the treading operation during the subsequent fitting of components. Moreover, separating the wire hole from the damascene holes can also avoid waste of the potting material or poor potting effect caused by the loss of the potting material from the wire hole when the potting process is carried out to obtain the potting layer. People skilled in the art should appreciate that the wire hole is not disposed on all of the first yokes and the second yokes. For example, according to the practical needs, usually the second yoke at one end of the electromagnetic coil assembly and the first yoke connected thereto are not provided with the wire hole. In order to improve the heat dispersion of an end portion of each of the one or more coils, the end portion of each of the one or more coils is also covered with the potting layer. At the same time, after the above potting layer is disposed, the length of the combination of the potting layer and the coils is controllable. Therefore, the length of the combination can be slightly larger than the depth of the damascene holes, so that when two adjacent first yokes are connected to one second yoke, the potting layer of the end portions may be compressed to seal the first yokes and the second yoke at the end portions of the coils. To improve the thermal conductivity of the potting layer, the potting layer is further embedded with thermally conductive insulating particles, wherein the thermally conductive insulating particles can use quartz sands, flint silica sands, PPS polyphenylene sulfide, PA46 nylon and the like. Furthermore, the present invention also provides a method of manufacturing an electromagnetic coil assembly for a control rod driving mechanism, wherein the method is used to manufacture any one of the aforementioned electromagnetic coil assemblies for the control rod driving mechanism. The method comprises following steps carried out in sequence: S1. winding the coil winding around an inner frame to obtain the one or more coils, and embedding the obtained one or more coils in the one or more damascene holes of the at least one first yokes; S2. integrally potting an assembly consisting of the one or more coils and the at least one first yokes to obtain a potting layer in a space between a hole wall of each of the one or more damascene holes and each of the one or more coils; and S3. completing an assembly of the at least one first yokes and the at least one second yokes. Such technical solution is the processing method of the electromagnetic coil assembly for the control rod driving mechanism provided above. Preferably, a specific form of the assembly in S3 uses a bolted connection. A sealing ring is disposed on a connecting surface of the at least one first yokes and the at least one second yokes. The present invention has the following beneficial effect: The structure of the electromagnetic coil assembly provided in the present invention and the product obtained by the manufacturing method provided in the present invention have the following features: Since the structure features provided in such technical solution avoids the direct contact between the coils and the sealing shell assembly, there is no direct heat conduction between the coils and the sealing shell assembly. Meanwhile, the thermal conductivity of the first yokes is stronger than the thermal conductivity of the second yokes, so that the heat transferred from the second yokes to the first yokes and the coils due to heat conduction can be effectively reduced, and the heat generated during working process of the coils can be dissipated by the first yokes which have the stronger thermal conductivity. As a result, the design of the yoke structure provided by the technical solution can effectively discharge the heat generated by the coils through the first yokes, while reduce the heat transferred from the sealing shell assembly to the coils, which allow the coils to operate at a relatively low temperature, thereby significantly reducing the temperature inside the coils, improving the reliability of CRDM electromagnetic coil assembly, and extending the service life of the CRDM electromagnetic coil assembly. Annotations in the figures and names of the corresponding parts are: 1. first yoke, 11. damascene hole, 12. wiring hole, 2. potting layer, 3. coil, 4. inner frame, 5. second yoke, 6. bolt, 7. sealing ring. The present invention is further described in detail below with reference to the embodiments, but it is understood that the structure of the present invention is not limited to the following embodiments. As shown in FIG. 1 to FIG. 3, an electromagnetic coil assembly for a control rod driving mechanism, comprising one or more coils 3 and a yoke for embedding the one or more coils 3, wherein one or more damascene holes 11 are disposed on the yoke, the one or more coils 3 are installed in the one or more damascene holes 11, the yoke comprises at least one first yokes 1 and at least one second yokes 5, and the one or more damascene holes 11 are disposed on the at least one first yokes 1; the at least one first yokes 1 are connected with the at least one second yokes 5, and a through hole for cooperating with a sealing shell assembly is disposed on the at least one second yokes 5, a projection of the one or more coils 3 toward the end surface of each of the second yokes 5 falls outside the through hole; among the at least one first yokes 1 and the at least one second yokes 5, a thermal conductivity of the at least one first yokes 1 is stronger than a thermal conductivity of the at least one second yokes 5. The designs relating to the reliability and service life of the electromagnetic coil assembly for the control rod driving mechanism, in the prior art, include the technical solutions such as the use of the high-temperature resistant non-metallic material as the coil framework, the use of the metallic material having a better temperature tolerance as the coil framework, and reinforcement of the heat dispersion of the coils 3. In the present technical solution, the first yokes 1 and the second yokes 5 are respectively used as parts of the yoke. Meanwhile, the coils 3 are all installed on the first yokes 1 through the damascene holes 11 disposed on the first yokes 1. Therefore, when such assembly is cooperated with the sealing shell assembly, the center holes on the coils 3 and the through holes of the second yokes 5 are served as the passage for the sealing shell assembly to pass through such assembly. In the technical solution, the through holes for cooperating with the sealing shell assembly are disposed on the second yokes 5. The projection of the coils 3 toward the end surface of each of the second yokes 5 falls outside the through hole, that is to say that the aforementioned projection of the coils 3 is situated outside the through hole, which means, in such assembly, the inner diameter of the passage at the positions of the coils 3 is larger than the inner diameter of the passage at the positions of the second yokes 5. Therefore, when the assembly is cooperated with the sealing shell assembly, the engagement positions of the assembly and the sealing shell assembly are located at the positions of the second yokes 5, preventing the coils 3 from directly contacting the sealing shell assembly. At the same time, in the prior art, since the operating temperature of the CRDM pressure resistant sealing shell assembly located in the center hole of such assembly is about 300° C., the heat transferred from the sealing shell assembly to the coils 3 of such assembly by heat conduction is particularly large. However, as the structure features provided in such technical solution avoids the direct contact between the coils 3 and the sealing shell assembly, there is no direct heat conduction between the coils 3 and the sealing shell assembly. In addition, the thermal conductivity of the first yokes 1 is stronger than the thermal conductivity of the second yokes 5, so that the heat transferred from the second yokes 5 to the first yokes 1 and the coils 3 due to heat conduction can be effectively reduced, and the heat generated during working process of the coils 3 can be dissipated by the first yokes 1 which have the stronger thermal conductivity. Consequently, the design of the yoke structure provided by the technical solution can effectively discharge the heat generated by the coils 3 through the first yokes 1, while reduce the heat transferred from the sealing shell assembly to the coils 3, which allow the coils 3 to operate at a relatively low temperature, thereby significantly reducing the temperature inside the coils 3, improving the reliability of CRDM electromagnetic coil assembly, and extending the service life of the CRDM electromagnetic coil assembly. Preferably, as specific implementation forms of the first yokes 1 and the second yokes 5, the material of the first yokes 1 may be pure iron, electromagnetic pure iron and the like, and a plurality of heat dissipating fins are processed on the outer surface of the first yokes 1 to increase the heat dissipating area of the first yokes 1; the material of the second yokes 5 may be cast iron, martensitic stainless steel, permeability alloy, soft magnetic ferrite, soft magnetic amorphous alloy and the like. As shown in FIG. 1 to FIG. 3, embodiment 2 is further defined on the basis of embodiment 1: as specific implementation forms of the damascene holes 11 and the through hole, either the one or more damascene holes 11 or the through hole is a round hole, and an axis of the one or more damascene holes 11 is collinear with an axis of the through hole; and each of the one or more coil 3 comprises an inner frame 4 and a coil winding wound around the inner frame 4, the inner frame 4 has a cylindrical shape, an axis of the inner frame 4 is collinear with the axis of the one or more damascene holes 11, and a diameter of a center hole of the inner frame 4 is larger than a diameter of the through hole. An amount of the at least one first yokes 1 exceeds an amount of the at least one second yokes 5 by one. Two adjacent first yokes 1 of the at least one first yokes 1 are connected by one of the at least one second yokes 5. The one or more damascene holes 11 are disposed on each of the at least one first yokes 1, and the one or more coils 3 are installed in each of the one or more damascene holes 11. In such technical solution, the first yokes 1 are in an interval distribution, allowing better external heat dissipation environment of the single first yoke. As described above, since the materials of the first yokes 1 and the second yokes 5 are different, the first yokes 1 and the second yokes 5 are connected by the bolt 6 which is a detachable connection form not affecting the properties of the first yokes 1 and second yokes 5 during connection. Also, a sealing ring 7 is disposed on the connecting surface of the first yokes 1 and the second yokes 5. The sealing ring 7 is used for sealing the inner side and the outer side of the assembly, that is to say that the sealing ring 7 provides a radial seal, having characteristics of improving the moisture resistance of the coils 3 of the assembly. Preferably, as the electromagnetic coil assembly for the control rod driving mechanism generally contains three coils 3, the end portions of the assembly are both the second yokes 5. Meanwhile, each of the second yokes 5 at the end portions is also provided with an annular sealing groove for installing the sealing ring to realize the axial sealing of the assembly. Such structure form also facilitates the increase of the contact area of the assembly and the sealing shell assembly, as well as the reliability of the connection of the assembly and the sealing shell assembly. As an implementation that the coils 3 can be completely enclosed in the first yokes 1, the depth of each of the one or more damascene holes 11 is equal to or greater than a length of each of the one or more coils 3. A space between a hole wall of each of the one or more damascene holes 11 and each of the one or more coils 3 is filled with a potting layer 2. Two end faces of each of the one or more coils 3 are both located between two end faces of each of the one or more damascene holes 11, or the two end faces of each of the one or more coils 3 are respectively coincident with the two end faces of each of the one or more damascene holes 11. In such technical solution, the first yokes 1 are able to protect the coils 3 and act as the outer frames of the coils 3, which facilitate simplifying the structure design of the coils 3. The above potting layer 2 not only facilitates the structure stability of the assembly, but also makes the seismic performance of the assembly stronger. Meanwhile, the above potting layer 2 is beneficial to heat conduction between the coils 3 and the first yokes 1. A wire hole 12 is disposed on the at least one first yokes 1 and the at least one second yokes 5. The wire hole 12 disposed on the at least one first yokes 1 and the one or more damascene holes 11 are holes relatively independent from each other. The wire hole 12 disposed on the at least one second yokes 5 and the through hole are holes relatively independent from each other. Further definition that the wire hole 12 and the damascene holes 11 are relatively independent holes aims to separate the wire hole 12 from the damascene holes 11, thereby effectively preventing the potting material from blocking the wire hole 12 when the potting layer 2 is disposed, which may affects the treading operation during the subsequent fitting of components. Moreover, separating the wire hole from the damascene holes can also avoid waste of the potting material or poor potting effect caused by the loss of the potting material from the wire hole 12 when the potting process is carried out to obtain the potting layer 2. In order to improve the heat dispersion of an end portion of each of the one or more coils 3, the end portion of each of the one or more coils 3 is also covered with the potting layer 2. At the same time, after the above potting layer 2 is disposed, the length of the combination of the potting layer 2 and the coils 3 is controllable. Therefore, the length of the combination can be slightly larger than the depth of the damascene holes 11, so that when two adjacent first yokes 1 are connected to one second yoke 5, the potting layer 2 of the end portions may be compressed to seal the first yokes 1 and the second yoke 5 at the end portions of the coils 3. To improve the thermal conductivity of the potting layer 2, the potting layer 2 is further embedded with thermally conductive insulating particles, wherein the thermally conductive insulating particle can use quartz sands, flint silica sands, PPS polyphenylene sulfide, PA46 nylon and the like. Embodiment 3 provides a method of manufacturing an electromagnetic coil assembly for a control rod driving mechanism, wherein the method is used to manufacture any one of the aforementioned electromagnetic coil assemblies for the control rod driving mechanism. The method comprises following steps carried out in sequence: S1. winding the coil winding around an inner frame 4 to obtain the one or more coils 3, and embedding the obtained one or more coils 3 in the one or more damascene holes 11 of the at least one first yokes 1; S2. integrally potting an assembly consisting of the one or more coils 3 and the at least one first yokes 1 to obtain a potting layer 2 in a space between a hole wall of each of the one or more damascene holes 11 and each of the one or more coils 3; and S3. completing an assembly of the at least one first yokes 1 and the at least one second yokes 5. Such technical solution is the processing method of any one of the electromagnetic coil assembly for the control rod driving mechanism provided by any one of the aforementioned embodiments. Preferably, a specific form of the assembly in S3 uses a bolted connection. A sealing ring 7 is disposed on a connecting surface of the at least one first yokes 1 and the at least one second yokes 5. The content described above is a further detailed description of the present invention with reference to the preferred embodiments, and the embodiments of the present invention should not be limited to the description. For the ordinary people skilled in the art, the other embodiments obtained without departing from the technical solutions of the present invention should be included in the scope of the present invention.
	description	The invention relates to a method for decontaminating a metal surface exposed to radioactive liquid or gas during operation of a nuclear facility, and in particular to a method for decontaminating a metal surface in the primary circuit of a nuclear reactor wherein the metal surface is covered with a radioactive metal oxide layer including chromium. The piping of a nuclear reactor is usually made of stainless steel or carbon steel. The steam generator tubes and main surfaces inside the primary circuit may include nickel alloys. When the nuclear reactor is operated, metal ions are released from these metal surfaces and transported into the coolant. Some of the metal ions are activated to form radioisotopes when passing the reactor core. A portion of the metal ions and radioisotopes is removed by the reactor water clean-up system (RWCU) during operation of the reactor. Another portion is deposited on the metal surfaces inside the reactor cooling system, and is later incorporated into metal oxide layers growing on the metal surfaces. Through the incorporation of radionuclides, these oxide layers become radioactive. The removal of the radioactive oxide layers is often necessary to reduce the level of personnel radiation exposure prior to carrying out inspection, maintenance, repair and dismantling procedures on the reactor cooling system. Depending on the type of alloy used for a component or system, the metal oxide layers contain mixed iron oxides with divalent and trivalent iron as well as other metal oxide species including chromium(III) and nickel(II) spinels. Especially the oxide deposits formed on the metal surfaces of the steam generator tubes may have a high Cr(III) or Ni(II) content which makes them very resistant and difficult to remove from the metal surfaces. Many procedures are described to remove metal oxide layers containing radioactive corrosion products from metal surfaces in the cooling system of a nuclear reactor. A commercially successful method is known as HP CORD UV and comprises the steps of treating the metal oxide layer with an aqueous solution of a permanganate oxidant in order to convert Cr(III) to Cr(VI), and subsequently dissolving the metal oxide layer under acidic conditions using an aqueous solution of an organic acid such as oxalic acid. The organic acid additionally serves to reduce a possible excess of the permanganate oxidant originating from the preceding oxidation step, and to reduce Cr(VI) dissolved in the oxidant solution to Cr(III). Additional or alternative reducing agents can be added for removing the permanganate oxidant and converting Cr(VI) to Cr(III). In a subsequent cleaning step, the decontamination solution containing the organic acid and corrosion products including metal ions and radioisotopes originating from the metal oxide layer, such as Fe(II), Fe(III), Ni(II), Co(II), Co(III) and Cr(III), is then passed through an ion exchange resin to remove the radioisotopes and some or all of the metal ions from the decontamination solution. The organic acid in the decontamination solution can be exposed to UV radiation and decomposed by photocatalytical oxidation to form carbon dioxide and water, thereby minimizing the amount of radioactive waste generated by the decontamination treatment. Ion exchange resin waste is commonly generated during the cleaning step as a result of removing the corrosion products from the decontamination solution. Depending on the corrosion products, cation and/or anion exchange resins are used to purify the decontamination solution. If chromium is present in the decontamination solution, the solution will initially contain anionic chromium complexes such as chromium oxalate Cr(III)(C2O4)33−. If the photocatalytical decomposition step is prolonged for sufficient time, the decontamination solution may also contain inorganic chromium compounds such as chromate salts Cr(VI)O42−. However, chromium oxalate is an extremely stable chelate complex, and it is often not possible to achieve a complete decomposition of the oxalate within the constraints of an industrial scale chemical decontamination application using this method alone. The anionic chromium complexes are picked up at the end of the cleaning step by the anion exchange resin as soon as the decontamination solution is depleted of free oxalic acid by photocatalytical oxidation, but before complete decomposition of the amount of oxalic acid bound in the chromium oxalate complex. Oxalic acid, or other organic acids and chelating agents employed in decontamination methods comparable to the one described above, may also be absorbed by the anion exchange resin, resulting in the presence of a substantial amount of the chelating agent in the final waste resin matrix. This may be undesirable in some jurisdictions for technical reasons or due to existing regulations. A further analysis of the published prior art reveals that processes for the removal of chromium in an inorganic non-chelated state during a chemical decontamination treatment of a nuclear facility have been proposed. Many of these processes employ ozone as an oxidizing agent for the oxidation of chromium in the oxide layers. For example, EP 1 054 413 B1 relates to a method of chemically decontaminating components of a radioactive material handling facility. Ozone gas having a high ozone concentration is generated by an electrolytic process. An ozone solution is prepared by injecting the ozone gas into an acidic solution of pH 6 or below. The ozone solution heated at a temperature in the range of 50° to 90° C. is supplied to a contaminated object to oxidize and dissolve a chromium oxide film by an oxidizing dissolving process. The ozone solution used in the oxidizing dissolving process is irradiated with ultraviolet rays to decompose ozone contained in the ozone solution, and the solution is passed through an ion-exchange resin to remove chromate ions contained in the ozone solution. Subsequently, an oxalic acid solution is supplied to the contaminated object to dissolve an iron oxide film by a reductive dissolving process. Oxalic acid remaining in the oxalic acid solution after the reductive dissolving process is decomposed by injecting ozone into the oxalic acid solution and irradiating the oxalic acid solution with ultraviolet rays, and ions contained in the oxalic acid solution are removed by an ion-exchange resin. EP 1 220 233 B1 directed to a chemical decontamination method for dissolving an oxide film adhered to a contaminated component. The method comprises the steps of preparing a decontamination solution in which ozone is dissolved and an oxidation additive agent is added, which suppresses corrosion of a metal base of the contaminated component, and applying the decontamination solution to the contaminated component, thereby to remove the oxide film by oxidation. The chromate ions formed in this step are captured on an anion exchange resin. However, the oxidation step is performed only after a reduction decontamination step using oxalic acid. EP 2 758 966 B1 relates to a method for decomposing an oxide layer containing chromium, iron, nickel, and radionuclides by means of an aqueous oxidative decontamination solution, which contains permanganic acid and a mineral acid, and which flows in a circuit, wherein the oxidative decontamination solution is set to a pH value 2.5. The decontamination solution is repeatedly passed through a cation exchange material for removing radioactive matter dissolved from the oxide layer, and is subsequently passed through an anion exchange resin to immobilize chromate ions formed during the oxidative decontamination step and regenerating the mineral acid. The method does not make use of any organic acid for dissolving metal oxide deposits other than hematite. U.S. Pat. No. 4,287,002 A relates to a method of decontaminating and removing corrosion products at least some of which are radioactive, from nuclear reactor surfaces exposed to coolant or moderator, said surfaces containing acid-insoluble metal oxides, including chromium oxide. The surfaces are decontaminated by treating the surface with ozone to oxidize acid-insoluble metal oxides to a more soluble state, removing oxidized solubilized metal oxides, and removing other surface oxides using low concentrations of decontaminating reagents. Chromic acid dissolved from the surfaces may be removed from the circulating water by contacting the solution with an anion exchange resin. EP 134 664 B1 is directed to a process for oxidizing chromium in deposits in the cooling system of a nuclear reactor using a solution of ozone, which consists of adding to the solution from 0.01 to 0.5% of a water-soluble cerium (IV) compound, from 0.1 to 0.5% of a water-soluble aromatic compound having at least one ketone group on an aromatic ring, or adding both. A process for decontaminating the cooling system of nuclear reactors comprises adding a decontamination composition to the coolant, circulating the coolant between the cooling system and a cation exchange resin, removing the decontamination composition by passing it through an anion exchange resin, adjusting the temperature to 40 to 100° C., adding the ozone oxidation composition, circulating the coolant through the cooling system, raising the temperature to at least 100° C., passing the coolant through an anion exchange resin or a mixed resin, adjusting the temperature to from 60 to 100° C. and repeating the addition of the decontamination composition and its removal. Due to the extremely limited half-life of ozone in water, ozone-based processes have proved as being ineffective for the decontamination of chromium rich oxide layers on a large scale, such for full system decontamination (FSD) of PWR (pressurized water reactor) type nuclear power plants. Processes trying to overcome this limitation of ozone through use of auxiliary substances, such as the use of cerium(IV) as a reaction intermediary, suffer from a greatly increased radioactive waste amount produced due to the auxiliary chemicals. These chemicals may also include nitrates or sulfates, which are either undesirable in the radioactive waste and/or raise compatibility concerns towards many of the materials present in primary circuit and auxiliary systems of the nuclear power plant. In addition, most of these processes involve a subsequent treatment with organic acids, in which chromium is present in a chelated state anyway. However, the main disadvantage of the ozone based processes is the use of ozone itself. Use of ozone in the oxidation step is costly and requires additional separate dosage stations and equipment since the ozone must be prepared on-site and cannot be stored in stock solutions at the nuclear facility. A further disadvantage of ozone is its nature as a toxic, even poisonous gas. Use of ozone in the closed containment of a nuclear power plant is therefore categorized as a safety risk and undesirable hazard. For this reason, liquid or non-gaseous alternatives are greatly preferred, which drastically reduce or completely eliminate the risk of gas poisoning for the involved personnel. The prior art processes using other oxidants present in the liquid phase are suitable to avoid the disadvantages of gaseous ozone. However, these processes are not optimized for waste reduction in a large scale application while making use of chromium removal in an inorganic, chelate-free stage. EP 2 923 360 B1 discloses a method for the chemical decontamination of a surface of a metal component having an oxide layer in the coolant system of a nuclear power plant. The method comprises at least one oxidation step in which the oxide layer is treated with an aqueous oxidizing solution, and a subsequent decontamination step, wherein the oxide layer is treated with an aqueous solution of an organic acid. The organic acid is capable of forming complexes with metal ions, especially nickel ions, in the form of a sparingly soluble precipitate. Prior to performing the decontamination step, metal ions such as Ni(II) are removed from the oxidizing solution using a cation exchange resin. While this process uses permanganate as an oxidizing agent, the removal of chromium in a chelate-free inorganic state is not taken into consideration. Rather, chromium released during the oxidation treatment is assimilated to the chromium released during the organic acid treatment which in all cases is present as a chelate complex. EP 090 512 A1 discloses a method of oxidizing chromium-containing corrosion products deposited on internal surfaces of a piping system through which an aqueous fluid is circulating. The method comprises the steps of adding to said circulating fluid a ferrate (VI) salt to form a dilute ferrate solution, while maintaining a pH of between 7 and 14, said ferrate reacting with chromium compounds contained in said corrosion products to form a chromate. The fluid is regenerated in situ by passing the fluid through an ion exchange resin to remove the products formed in the oxidation reaction and unreacted ferrate (VI). After the regeneration of the fluid, a CAN-DECON™ decontamination process may follow. According to this process, chromium is removed in an inorganic non-chelated state. However, the process generates much higher amounts of radioactive waste than a permanganate-based process, due to higher amount of oxidants employed and the auxiliary chemicals required to maintain the pH of the ferrate solution, while providing less satisfactory decontamination results than permanganate-based treatments and being more corrosive. The subsequent treatment of the surfaces with a CAN-DECON solution is proposed as an option, but is necessary to achieve an acceptable decontamination effect. Use of the CAN-DECON solution again results in the generation of chromium in a chelated state. The inventors therefore contemplate that the HP CORD UV process, or similar processes based on permanganate oxidation solutions, constitute the starting point and benchmark for the development of any improved processes for the decontamination of metal surfaces in a nuclear facility, such as the primary circuit of a nuclear reactor, wherein the metal surface is covered with a radioactive metal oxide layer including chromium. The examination of the prior art reveals that there is no single process for removal of chromium in an inorganic chelate-free state which is optimized for application at an FSD scale. In fact, the prior art processes are either more corrosive, or riskier through the use of poisonous gas, or produce more waste. None of these processes would be more effective and quicker for a chemical decontamination application than the known permanganate-based HP CORD UV process, and none would be able to guarantee the efficient removal of chromium in a chelate-free state. It is therefore an object of the present invention to provide a cost effective decontamination method for a nuclear facility and its components suitable for applications up to a Full System Decontamination scale which allows for savings of radioactive waste and also savings of time for the decontamination treatment cycles. As a further object, the invention aims at providing a decontamination method which generates chelate free ion exchange material waste after chemical decontamination of the primary cooling system or its components of a nuclear power plant. These objects are solved by a decontamination method according to claim 1. Advantageous and expedient embodiments of the invention are indicated in the dependent claims which can be combined with each other independently. In one aspect, the invention provides a method for decontaminating a metal surface exposed to radioactive liquid or gas during operation of a nuclear facility, wherein the metal surface is covered with a metal oxide layer including chromium and radioactive matter, the method comprising: a) an oxidation step wherein the metal oxide layer is contacted with an aqueous oxidation solution for converting chromium into a Cr(VI) compound and dissolving the Cr(VI) compound in the oxidation solution, wherein the aqueous oxidation solution comprises a permanganate oxidant but no additional mineral acid; b) a first cleaning step wherein the aqueous oxidation solution containing the Cr(VI) compound is passed directly over an anion exchange material and the Cr(VI) compound is immobilized on the anion exchange material; c) a decontamination step following the first cleaning step wherein the metal oxide layer subjected to the oxidation step is contacted with an aqueous solution of an organic acid for dissolving the metal oxide layer, thereby forming a decontamination solution containing the organic acid, metal ions and radioactive matter, and wherein the decontamination solution is passed over a cation exchange material for immobilizing the metal ions and radioactive matter; d) a second cleaning step wherein the organic acid contained in the decontamination solution is decomposed; and e) optionally repeating steps a) to d). The present invention provides a safe and reliable chemical decontamination process that can be applied at an industrial scale up to full system decontamination (FSD), including the simultaneous treatment of a complete primary coolant circuit including auxiliary systems of a nuclear power plant, and that guarantees the absence of chelating agents originating from the decontamination chemicals in the resulting radioactive waste as well as the absence of corrosive mineral acids. The amount of radioactive waste generated as a result of the decontamination treatment is furthermore kept as low as possible to reduce the high disposal costs involved. The inventors contemplate that one of the key factors for solving the above problem consists in achieving a chemical state for chromium, in which it can be completely removed from the process solution in an inorganic non-chelated state. One viable option is the removal of chromium as a Cr(VI) compound such as chromate. The decontamination method of the present invention avoids the presence of a substantial amount of organic anionic chromium complexes in the second cleaning step which would have to be picked up by an anion exchange material, and which would then create additional resin waste due to the presence of a chelating agent such as an organic acid. Since chromium is removed already at the end of or during the oxidation step, only a residual amount of chromium complexes such as chromium oxalate is present in the second cleaning step at the end of the decontamination treatment cycle. This residual amount of organic anionic chromium complexes can be decomposed in a considerably shorter time using a suitable technology such as the described photocatalytical decomposition, or preferably can be transferred to the next treatment cycle starting with the oxidation step, wherein the chelating agent is completely decomposed very effectively and in a very short time by the permanganate oxidant. During the same process any chelated Cr(III) is converted to a Cr(VI) compound, that can then be removed from solution in a chelate-free state in the following cleaning step. A combination of both techniques can also be employed, wherein the amount of organic acid is first reduced using a technology such as photocatalytical decomposition, and the residual amount of chromium complexes is subsequently decomposed by adding the oxidation solution comprising a permanganate oxidant. This combination results in a faster treatment than the photocatalytical decomposition technology alone, and produces less additional waste as if the chromium chelate complexes were decomposed by adding only the permanganate-based oxidation solution. Accordingly, the potential presence of organic anionic chromium complexes at the end of the decontamination treatment cycle has only minimal impact on the waste produced during the treatment cycle. The method according to the invention guarantees that no organic acids or chelating agents are present in the spent ion exchange material waste. According to a preferred embodiment, this can be ensured either through sluicing of the ion exchange material immediately after its use and before any injection of chelating substances has taken place, or through other methods or technical restrictions such as appropriate valve positioning, so that the ion exchange material is not exposed to an organic acid solution at a later stage of the decontamination process. Moreover, when chromium is removed from the process in the form of a Cr(VI) compound such as a chromate or dichromate at the end of or during the oxidation step, the consumption of anion exchange material is considerably lower as compared to removal of chromium complexes. The Cr(VI) compound is picked up by the anion exchange material as only one equivalent per chromium atom, instead of three equivalents in the case of e.g. chromium(III) trioxalate. In a practical example, this means a consumption of only 100 L of anion exchange material as compared to up to 300 L according to the prior art decontamination process, for the same amount of chromium removed. Further, since regulatory provisions in some countries limit the total amount of chelating agents in the radioactive waste, the commercial prior art process may require a decomposition of the chromium complexes during the final cleaning step, for example by photocatalytical oxidation. This additional decomposition step requires a considerable amount of time which may range from hours to days per treatment cycle. In contrast thereto, the inventive process allows for considerable time savings because the amount of chromium complexes at this stage is lower due to the removal of a large fraction of the chromium as a Cr(VI) compound before the injection of the chelating acids, and because any residual amount of chromium present in the decontamination solution after the cleaning step can be transferred to the next decontamination cycle. At the start of the next cycle, chromium is again oxidized within minutes to form a Cr(VI) compound, which is then captured in an inorganic chelate-free state. In a second aspect, the invention provides a method of reducing an amount of spent ion exchange material waste from decontamination of a metal surface exposed to radioactive liquid or gas, wherein the metal surface is covered with a metal oxide layer including chromium and radioactive matter, and wherein the decontamination comprises a plurality of treatment cycles, each treatment cycle comprising: an oxidation step to convert chromium in the metal oxide layer to a Cr(IV) compound; a first cleaning step wherein a substantial amount of the Cr(VI) compound is immobilized on an anion exchange material without contacting the anion exchange material with a chelating organic acid; and a decontamination step following the first cleaning step wherein a decontamination solution comprising an organic acid and metal ions dissolved from the metal oxide layer is passed over a cation exchange resin for immobilizing the metal ions; wherein any chelated chromium contained in the decontamination solution is carried to the oxidation step of a following treatment cycle. The inventors contemplate that the inventive decontamination method results in a reduction of spent ion exchange material waste of greater than 20 percent by volume as compared to a decontamination method including a step of contacting the Cr(VI) compound with a chelating organic acid, preferably greater than 30 percent, and more preferably 30 to 40 percent. Preferably, in both aspects of the invention, the anion exchange material is an inorganic anion exchange material. Use of an inorganic anion exchange material is made possible through the removal of chromium in an inorganic non-chelated state and through the absence of an organic acid. Use of an inorganic anion exchange material has not been reported so far for any large scale chemical decontamination applications due to its incompatibility with organic acids. Furthermore, use of an inorganic anion exchange material also enables the removal of permanganate from the process solution through ion exchange mechanisms, instead of requiring reduction of permanganate to manganese in a lower oxidation state which is then either filtered out or removed from the process solution via cation exchange. The removal of permanganate prior to the decontamination step using an inorganic anion exchange material results in additional waste savings of greater than 30 percent as compared to the removal of manganese via cation exchange. Additional waste savings of preferably greater than 40 percent, and more preferably greater than 60 percent and more can be achieved in this way. Moreover, removal of residual permanganate on an ion exchange material instead of reducing it through the addition of an organic acid may also avoid emissions of gaseous carbon dioxide from decomposition of the organic acid. Preferably, the Cr(VI) compound has a greater affinity towards said anion exchange material than permanganate. More preferably, the affinity of the Cr(VI) compound to the, preferably inorganic, anion exchange material is between five to ten times higher than the affinity of permanganate. The higher affinity of the Cr(VI) compound allows for separating Cr(VI) from permanganate during the course of the decontamination process, by limiting the amount of anion exchange material available for bonding chromium. This feature of the decontamination method may be of particular interest for nuclear power plants in operation, or close to the operational period, due to the higher contents of radioactive chromium-51 present then. The possibility of separating radioactive chromium compounds in a waste fraction having a high activity content from a permanganate waste fraction having a much lower activity content can have considerable advantages with respect to waste disposal, depending on applicable regulations on the site. The amount of anion exchange material required for bonding the total amount of chromium present in the oxidation step can be determined based on the amount of chromium analyzed in the oxidation solution. Permanganate initially fixed on the anion exchange material is displaced by the Cr(VI) compound when it arrives at the ion exchanger. The selectivity of the anion exchange material towards the chromium compound makes it possible to remove the Cr(VI) compound from the oxidation solution while maintaining a permanganate concentration at a sufficiently high level to enable the oxidation process to continue. Thus, in a preferred embodiment, the first cleaning step can be started already during the oxidation step so that the oxidation step and the first cleaning step are at least partially conducted simultaneously. This can be used to achieve additional time savings. Preferably, the oxidation solution containing the Cr(VI) compound and the permanganate oxidant is passed over the anion exchange material, preferably an inorganic anion exchange material, and at least the Cr(VI) compound is immobilized on the anion exchange material. More preferably, the oxidation solution is passed over the anion exchange material before a concentration of the Cr(IV) in the oxidation solution has stabilized at an essentially constant level. The construction and method of operation of the invention together with additional objects and advantages thereof will be best understood from the following description of specific embodiments which are given for illustrative purposes only and which are not intended to limit the scope of the present invention. According to the method of the present invention, a metal oxide layer containing radioisotopes is effectively removed from metal surfaces of a nuclear facility, and in particular from metal surfaces located in the primary cooling system of a nuclear reactor. The primary cooling system is understood as comprising all systems and components which are in contact with the primary coolant during reactor operation, including but not limited to the reactor vessel, reactor coolant pumps, pipework and steam generators, as well as auxiliary systems such as the volume control system, pressure reducing station and reactor water clean-up system. The decontamination method of the present invention is particularly useful for decontamination of the primary cooling system or components thereof in a boiling water reactor or a pressurized water reactor, and preferably a nuclear reactor comprising steam generator piping having metal surfaces of nickel alloys such as Inconel™ 600, Inconel™ 690 or Incoloy™ 800, and/or materials with a high chromium content, or large surfaces of chromium containing materials. The inventors contemplate that the method of the present invention can also be used for decontamination of the coolant and/or moderator circuit of a heavy water reactor such as a CANDU™ nuclear reactor or any other heavy water reactors, but is not limited to these reactor types. The decontamination treatment can be carried out on reactor subsystems and components. Preferably, the decontamination method of the present invention is carried out as full system decontamination. During full system decontamination the contaminated metal oxide layer is removed from all metal surfaces in the reactor cooling system that are in contact with the primary coolant during reactor operation. Typically, full system decontamination involves all parts of the primary coolant circuit and the steam generator as well as the volume control system, the pressure reducing station and possibly other systems which are contaminated to a certain extent. According to a preferred embodiment, the decontamination method can be applied using an external decontamination equipment for injection of decontamination chemicals, for monitoring the decontamination treatment, for increasing the available ion exchange rate, and for achieving the decontamination targets in a faster, more economical and safer way. The process temperatures are preferably kept below the boiling point of water at atmospheric pressure in order to eliminate the need of using complex and expensive pressure-proof components for the external decontamination equipment. The chemicals used for the decontamination treatment can be injected into the primary coolant circuit of the nuclear reactor at a dosing station located in the low-pressure part of the coolant circuit. Preferably, the external decontamination equipment is used for dosing the decontamination chemicals. Ion exchange materials and chemicals used in the decontamination method of the present invention are commercially available and can be held in stock at the nuclear power plant facilities. In general, one or more decontamination treatment cycles are carried out in order to achieve a satisfactory reduction of activity on the metal surfaces. The reduction of surface activity and/or the dose reduction correlating to surface activity reduction is referred to as “decontamination factor”. The decontamination factor is calculated either by the specific surface activity before decontamination treatment divided by the specific surface activity after the decontamination treatment, or by the dose rate before decontamination treatment divided by the dose rate after decontamination treatment. Preferably, the decontamination factor of a technically satisfying decontamination treatment is greater than 10. The various steps of the decontamination method of the present invention are now described in greater detail below. Oxidation Step For carrying out the oxidation step, an aqueous solution of the permanganate oxidant is injected into the primary coolant within the primary coolant circuit or the subsystem which is to be decontaminated, and the aqueous oxidation solution comprising the permanganate oxidant is circulated through the system. Preferably, the permanganate oxidant is injected into a low-pressure section of the cooling and/or moderator system. Examples for suitable injection positions are the volume control system, the reactor water cleanup system and/or a residual heat removal system. More preferably, the solution of the permanganate oxidant can be introduced into the primary cooling system or moderator system by means of an external decontamination device. The oxidation step is carried out as a mere pre-oxidation step. Thus, during the oxidation step, the metal oxide layer substantially remains on the metal surface to be decontaminated, and no activity is removed from the system to be decontaminated. Rather, the permanganate oxidant acid reacts with spinel-type metal oxides in the metal oxide layer which are almost inert to organic acids to break up the oxide structure and convert the spinel-type metal oxides into more soluble oxides. Cr(III) in the metal oxide layer is oxidized to form soluble Cr(VI) compounds, and the Cr(VI) compounds are dissolved in the permanganate-based oxidation solution. Depending on the pH value of the oxidation solution, the Cr(VI) compound may comprise chromic acid, dichromic acid and/or salts thereof. Preferably, the permanganate oxidant is selected from permanganic acid, HMnO4, and alkali metal permanganate, optionally in combination with an alkali metal hydroxide. Permanganic acid is preferred over alkali metal permanganate salts because less waste is produced. Depending on the nature of the metal oxide layer, however, an alkaline oxidation solution may also be used for oxidizing the metal oxide layer. The alkaline oxidation solution may include an alkali metal permanganate salt such as sodium or potassium permanganate, as well as an alkali metal hydroxide. It may also be useful to switch between acidic oxidation conditions and alkaline oxidation conditions in the oxidation steps of subsequent decontamination treatment cycles. Still more preferably, the permanganic acid is prepared on demand by ion exchange reaction between an alkali metal permanganate salt and a cation exchange resin. The permanganic acid can be prepared on site, or can be provided as an aqueous stock solution having a concentration of from 1 to 45 g/L, preferably a concentration of from 30 to 40 g/L. According to the invention, no additional mineral acid such as sulfuric acid, nitric acid, hydrochloric acid or phosphoric acid is added to the oxidation solution. Preferably, the pH of the oxidation solution is kept at or above 2.5 which can be achieved using permanganic acid as the sole oxidant. Carrying out of the oxidation step at a pH>2.5 can avoid substantial corrosion of the metal surface to be decontaminated. In addition, the absence of an additional mineral acid in the oxidation solution avoids too high dissolution rates of the metal oxide layer which could be detrimental in FSD operations. Preferably, the oxidation step is carried out at a temperature of between about 20 to 120° C., more preferably at a temperature of from 80 to 95° C. The oxidation step is faster at higher temperatures. Accordingly, higher oxidation temperatures are preferred. Moreover, the boiling point of an aqueous solution of permanganic acid under atmospheric pressure is higher than 95° C., which makes it easier to circulate the oxidation solution through the cooling system using the pumps of the external decontamination device. However, it is also possible to carry out the oxidation step at temperatures of up to 120° C. at a higher than atmospheric pressure, with or without the use of an external decontamination device. Preferably, the concentration of the permanganate oxidant in the oxidation solution within the primary cooling system is controlled to be in the range of from 10 to 800 mg/kg during the oxidation step, and preferably to range from 50 to 200 mg/kg. If the concentration of the permanganate oxidant in the oxidation solution is lower than 10 mg/kg, the reaction rate of the oxidation may be too low and several additional injections may be required. If the concentration of the permanganate oxidant in the oxidant solution exceeds 800 mg/kg, a large excess of the oxidant may be present at the end of the oxidation step which can generate an unnecessary amount of waste. Preferably, the amount of the permanganate oxidant is controlled to be as low as possible at the end of the oxidation step because removal of excess permanganate oxidant will increase the amount of secondary waste. Preferably, the progress of the oxidation step is monitored by controlling the amount of the permanganate oxidant remaining in the oxidation solution, and by monitoring the concentration of Cr(VI) dissolved in the permanganate-based oxidation solution. As long as the oxidation reaction continues and the oxidation of the metal oxide layer is incomplete, the permanganate oxidant continues to be consumed and, in most cases, the concentration of Cr(VI) compounds increases. The residence time of the oxidation solution in the cooling system during the oxidation step may comprise a plurality of hours, preferably 30 hours or more in large and complex applications such as full system decontaminations. It is desired that the oxidation of the metal oxide layer is substantially complete so that as much as possible of the metal oxide layer thickness is reacted during the oxidation step. Preferably, the oxidation step is terminated when no further increase of the Cr(VI) concentration in the oxidation solution can be determined, more preferably when the permanganate concentration in the oxidation solution has stabilized additionally at an essentially constant concentration level and permanganate oxidant is no longer being consumed, and most preferably when the permanganate oxidant has been completely consumed. Instead of, or in addition to, monitoring the concentration of Cr(VI) and/or permanganate, it is also possible to monitor the presence of the radioisotope Cr-51 in the oxidation solution by means of gamma spectroscopy. First Cleaning Step In the first cleaning step, the aqueous oxidation solution containing the Cr(VI) compound is passed directly over an anion exchange material, before or after removal of the permanganate oxidant, in order to capture at least the chromate or dichromate ions present in the oxidation solution and optionally any excess of permanganate ions still contained in the oxidation solution. Passing the oxidation solution directly over an anion exchange material means that no cation exchange is performed during the first cleaning step or the oxidation step. Treating the oxidation solution by passing over a cation exchange material is not necessary in this stage of the decontamination process, since the amount of divalent metal ions or activity dissolved from the metal oxide layer in the oxidation solution is rather low. Suitable anion exchange materials for use in the decontamination method of the present invention are resistant to the harsh oxidizing and optionally acidic conditions present in the oxidation solution. It is also possible to use different anion exchange materials or combinations of anion exchange materials each being optimized for the specific conditions in the different process steps. Anion exchange materials suitable for use in the decontamination method of the present invention are commercially available, such as Levatite™ M800 from Lanxess, Diaion SA 10AOH from Mitsubishi Chemicals or NRW 8000 from Purolite. The anion exchange materials can be included in the external decontamination device, and may be configured as membranes or ion exchange columns filled with the anion exchange material. Alternatively or additionally, the inventors contemplate use of the anion exchange materials which are present in the reactor water clean-up system or any other suitable internal system of the nuclear facility. In a preferred embodiment of this invention, the anion exchange material is contained within an external module which is preferably configured for a prompt charge and discharge of different amounts of said material. More preferably, the external module is an integral part of the external decontamination equipment. The first cleaning step is controlled by monitoring the removal of the Cr(VI) compound and/or the permanganate oxidant from the oxidation solution, preferably by photometric measurements, by determining the oxidation potential of the oxidation solution relative to a reference electrode, and/or by determining a concentration of chromium and manganese through an instrumental analysis technique such as atomic absorption spectrometry (AAS) or inductively coupled plasma (ICP) mass spectrometry. The anion exchange material may be an anion exchange resin. Preferably, the anion exchange material is an anion exchange resin which is employed during power generating operation of the nuclear facility. In a preferred embodiment, the anion exchange material is an inorganic anion exchange material. Use of an inorganic anion exchange material is advantageous in that it is resistant to harsh oxidizing conditions and chemically stable over long disposal times. More preferably, the anion exchange material has an affinity to the Cr(VI) compound which is higher than an affinity to permanganate. More preferably, the affinity of the anion exchange material to the Cr(VI) compound is at least between five to ten times higher than the affinity to permanganate. The higher affinity towards the Cr(VI) compound makes possible to separate the Cr(VI) compound from the permanganate oxidant during the course of the first cleaning step. The first cleaning step can be started when the oxidation step is terminated, that is when no further increase of the Cr(VI) concentration in the oxidation solution can be determined. According to a preferred embodiment, however, the first cleaning step is started already during the oxidation step. Preferably, the aqueous oxidation solution containing the Cr(VI) compound and the permanganate oxidant is passed over the anion exchange material preferably before the chromium concentration has stabilized in the oxidation solution, i.e. while the chromium concentration is still increasing. Accordingly, the oxidation step and the first cleaning step are at least partially conducted simultaneously. This can be used to achieve additional time savings. More preferably, an amount of the Cr(VI) compound in the oxidation solution is determined, and the amount of anion exchange material used in the first cleaning step is controlled on the basis of the amount of the Cr(VI) compound determined in the oxidation solution. Still more preferably, the amount of the anion exchange material is controlled so as to substantially immobilize the Cr(VI) compound only, and retain at least part of, or substantially all of, the permanganate oxidant in the oxidation solution. Using slightly less of the anion exchange material than required for bonding all of the Cr(VI) compound contained in the oxidation solution guarantees that the permanganate oxidant is not removed from the oxidation solution. Due to the higher affinity of the anion exchange material to the Cr(VI) compound, any permanganate initially captured on the anion exchange material is displaced by the Cr(VI) compound when it passes the anion exchange material. Therefore, the Cr(VI) compound is selectively removed from the oxidation solution while a concentration of the permanganate oxidant in the oxidation solution remains sufficiently high to further oxidize the metal oxide layer. Most preferably, the amount of the anion exchange material is controlled so as to immobilize about 80-95 weight percent, preferably 85 to 100 weight percent, of the Cr(VI) compound contained in the oxidation solution. The permanganate oxidant is preferably removed from the oxidation solution after immobilizing of the Cr(VI) compound by immobilizing on an anion exchange material, before commencing the decomposition step. According to a further preferred embodiment, the first cleaning step is started when the oxidation step is terminated and the permanganate oxidant is removed substantially completely from the oxidation solution. In this embodiment, the oxidation solution containing the Cr(VI) compound is passed over the anion exchange material after complete removal of the permanganate oxidant. Preferably, complete removal of the permanganate oxidant is effected by reacting permanganate with a stoichiometric or under-stoichiometric amount of a reducing agent without changing the oxidation state of the Cr(VI) compound. The reducing agent can be an inorganic or an organic reducing agent. More preferably, the reducing agent is a compound that does not release any metal cations when being reacted with the permanganate oxidant. Still more preferably, the reducing agent is selected from the group consisting of hydrogen, hydrogen peroxide, hydrazine, non-chelating monocarboxylic acids, non-chelating dicarboxylic acids, and derivatives thereof. According to an alternative embodiment, the reducing agent comprises a metal cation which changes its oxidation state when reacted with permanganate, and more preferably a cation selected from the group consisting of iron(II) and chromium(III). This embodiment is less preferred because additional waste is generated. Complete removal of the permanganate oxidant can also be effected by means of electrochemical reduction using an electrode or other electrochemical means. The above described reducing agents and/or electrochemical reduction can also be used for removing the permanganate oxidant from the oxidant solution after immobilizing of the Cr(VI) compound on an anion exchange material, before commencing the decomposition step. Preferably, however, the oxidation solution containing the Cr(VI) compound is not contacted with any organic acid prior to the subsequent decontamination step. In a further preferred embodiment, the anion exchange material used to remove the Cr(VI) compound and/or the permanganate oxidant is never exposed to an organic acid solution, neither in the first cleaning step, nor in a subsequent step of the decontamination treatment. An exposure of the anion exchange material to an organic acid would wash down manganese from the material. In addition, chromium would be washed down from the anion exchange material and additionally form a chromium chelate complex, which is to be avoided. Therefore, the anion exchange material used in the first cleaning step preferably is either discarded directly after its use, or any process solution containing organic acid is prevented from flowing through the anion exchange material used in the first cleaning step by appropriate valve positioning in the decontamination circuit, so that use of the anion exchange material can be resumed in a posterior treatment cycle if its capacity has not yet been exhausted. Further, preventing the anion exchange material from being exposed to the organic acid facilitates and/or enables the use of inorganic anion exchange materials. These materials are suitable for the first cleaning step of the present invention, but have not been employed for any reported chemical decontamination application due to their general incompatibility with organic acids used in the decontamination step. As soon as the removal of the Cr(VI) compound is completed or the concentration of the Cr(VI) compound is below a predetermined target value, the decontamination step is started. Decontamination Step In the decontamination step, the metal oxide layer subjected to the oxidation step is contacted with an aqueous solution of an organic acid. The organic acid serves as a decontamination reagent and reacts with the metal oxides and radioactive matter incorporated in the metal oxide layer, thereby forming a decontamination solution containing the decontamination reagent, one or more metal ions dissolved from the metal oxide layer, and the radioactive matter. Preferably, the organic acid is an organic acid that can be treated in situ at a later stage to minimize or completely eliminate the waste volume associated to it. According to a further preferred embodiment, the organic acid used in the decontamination step is selected from the group consisting of monocarboxylic acids such as formic acid and glyoxylic acid, aliphatic dicarboxylic acids such as oxalic acid, alkali metal salts of monocarboxylic acids and aliphatic dicarboxylic acids, and mixtures thereof. More preferably, the organic acid is an aliphatic dicarboxylic acid selected from linear aliphatic dicarboxylic acids having 2 to 6 carbon atoms. Most preferably, the organic acid is oxalic acid. The decontamination step further comprises passing the decontamination solution over a cation exchange material to immobilize the metal ions and the radioisotopes dissolved therein. During this step, all cations dissolved in the decontamination solution, including Mn(II) generated from the decomposition products of the permanganate oxidant consumed during the oxidation step as well as the radioisotopes dissolved in the decontamination solution, are removed from the decontamination solution and are permanently captured on the cation exchange material. The cation exchange material may be a cation exchange resin of the type employed in the nuclear power plant during power generating operation, or any other suitable cation exchange material. Preferably, the cation exchange material used in the decontamination step is a cation exchange resin which is present in the water clean-up system of the nuclear reactor. The organic acid dissolved in the decontamination solution is regenerated by release of hydrogen ions during the cation exchange reaction. Therefore, the organic acid is not depleted in the decontamination step, and can be used continuously for dissolution of the metal oxide layer. Accordingly, it is possible to employing sub-stoichiometric amounts of the organic acid. The decontamination of the metal surface covered with the metal oxide layer is only limited by a decrease of the solubility of the metal oxide layer which is due to the fact that the metal oxide layer reacted in the oxidation step is completely removed at the end of the decontamination step. Therefore, a further oxidation of the remaining metal oxide layer is often required to dissolve additional metal ions from the metal oxide layer into the decontamination solution. The progress of the decontamination step and the cation exchange reaction can be monitored by measuring the concentration of selected radioisotopes and metal ions. Samples can be taken from the decontamination solution and analyzed by spectroscopic methods such as atomic absorption spectroscopy (AAS) and inductively coupled plasma (ICP) mass spectrometry. The amount of radioisotopes dissolved in the decontamination solution can be determined by different methods of gamma spectroscopy, such as by means of high purity germanium detectors, sodium iodide detectors, or by other suitable methods depending on the nature of the radioisotopes present. The decontamination step is terminated as soon as no substantial increase of the amount of metal ions removed from the decontamination solution and immobilized on the cation exchange material is determined, and/or no further increase of the activity of the radioisotopes immobilized at the ion exchange materials can be measured. Second (Intermediate or Final) Cleaning Step Before starting a further oxidation step to solubilize the metal oxide layer now exposed by the decontamination solution, the organic acid must be removed from the decontamination solution. For example, the system can be drained and rinsed with additional water until the organic acid is completely removed. However, this is the least favored option, because it would generate a large amount of radioactive liquid waste. The water would have to be treated at a later stage in such a way that no chelates are generated. The organic acid can also be removed by ion exchange mechanisms, but this would generate undesired chelate-containing waste. According to another option, the organic acid can be removed from the decontamination solution by reacting the organic acid with permanganic acid or another permanganate or oxidizing compound. The process of decomposing the organic acid by reacting with permanganate can preferably be used for decontamination systems having small volumes, e.g. during the decontamination of isolated heat exchangers and the like. However, this reaction requires a substantial amount of permanganic acid or other permanganate compound and also generates additional secondary waste in the form of e.g. manganese ions that have to be removed from the solution via ion exchange, in a way comparable to the other metal cations generated from the metal oxide layer. Therefore, the preferred embodiment of the decontamination method comprises a decomposition step using another method for the reduction of the organic acid present in the decontamination solution, such as photocatalytical oxidation of the organic acid. An oxidation of the organic acid itself, photocatalytically or otherwise, does not necessarily generate additional radioactive waste since the decomposition of the organic matter results in the formation of water and carbon dioxide. Therefore, selecting an appropriate decomposition method makes it possible to avoid the formation of any unnecessary secondary radioactive waste in this stage. According to a preferred embodiment, the organic acid is reacted with an oxidant that does not contribute to the amount of radioactive waste generated during the decontamination process. Preferably, the organic acid is decomposed to form carbon dioxide and water. More preferably, the organic acid is decomposed by reacting the organic acid with an oxidant such as hydrogen peroxide, most preferably while simultaneously exposing the decontamination solution to UV radiation. Use of hydrogen peroxide is advantageous because it is an industrial chemical which is commercially available and can be stored in stock solutions at the nuclear plant facilities. Oxygen or ozone could also be used for decomposing the organic acid, but are less preferred because these oxidants require additional equipment and are associated with other risks, especially in the case of ozone. Preferably, a photocatalytical oxidation is employed to increase the reaction speed. Preferably, the temperature of the decontamination solution during decomposition of the organic acid is kept between 20 and 95° C. A UV reactor is preferably immersed into the decontamination solution to maximize the area of exposure to UV light, and hydrogen peroxide is injected into the decontamination solution upstream of the UV reactor such that the hydrogen peroxide is thoroughly mixed with the decontamination solution prior to reaching the UV reactor. The injection of hydrogen peroxide into the decontamination solution is preferably controlled so that no hydrogen peroxide is determined downstream of the UV reactor. Preferably, hydrogen peroxide downstream of the UV reactor is monitored continuously, and the rate of the hydrogen peroxide injection is adjusted accordingly. Through the application of the invention, the duration of the decomposition step can be reduced when compared with the prior art. This is a consequence of the significantly lower amount of chromium complexes present in solution. In the prior art decontamination process both, chromium released during the oxidation phase and chromium released during dissolution of the metal oxide layer in the decontamination step, are commonly present at this stage. According to the invention, only chromium compounds released from the metal oxide layer during the decontamination step are present in the process solution at the time of the decomposition of the organic acid. The decomposition of the organic acid is preferably terminated if the decontamination solution is completely depleted of the organic acid, including the organic acid bound in chelate complexes. While less preferred, but also possible depending on project objectives and local specific considerations, the decontamination solution can be depleted to a concentration of the free organic acid in the solution of up to 50 mg/kg or less. Higher concentrations of free organic acid are also possible but even less preferable, due to an increase of permanganate consumption in a subsequent treatment cycle. Chromate resulting from the decomposition of the organic acid and any chromium complexes still present in the decontamination solution after the decomposition of the organic acid, such as Cr(III) oxalate complexes, are preferably carried to the next oxidation step. In the oxidation step, any remaining amount of the chelating organic acid, if present, is decomposed by the action of the permanganate oxidant, and eventually remaining Cr(III) compounds are oxidized to form Cr(VI) compounds. Thus, no organic acid or other chelating agent is transferred to the ion exchange material waste as a result of the second cleaning step. In a final cleaning step, when the metal oxide layer is completely removed from the metal surface and/or the desired decontamination factor is achieved, the conductivity of the primary coolant may be controlled to be 10 μS/cm at 25° C. or lower, although final water quality criteria can vary from facility to facility. Preferably, the final cleaning step is conducted at a temperature of 70° C. or less, more preferably 60° C. or less. The second cleaning step may already be started during the decontamination step. The decontamination solution is then passed over a cation exchange resin while the organic acid is simultaneously decomposed, for instance by photocatalytic oxidation. The removal of the metal ions and radioisotopes in the second cleaning step and/or the decontamination step may take place in a bypass conduit in the low-pressure part of the reactor, most preferably using cation exchange columns present in the water cleaning system of the nuclear reactor. It is also possible to operate external ion exchange modules, alone or in parallel to the ion exchange columns of the reactor water cleaning system. The oxidation step, the first cleaning step, the decontamination step and the second cleaning step will form a decontamination treatment cycle. These steps may optionally be repeated so that the decontamination method may comprise two or more decontamination treatment cycles, preferably two to five treatment cycles. It has been found that a satisfactory decontamination factor can be achieved with this number of treatment cycles in full system decontamination and/or decontamination of subsystems or components of a pressurized water reactor. However, the number of decontamination treatment cycles is not limited to the numbers given above, but may also depend on the reactor design, the level of radioactive contamination and the decontamination objectives. The decontamination method of the present invention is preferably applied to the decontamination of the primary coolant circuit of a nuclear reactor. The primary coolant circuit is provided for cooling of the reactor core including the fuel bundles and for transferring the hot coolant to the steam generator where energy is transferred from the primary coolant to a secondary cooling circuit passing through the steam generator. Calculations have been performed on a full system contamination of a primary coolant system having a system volume of 360 m3, and using oxalic acid as the organic acid and conventional anion exchange resin such as the one used during operation of a nuclear power plants as the sole anion exchange material during 5 decontamination treatment cycles. The calculations show that the resin consumption for capturing chromium and additional manganese spent for oxidizing residual chromium oxalate in the oxidation step, according to the present invention, will result in a consumption of about 1560 liters of conventional anion exchange resin and about 1400 liters of conventional cation exchange resin, adding to a total of about 2960 liters of spent conventional waste resin. A decontamination treatment of the same system using a commercial prior art process would result in a consumption of about 4460 liters of conventional anion exchange resin for capturing chromium oxalate at the cleaning step. Thus, the waste savings amount to a total of about 1500 liters of conventional ion exchange resins to be used in the decontamination process corresponding to waste savings of about 34 percent in volume. In addition, since the spent anion exchange resin is free of chelating agents, disposal of the resin is significantly simplified, or even made only possible through the application of this method. Although the invention is illustrated and described herein as embodied in a method for surface decontamination, it is not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the scope of the appended claims.
	claims	1. A modular gaseous electrolysis apparatus, wherein an electrolyte is gas, the gaseous electrolysis apparatus comprising:an actively-cooled header module with at least one electrical connector or coupling, andwherein said actively-cooled header module is configured to be one of:opened/removed, andclosed/installed;a heat exchanger module configured to:remove heat from a surface of a reaction chamber module; andfacilitate and control a thermal gradient through a removable cathode module, and a wall of the reaction chamber module, wherein said heat exchanger module is separate from the removable cathode module, and wherein the reaction chamber module is configured to receive the gas electrolyte;wherein said heat exchanger module is capable of assembly, and disassembly, andwherein the gaseous electrolysis apparatus comprises at least one or more of:wherein the gaseous electrolysis apparatus is configured to include hermetic seals to maintain integrity in an elevated pressure and temperature environment;a flash boiler is configured to provide a mist of water or other coolant to the outer surface of the reaction chamber module; ora plurality of spray nozzles is configured to cool at least one portion of the reaction chamber module and to facilitate and control thermal diffusion through the removable cathode module; andat least one or more of:at least one steam pressure port; orat least one thruster port configured to provide pressure output;a gas handling system, configured to provide the gas electrolyte, said gas handling system mechanically coupled to the reaction chamber module and separate from the reaction chamber module,wherein said gas handling system comprises at least one or more of:a measurement container configured to temporarily store reactant gas and periodically permit extraction of the reactant gas from said measurement container; ora subsystem comprising an acoustic sensor or other type of electronic interface, configured to facilitate estimation of quantities of reactant gas;an electronic control circuit module electrically coupled or connected to said actively-cooled header module and said gas handling system and configured to electronically control said gas handling system;a modular, removable anode module comprising:an electric heater disposed within the modular, removable anode module;wherein said actively-cooled header module, said heat exchanger module, the removable cathode module, said modular, removable anode module, said electric heater, said gas handling system, and said electronic control circuit module are removably and mechanically coupled to the reaction chamber module; andwherein said gas handling system comprises:a gas manifold module that controls gas flow external to the reaction chamber module,wherein said gas manifold module comprises:a reaction gas product collection manifold module, mechanically coupled to said reaction chamber module, and coupled to a reaction gas product collector and further comprising:at least one mechanical valve;at least one electronically controlled valve;at least one separator valve with an electronic interface for matter output (EIMO);at least one pressure sensor;at least one exemplary reaction gas collection tank or bottle comprising a thermal sensor; andwherein said at least one electronic interface for matter output (EIMO) comprises at least one or more of:the at least one acoustic sensor or other type of electronic interface;at least one gas detector, orat least one reaction product sensor; andwherein said at least one electronic interface for matter output (EIMO) comprises being configured to at least one or more of:manage material output through said at least one separator valve; orfacilitate estimation of quantities of reaction gas being collected. 2. The gaseous electrolysis apparatus according to claim 1, wherein said actively-cooled header module is configured to enable the gaseous electrolysis apparatus to be operated for a period of time between maintenance periods, and wherein said actively-cooled header module comprises at least one or more of:a physically extended cooling manifold or water jacket, to improve thermal efficiency; orat least one feedthrough to a header module cooling manifold proximate to said actively-cooled header module. 3. The gaseous electrolysis apparatus according to claim 2, comprising said at least one feedthrough configured and constructed with at least one conductor surrounded by at least one insulating material to maintain integrity of electronics fed therethrough during a period of at least one or more of: a variable duration, or a long duration, of elevated pressure and temperature environment, andwherein said at least one feedthrough comprises at least one or more of:is welded into a thermal plate;wherein one end of said feedthrough extends beyond said actively-cooled header module for connection with the electronic control circuit module; orwherein said at least one feedthrough comprises a threaded coupling. 4. The gaseous electrolysis apparatus according to claim 1, wherein said actively-cooled header module comprises:a gasket to seal the actively-cooled header module to a top of a body of the reaction chamber module;at least one anode module connection;at least one anode module heater wire connection configured such that a heater is configured to raise a temperature of reaction material in the removable cathode module;at least one thermal sensor connection;at least one ceramic-encased microwave loop antenna configured to:facilitate transport of the gas electrolyte between an anode module and the removable cathode module; andsupport diffusion of the gas electrolyte into a reaction material; andat least one insulator configured to at least one or more of:electrically isolate;minimize a volume where the gas electrolyte resides; orprovide mechanical support for components of the heater module within the reaction chamber module. 5. The gaseous electrolysis apparatus according to claim 1, wherein themodular, removable anode module further compriseswherein edges at the ends of said modular, removable anode module facing the removable cathode module are tapered or curved to help prevent high voltage breakdown between the modular, removable anode module and the removable cathode module. 6. The gaseous electrolysis apparatus according to claim 1, wherein the removable cathode module comprises:a modular, removable, hollow-shaped, cylindrical removable cathode module, electrically coupled, connected, or grounded to the reaction chamber module with a central cavity configured to receive an anode module;wherein the removable cathode module is-encased by an outer metal supporting sleeve;wherein the removable cathode module is bounded at a base and at a top of the removable cathode module with at least one insulator endcap; andwherein edges at the ends of a reaction material part of the removable cathode module facing the anode module are at least one of tapered or curved to help prevent high voltage breakdown. 7. The gaseous electrolysis apparatus according to claim 1, wherein said heat exchanger module comprises at least a portion of a space surrounding the reaction chamber module, andis co-disposed around the reaction chamber module. 8. The gaseous electrolysis apparatus according to claim 1, wherein said gas handling system comprises: four (4) separate gas manifold modules, coupled to said reaction chamber module, said four separate gas manifold modules are configured to control gas flow external to the reaction chamber module, while serving to minimize gas volume external to said reaction chamber module,wherein said four separate gas manifold modules comprise:a hydrogen/deuterium gas supply manifold module, mechanically coupled to said reaction chamber module, and coupled to a hydrogen/deuterium supply gas supply source;an inert carrier gas manifold module, mechanically coupled to said reaction chamber module, and coupled to an inert gas supply source;a reaction gas product collection manifold module, mechanically coupled to said reaction chamber module, and coupled to a reaction gas product collector; anda gas measurement and evacuation manifold module, mechanically coupled to said reaction chamber module, and coupled to a gas measurement and evacuation device. 9. The gaseous electrolysis apparatus according to claim 8, wherein each of said four separate gas manifold modules, coupled to said reaction chamber module, comprises:a cooling chamber or water jacket to provide cooling for gas tubing and pipes connected or coupled to the reaction chamber module;at least one normally closed, gas compatible valve, at least one pressure sensor and at least one temperature sensor connected to, or coupled to said electronic control circuit module/subsystem; andat least one tank or at least one container whose known volume enables small quantities of gas to be determined by calculating pressure, temperature and volume before gas is transferred into or out of the reaction chamber module. 10. The gas electrolysis apparatus according to claim 8, wherein the reaction gas product collection manifold module comprises:the measurement container configured to temporarily store reactant gas and periodically permit extraction of the reactant gas from said measurement container; andthe subsystem comprising the acoustic sensor or the other type of electronic interface configured to facilitate estimation of quantities of the reactant gas. 11. The gaseous electrolysis apparatus according to claim 1, wherein said electronic control circuit (ECC) module comprises:an automated special-purpose computer and display monitor, and control software;automated gas handling system electronics, wherein said automated gas handling system electronics is coupled to, and related to associated electric valves, temperature sensors and pressure sensors;automated anode-to-cathode voltage/current supply;automated heater supply;automated microwave starter or initiator electronics; andautomated heat exchanger module electronics. 12. The gaseous electrolysis apparatus according to claim 1, wherein said gas handling system comprises at least one gas manifold module, coupled to said reaction chamber module, and is configured to control gas flow, external to the reaction chamber module, while serving to minimize gas volume external to said reaction chamber module,wherein said at least one gas manifold module comprises at least one or more of:a hydrogen/deuterium gas supply manifold module, mechanically coupled to said reaction chamber module, and coupled to a hydrogen/deuterium supply gas supply source;an inert carrier gas manifold module, mechanically coupled to said reaction chamber module, and coupled to an inert gas supply source;a reaction gas product collection manifold module, mechanically coupled to said reaction chamber module, and coupled to a reaction gas product collector; ora gas measurement and evacuation manifold module, mechanically coupled to said reaction chamber module, and coupled to a gas measurement and evacuation device. 13. The gaseous electrolysis apparatus according to claim 12, wherein said at least one gas manifold module, coupled to said reaction chamber module, comprises at least one or more of:a cooling chamber or water jacket to provide cooling for gas tubing and pipes connected or coupled to the reaction chamber module;at least one normally closed gas compatible valve, at least one pressure sensor and at least one temperature sensor connected to, or coupled to said electronic control circuit module/subsystem; orat least one tank or at least one container whose known volume enables small quantities of gas to be determined by calculating pressure, temperature and volume before gas is transferred into or out of the reaction chamber module. 14. The gaseous electrolysis apparatus according to claim 1, further comprising:wherein said modular, removable anode module compriseswherein said modular, removable anode module comprises wherein edges at the ends of said modular, removable anode module facing the removable cathode module are tapered or curved to help prevent high voltage breakdown between said modular, removable anode module and the removable cathode module; andwherein said electric heater module disposed within said modular, removable anode module comprises:wherein said electric heater module is configured to raise the temperature of the reaction material in the removable cathode module, by at least one of thermal radiation or diffusion; andwherein said electric heater module is electrically connected through a feedthrough in the header module to a power supply configured to provide power to said electric heater module. 15. The gaseous electrolysis apparatus according to claim 1, wherein the removable cathode module comprises at least one or more of:a modular, removable, hollow-shaped, cylindrical removable cathode module, electrically coupled, connected, or grounded to the reaction chamber module with a central cavity configured to receive an anode module;wherein the removable cathode module is encased by an outer metal supporting sleeve;wherein the removable cathode module is bounded at a base and at a top of the removable cathode module with at least one insulator endcap; orwherein edges at each ends of a reaction material part of the removable cathode module facing the anode module are at least one of tapered or curved away from the anode module to help prevent high voltage breakdown. 16. The gaseous electrolysis apparatus according to claim 1, wherein said heat exchanger module is modular andcomprises:a flash boiler configured to provide a mist of water or other coolant to the outer surface of the reaction chamber module; anda plurality of spray nozzles to cool at least one portion of the reaction chamber module and to facilitate and control thermal diffusion through the removable cathode module; and one or more ofthe at least one steam pressure port; orthe at least one thruster port configured to provide pressure output. 17. The gas electrolysis apparatus according to claim 1, wherein said reaction gas product collection manifold module further comprises at least one or more of:a measurement container, coupled to said reaction gas product collection manifold module, configured to temporarily store reactant gas and periodically permit extraction of the reactant gas from said measurement container; ora subsystem, coupled to said reaction gas product collection manifold module, comprising said acoustic interface, wherein said acoustic interface comprises an acoustic sensor configured to facilitate estimation of quantities of the reactant gas. 18. The gas electrolysis apparatus according to claim 1, wherein said reaction gas product collection manifold module further comprises:a measurement container configured to temporarily store reactant gas and periodically permit extraction of the reactant gas from said container; anda subsystem comprising at least one of:an acoustic sensor, or other type of electronic interface, said subsystem configured to facilitate estimation of quantities of the reactant gas. 19. The gaseous electrolysis apparatus according to claim 1, wherein said electronic control circuit (ECC) module comprises at least one or more of:an automated special-purpose computer and display monitor, and control software;automated gas handling system electronics, wherein said automated gas handling system electronics is coupled to, and related to associated electric valves, temperature sensors and pressure sensors;automated anode-to-cathode voltage/current supply;automated anode module heater supply;automated microwave starter or initiator electronics; orautomated heat exchanger module electronics. 20. A modular gaseous electrolysis apparatus, wherein an electrolyte comprises gas, the gaseous electrolysis apparatus comprising:an actively-cooled header module comprising at least one electrical connector or coupling, andwherein said actively-cooled header module is configured to be at least one or more of:opened,removed,closed, orinstalled;a heat exchanger module configured to:remove heat from a surface of a reaction chamber module; andfacilitate and control a thermal gradient through a removable cathode module, anda wall of the reaction chamber module,wherein said heat exchanger module is separate from the removable cathode module, andwherein the reaction chamber module is configured to receive the gas electrolyte;wherein said heat exchanger module is configured to be at least one or more of assembled, or disassembled, andwherein the gaseous electrolysis apparatus comprises at least one or more of:wherein the gaseous electrolysis apparatus is configured to include hermetic seals to maintain integrity in an elevated pressure and temperature environment;a flash boiler configured to provide a mist of water or other coolant to the outer surface of the reaction chamber module; ora plurality of spray nozzles configured to at least one or more of:cool at least one portion of the reaction chamber module;facilitate thermal diffusion; orcontrol thermal diffusion through the removable cathode module; andat least one or more of:at least one steam pressure port; orat least one thruster port configured to provide pressure output;a gas handling system, configured to provide the gas electrolyte, said gas handling system mechanically coupled to the reaction chamber module and separate from the reaction chamber module,wherein said gas handling system comprises at least one or more of:a measurement container configured to temporarily store reactant gas;a measurement container configured to permit extraction of the reactant gas; ora subsystem comprising an acoustic sensor or other type of electronic interface, configured to facilitate estimation of quantities of reactant gas;an electronic control circuit module electrically coupled or connected to said actively-cooled header module and said gas handling system and configured to electronically control said gas handling system;a modular, removable anode module comprising:an electric heater disposed within the modular, removable anode module;wherein said actively-cooled header module, said heat exchanger module, the removable cathode module, said modular, removable anode module, said electric heater, said gas handling system, and said electronic control circuit module are removably and mechanically coupled to the reaction chamber module; andwherein said gas handling system comprises:a gas manifold module that controls gas flow external to the reaction chamber module,wherein said gas manifold module comprises:a reaction gas product collection manifold module, mechanically coupled to the reaction chamber module, and coupled to a reaction gas product collector and further comprising:at least one mechanical valve;at least one electronically controlled valve;at least one separator valve with an electronic interface for matter output (EIMO);at least one pressure sensor;at least one exemplary reaction gas collection tank or bottle comprising a thermal sensor; andat least one electronic interface for matter output (EIMO) comprising at least one or more of:the at least one acoustic sensor or the other type of electronic interface;at least one gas detector, orat least one reaction product sensor; andwherein the at least one electronic interface for matter output (EIMO) comprises being configured to at least one or more of:manage material output through said at least one separator valve; orfacilitate estimation of quantities of reaction gas being collected.
043115586	claims	1. A pressure vessel for a nuclear power plant comprising a generally cylindrical shaped first partial section, a cavity for housing a high temperature reactor within said first partial section, a second partial section forming the lower part of the pressure vessel, a horizontal tunnel for housing a gas turbine assembly within said second partial section and a plurality of recesses for housing components of a heat exchanger apparatus wherein said cavity for housing the high temperature reactor is eccentrically arranged with respect to said generally cylindrical shape in said first partial section of said pressure vessel. 2. The pressure vessel of claim 1 comprising a reinforced concrete pressure vessel. 3. The reinforced concrete pressure vessel of claim 2 wherein said recesses extend in the longitudinal direction throughout said pressure vessel. 4. The reinforced concrete pressure vessel of claim 2 wherein said cavity is arranged above said partial section housing said horizontal tunnel. 5. The reinforced concrete pressure vessel according to claim 2 wherein said cavity for housing the high temperature reactor is eccentrically displaced in the direction of said horizontal tunnel from the center of the reinforced concrete pressure vessel. 6. The reinforced concrete pressure vessel according to claim 2, wherein said recesses for housing components of the heat exchanger apparatus are arranged on both sides of said horizontal tunnel and symmetrically perpendicular with respect to said tunnel. 7. The reinforced concrete pressure vessel according to claim 6, wherein said recesses are arranged to generally surround said cavity for housing the high temperature reactor. 8. The reinforced concrete pressure vessel according to claim 6, wherein said recesses and said cavity for housing the high temperature reactor are located in two adjacent areas of the reinforced concrete pressure vessel. 9. The reinforced concrete pressure vessel according to claim 8, wherein said reinforced concrete pressure vessel, in its area located underneath said cavity for housing the high temperature reactor, is offset transversely to the longitudinal axis of said horizontal tunnel. 10. The reinforced concrete pressure vessel according to claim 9, wherein said horizontal tunnel is located entirely within said second partial section of the pressure vessel and said recesses for housing the components of the heat exchanger apparatus extend vertically within said first and second partial sections of the pressure vessel. 11. The reinforced concrete pressure vessel according to claim 2, wherein said first and second partial sections of the reinforced concrete pressure vessel are prestressed to different degrees. 12. The reinforced concrete pressure vessel according to claim 11, wherein said prestressing in said first and second partial sections of the reinforced concrete pressure vessel is adapted to the pressures prevailing in said cavity, tunnel and recesses.
049873091	claims	1. Radiation therapy unit with a beam of rays propagating from a focal point along a beam axis and a radiator head arranged on the beam axis, having the following features: (a) the radiator head comprises a double-focus multi-leaf collimator; (b) the multi-leaf collimator exhibits a plurality of adjacently arranged diaphragm plates which in each case have two side faces, two front faces and an inside and an outside face; (c) means are provided for displacing each individual diaphragm plate; wherein (d) each side face of each diaphragm plate forms a part of a surface area of a cone, all such cones having both a common cone axis which extends perpendicularly to the beam axis through the focal point and a common cone point which coincides with the focal point, and (e) means are provided for guiding the diaphragm plates so that each diaphragm plate performs a pure rotation about the cone axis during its displacement. (a) the outside faces of all diaphragm plates arranged laterally adjacently have overall as an enveloping surface a part of a surface area of an outer cylinder the axis of which is the cone axis, and (b) the means for displacing the diaphragm plates engage the outside faces. (a) the inside faces of all diaphragm plates arranged laterally adjacently have overall as an enveloping surface a part of a surface area of an inner cylinder the axis of which is the cone axis, and (b) the means for guiding adjacently arranged diaphragm plates (23) comprise at least two holding yokes (24b, 24c) so that each diaphragm plate is supported at at least four points. (a) an adjusting aid attached to the outside face of the diaphragm plate (23), (b) a gear which engages this adjusting aid and (c) a drive actuating this gear. (a) the adjusting aid is a toothed rail mounted on the outside face, (b) the gear is a self-inhibiting worm-rack drive and (c) the drive is a stepping motor. (a) multi-leaf collimators have a reduced height, the reduced height of a single multi-leaf collimator not being sufficient but the sum of the heights of the multi-leaf collimators arranged one above the other being sufficient for attenuating the beam of rays to a required level and (b) a control circuit is provided which positions the individual diaphragm plates of the multi-leaf collimators in such a manner that a field point (34) to be shielded in the beam of rays is always covered by two diaphragm plates (23d, 23e). 2. Radiation therapy unit as claimed in claim 1, wherein 3. Radiation therapy unit as claimed in claim 1, wherein 4. Radiation therapy unit as claimed in claim 2, wherein means for displacing comprise for each diaphragm plate (23) 5. Radiation therapy unit as claimed in claim 4, wherein 6. Radiation therapy unit as claimed in claim 1, wherein the radiator head (3) exhibits two multi-leaf collimators which are arranged one above the other and are aligned perpendicularly with respect to one another. 7. Radiation therapy unit as claimed in claim 6, wherein the multi-leaf collimators have a reduced height, the reduced height of a single multi-leaf collimator not being sufficient but the sum of the heights of the multi-leaf collimators arranged above one another being sufficient for attenuating the beam of rays to a required level. 8. Radiation therapy unit as claimed in claim 1, wherein a matrix ionization chamber, by means of which the multi-leaf collimators are monitored, is arranged on the beam axis (S) opposite to the radiator head (3) and behind an isocenter. 9. Radiation therapy unit as claimed in claim 5, wherein the toothed rails of adjacent diaphragm plates of a multi-leaf collimator exhibit a different height and a resultant different pitch, the pitch being designed in such a manner that a given linear relationship which is valid for all diaphragm plates always exists between motor speed and plate advance. 10. Radiation therapy unit as claimed in claim 6, wherein the 11. Radiation therapy unit as claimed in claim 6 or 10, wherein the reduced height is about two thirds of the height needed for the required attenuation of the beam of rays.
	abstract	A system and method for managing spent nuclear fuel includes a small capacity canister that preferably encloses or encapsulates a single spent nuclear fuel rod assembly but can enclose up to six spent nuclear fuel rod assemblies. The canister is air tight and prevents radioactive material from escaping. The canister is loaded by positioning a single spent nuclear fuel rod assembly in the canister and then closing the canister to make it air tight.
	abstract	An exoskeleton frame having a frame body that includes a front torso member and a back torso member. The frame body further includes a first shoulder band and a second shoulder band each extending from the front torso member to the back torso member, and an opening positioned between the first shoulder band and second shoulder band. The exoskeleton frame further has an adjustable belt removably attached to the front torso and back torso member that is configured to direct a weight of apparel that is worn by the user over the exoskeleton frame to a weight-bearing area of the user located between knees and abdomen of the user, wherein the first shoulder band and second shoulder band are configured to support the weight of the apparel such that the weight of the apparel is not applied to a first shoulder or a second shoulder of the user.
	description	This application is a continuation of U.S. application Ser. No. 14/263,594, filed on Apr. 28, 2014, which is a continuation of U.S. application Ser. No. 13/696,931, filed on Nov. 8, 2012, which claims priority from International Application No. PCT/JP2010/063874, filed on Aug. 17, 2010, the contents of which are herein incorporated in their entirety by reference. The present invention relates to a multi-leaf collimator that is utilized in order to form an irradiation field in a particle beam therapy system utilizing a charged particle beam, relates to a particle beam therapy system utilizing the multi-leaf collimator, and relates to a treatment planning apparatus for determining the operation condition of the particle beam therapy system. In the particle beam therapy, therapy is implemented by irradiating a charged particle beam onto a diseased site, which is a therapy subject, so as to kill tissues of the diseased site; in order to deliver a sufficient dose to the tissues of the diseased site without causing damage to the peripheral tissues thereof, there is required a particle beam therapy system that can appropriately control an irradiation dose and irradiation coverage (referred to as an irradiation field, hereinafter). In a so-called broad-irradiation-type particle beam therapy system, among particle beam therapy systems, that utilizes an irradiation nozzle provided with a scanning electromagnet such as a wobbler electromagnet, the irradiation nozzle enlarges the irradiation field and a multi-leaf collimator that changes a penetration shape is disposed in the enlarged irradiation field, so that an irradiation field corresponding to the shape of a diseased site is formed. A multi-leaf collimator, in which two leaf lines formed of leaf plates laminated in the thickness direction are arranged in such a way as to face each other and the leaf plates are driven in a direction in which they approach each other or in a direction in which they are separated from each other, forms a predetermined penetration shape. Accordingly, by controlling the respective physical positions of the leaf plates, an irradiation field can readily be formed. However, in the case of a linear-driven leaf plate, in the contour portion that is away from the center of the irradiation field, a so-called penumbra is caused in which a charged particle beam having an angle toward the spreading direction hits part of the end face of the leaf plate and hence the dose of the charged particle beam is continuously attenuated. Thus, a so-called cone-shaped multi-leaf collimator has been proposed (e.g., refer to Patent Document 1 or 2) in which the spread of a beam is taken into consideration and leaves formed in a shape obtained through division at the side surface of an arc or a cone are driven on a circular orbit. [Patent Document 1] Japanese Patent Application Laid-Open No. S60-063500 (top right of Page 2, down right of Page 2 to top left of Page 3, and FIGS. 2 and 4) [Patent Document 2] Japanese Patent Application Laid-Open No. S63-225199 (down right of Page 3 to top right of Page 4, down left to down right of Page 7, FIGS. 1 through 3, and FIGS. 12 and 13) [Patent Document 3] Japanese Patent Application Laid-Open No. 10-255707 (Paragraphs 0009 through 0020, and FIGS. 1 and 5) [Patent Document 4] Japanese Patent Application Laid-Open No. 2006-166947 (paragraphs 0015 to 0016, and FIG. 1) However, in the case of the foregoing cone-shaped multi-leaf collimator, it is assumed that a beam spreads from a point light source. Alternatively, even in the case where it is assumed that the light source is a volume light source, it is not taken into consideration that the spreading manner differs depending on the direction. On the other hand, in order to enlarge an irradiation field in a particle beam therapy system that utilizes a charged particle beam, there is required an electromagnet for scanning a thin beam supplied from the accelerator, as described in Patent Documents 3 and 4. On top of that, there are required respective electromagnets such as an X-direction electromagnet and a Y-direction electromagnet for two directions in a plane perpendicular to the beam axis; thus, the spread starting point in the X direction and the spread starting point in the Y direction differ from each other. Accordingly, there has been a problem that even when the foregoing multi-leaf collimator is applied to a particle beam therapy system, the beam spreading manner and the penetration shape of the multi-leaf collimator do not coincide with each other and hence a penumbra remains. The present invention has been implemented in order to solve the foregoing problems; the objective thereof is to obtain a multi-leaf collimator and a particle beam therapy system in which a high-contrast irradiation field can be formed without undergoing the effect of a penumbra. A multi-leaf collimator according to the present invention is disposed in a particle beam that is irradiated so as to enlarge an irradiation field of that, in order to form the irradiation field so as to be conformed to an irradiation subject; the multi-leaf collimator is characterized in that there are provided a leaf row in which a plurality of leaf plates are arranged in the thickness direction thereof in such a way that the respective one end faces of the leaf plates are trued up and a leaf plate drive mechanism that drives each of the plurality of leaf plates in such a way that the one end face approaches or departs from a beam axis of the particle beam, in that in each of the leaf plates, a facing side facing a leaf plate that is adjacent to that leaf plate in the thickness direction is formed of a plane including a first axis that is perpendicular to the beam axis and is set at a first position on the beam axis, and in that the leaf plate drive mechanism drives the leaf plate along a circumferential orbit around a second axis that is perpendicular to the beam axis and the first axis and is set at a second position on the beam axis. A particle beam therapy system according to the present invention is characterized by including an irradiation nozzle that scans a particle beam supplied from an accelerator, by use of two electromagnets whose scanning directions are different from each other, and that irradiates the particle beam in such a way as to enlarge an irradiation field and the multi-leaf collimator, disposed in a particle beam irradiated from the irradiation nozzle, and characterized in that the multi-leaf collimator is disposed in such a way that the first axis coincides with the scanning axis of one of the two electromagnets and the second axis coincides with the scanning axis of the other one of the two electromagnets. A particle beam therapy system according to the present invention is characterized by including a three-dimensional data generation unit for generating three-dimensional data from image data on an irradiation subject, an irradiation condition setting unit that sets an irradiation condition, based on the generated three-dimensional data, and a control data generation unit that generates control data for controlling leaf driving for the multi-leaf collimator in the foregoing particle beam therapy system, based on the set irradiation condition, and characterized in that the three-dimensional data generation unit generates the three-dimensional data by utilizing at least a beam deflection angle with respect to the first axis and a beam deflection angle with respect to the second axis. In a multi-leaf collimator, a particle beam therapy system, and a treatment planning apparatus according to the present invention, the directions of the faces of leaf plates that configure a contour at a time when the multi-leaf collimator forms a penetration shape coincide with the directions of a particle beam spreading and passing through the vicinity of the faces; thus, a high-contrast irradiation field can be formed without undergoing the effect of a penumbra. The configurations of a multi-leaf collimator and a particle beam therapy system according to Embodiment 1 of the present invention will be explained below. FIGS. 1 through 5 are to explain the configurations of a multi-leaf collimator and a particle beam therapy system according to Embodiment 1 of the present invention. FIG. 1 is a view illustrating the configuration of an irradiation system, of a particle beam therapy system, that is provided with a multi-leaf collimator. FIG. 2 is a set of views for illustrating the configurations of a particle beam therapy system and a multi-leaf collimator when they are viewed from directions that are perpendicular to each other with respect to the center (z direction) of a charged particle beam in FIG. 1; FIG. 2(a) is a side view when viewed from the y direction; FIG. 2(b) is a side view when viewed from the x direction. FIG. 3 is to explain the shape of a beam bundle in an irradiation system of a particle beam therapy system; FIG. 3(a) is a view illustrating the overall appearance of a beam bundle; FIG. 3(b) and FIG. 3(c) are views of the beam bundle when viewed from directions that are perpendicular to each other with respect to the center (z direction) of a charged particle beam in FIG. 3(a); FIG. 3(b) is a side view when viewed from the y direction; FIG. 3(c) is a side view when viewed from the x direction. Each of FIGS. 4 and 5 is a set of views for explaining the configurations of a multi-leaf collimator and a leaf plate, which is a main constituent member of the multi-leaf collimator, when viewed from various directions. At first, as an assumption for a detailed explanation about the configuration of a multi-leaf collimator, an irradiation system, of a particle beam therapy system, that includes a multi-leaf collimator will be explained. As illustrated in FIGS. 1 and 2, a particle beam therapy system 10 is provided with a wobbler electromagnet 1 (an upstream wobbler electromagnet 1a, a downstream wobbler electromagnet 1b) that functions as an irradiation nozzle for enlarging an irradiation field by circularly scanning the charged particle beam B, which is supplied from an unillustrated accelerator and has a so-called pencil-looking shape; a ridge filter 2 for enlarging the width of a Bragg peak in accordance with the thickness of an irradiation subject; a range shifter 3 for changing the energy (range) of the charged particle beam B in accordance with the depth (irradiation depth) of the irradiation subject; a block collimator 4 for limiting the enlarged irradiation field to a predetermined area so as to prevent superfluous irradiation onto normal tissues; a multi-leaf collimator 5 that is configured with a plurality of leaf plates and a leaf drive mechanism for driving each of the leaf plates and that limits an irradiation field in such a way that the irradiation field coincides with the shape of a diseased site; and a bolus 6 that limits the range of the charged particle beam B in such a way that the range coincides with the depth-direction shape of an irradiation subject. Next, there will be explained the operation and the principle of an irradiation system that enlarges an irradiation field by means of an irradiation nozzle in which the Wobbling method is utilized. The charged particle beam B is accelerated by the unillustrated accelerator; then, as a so-called pencil beam having a diameter smaller than several millimeters, it is introduced to the irradiation system through a transport system. The beam introduced to the irradiation system is scanned by the wobbler electromagnet 1 in such a way as to draw a circular orbit. As illustrated in FIG. 1 or 2, the wobbler electromagnet 1 is usually provided with an x-direction electromagnet 1a and a y-direction electromagnet 1b; the two electromagnets are arranged in such a way as to be superimposed on each other along the center axis XB of the charged particle beam B. Here, for clarity of description, the x direction and the y direction will be defined. In various specifications, coordinate systems are defined; however, in this “DESCRIPTION”, the coordinate system is defined in the following manner. The direction in which the charged particle beam B travels is defined as the positive direction of the z axis. The x axis and the y axis are axes that are perpendicular to the z axis; the x axis and the y axis are also perpendicular to each other. Then, the xyz coordinate system is established in such a way as to be a right-handed coordinate system. In each of the examples in FIGS. 1 and 2, the upstream wobbler electromagnet 1a and the downstream wobbler electromagnet 1b scan a beam in the x direction and in the y direction, respectively. Due to the scanning by the electromagnets 1a and 1b, the irradiation field is expanded in the xy direction (planar direction). The charged particle beam B whose irradiation field has been enlarged passes through the ridge filter 2. A ridge filter is formed, for example, in such a way that a great number of cones or plates whose cross sections are triangles are arranged on a plane; assuming that, for example, an irradiation field is divided into a great number of sub-areas, in which there exist beams that pass through different thicknesses from one another. For easier understanding, FIG. 1 or 2 illustrates cones that are arranged as in a pin holder (“kenzan”). In such a manner as described above, the width of a Bragg peak SOBP (Spread-Out Bragg Peak) is enlarged. That is to say, the ridge filter 2 enlarges the irradiation field also in the z direction. Next, the charged particle beam B whose irradiation field has been enlarged passes through the range shifter 3. The range shifter 3 is a device that changes the energy of the charged particle beam B. Due to the range shifter 3, the enlarged irradiation field can be irradiated onto a position of a desired inner-body depth. Next, the beam that has passed through the range shifter 3 passes through the block collimator 4. The block collimator 4 is, for example, a metal block in which a passing hole PH is provided, and limits the planar-direction (the xy plane) spread of the irradiation field. This is because superfluous irradiation onto normal tissues can be prevented by preliminarily limiting the irradiation coverage. Next, the charged particle beam passes through the multi-leaf collimator 5. As described later, through a penetration shape PS formed based on the positions of a plurality of leaves 5L, the multi-leaf collimator 5 limits the shape of the irradiation field in accordance with the shape of the diseased site. That is to say, the multi-leaf collimator 5 performs limitation and formation of the irradiation field in the xy direction. The multi-leaf collimator 5 is provided with at least the plurality of leaf plates 5L (collectively referred as a leaf group 5G) and a leaf drive mechanism 5D. However, the configuration of the leaf drive mechanism 5D itself is not important, as long as the driving orbit of a leaf can be specified. If the leaf drive mechanism 5D itself is drawn in a figure, it becomes difficult to illustrate the arrangement of the leaf plates 5L; therefore, in FIGS. 1, 2, and thereafter, for the sake of simplicity, only a leaf plate 5L or only the leaf group 5G in which the leaf plates 5L are integrated, out of the multi-leaf collimator 5, is illustrated. Lastly, the charged particle beam B passes through the bolus 6. The bolus 6 is a limiter that is formed of resin or the like; it is formed in such a shape as to compensate the depth-direction shape of a diseased site, for example, the distal shape of the diseased site. The distal shape denotes the depression-protrusion contour of the deepest portion. In this situation, the energy of the irradiation field is limited (formed in the z direction) to have a shape the same as the distal shape. That is to say, the bolus 6 performs limitation and formation of the irradiation field in the z direction. The function of the irradiation system of a particle beam therapy system is to form an irradiation field in accordance with the diseased site onto which a beam is irradiated. In the Wobbling method that is adopted, as the method therefor, in a particle beam therapy system according to Embodiment 1, an irradiation field is enlarged only by the wobbler electromagnet 1. For example, the “large-area uniform irradiation method through spiral beam scanning” disclosed in Patent Document 3 is a specific example of this method, which is referred to as the “spiral Wobbling method”, among the Wobbling methods. Briefly speaking, the spiral Wobbling method is to scan a beam in a spiral manner so as to enlarge an irradiation field; the scanning orbit (scanning locus) in the irradiation field is contrived so that the flatness is secured. Additionally, a beam scanning orbit according to the spiral Wobbling method can be seen in FIG. 1 and the like of Patent Document 1. Meanwhile, in general, the method which is referred to as the “Wobbling method” often signifies the “single-circle Wobbling method”; in that case, when an irradiation field is enlarged, the flatness is secured by means of a scatterer. Therefore, among the Wobbling methods, there exist not only a Wobbling method in which a scatterer is utilized but also a Wobbling method in which no scatterer is utilized; thus, the directional behavior of a beam differs depending on whether or not there exists a scatterer. In the case where a scatterer is utilized, the beam spreads on the whole surface of the scatterer; thus, there exists a width in the irradiation direction of a beam that passes through a given point. In contrast, in the case where as the spiral Wobbling method, a beam is enlarged only by means of a scanning electromagnet without utilizing any scatterer, the irradiation direction of the beam that passes through a given point is a single direction that is determined mainly by the position thereof with respect to the scanning electromagnet. FIG. 3 is a set of schematic diagrams illustrating the spreading manner (the shape of a beam bundle FB) in which a beam is enlarged by the couple of scanning electromagnets 1 in the irradiation system of the particle beam therapy system 10 according to Embodiment 1. In the spiral Wobbling method, the beam is enlarged not in a point-light-source manner but in such a manner as illustrated in FIG. 3. For the sake of simplicity, the spreading manner of the beam, illustrated in FIG. 3, will be referred to as a “series-of-scanners spreading manner”. In the case where a beam is enlarged not in a point-light-source manner but in a series-of-scanners spreading manner, a limiter suitable therefor needs to be designed. The series-of-scanners spread will be explained in more detail hereinafter. As illustrated in FIG. 3, the beam B is irradiated from the top to the bottom (in the z direction). Originally, the beam B is supplied as a thin beam, which is called a pencil beam. Reference points CPa and CPb are set on the beam axis XB. The reference point CPa may be regarded as a position where the upstream wobbler electromagnet 1a (strictly speaking, a scanning axis ASa) is disposed; similarly, the reference point CPb may be regarded as a position where the downstream wobbler electromagnet 1b (strictly speaking, a scanning axis ASb) is disposed. The upstream wobbler electromagnet 1a disposed at the reference point CPa scans the beam B with respect to the reference point CPa. The scanning direction, of the upstream wobbler electromagnet 1a, in which the beam B is scanned is on a plane (the xz plane) of FIG. 3(b) and passes through the reference point CPa on the beam axis XB; the axis ASa, which is perpendicular to the beam axis XB, is the action axis (scanning axis) of the upstream wobbler electromagnet 1a. The downstream wobbler electromagnet 1b disposed at the reference point CPb scans the beam B with respect to the reference point CPb. The scanning direction, of the downstream wobbler electromagnet 1b, in which the beam B is scanned is on a plane (the yz plane) of FIG. 3(c) and passes through the reference point CPb on the beam axis XB; the axis ASb, which is perpendicular to the beam axis XB and the axis ASa, is the action axis (scanning axis) of the downstream wobbler electromagnet 1b. In other words, the scanning direction (x) of the upstream wobbler electromagnet 1a and the scanning direction (y) of the downstream wobbler electromagnet 1b are perpendicular to the beam axis XB; the scanning direction (x) of the upstream wobbler electromagnet 1a and the scanning direction (y) of the downstream wobbler electromagnet 1b are perpendicular to each other. Furthermore, the shape of the beam bundle FB will geometrically be explained with reference to FIG. 3. As illustrated in FIG. 3(b), there is drawn a vertical (z-direction) line whose top end point is the reference point CPa, and then the reference point CPb is provided at a position other than the reference point CPa on the vertical line. There is obtained a sector Fsa through which the line passes when the line is pivoted by ±α° with respect to the reference point CPa. In the case where only the upstream wobbler electromagnet 1a is utilized, the sector Fsa corresponds to the spread of the beam. Next, the sector Fsa is divided into the upper part and the lower part by the reference axis ASb that passes through the reference point CPb. There is obtained a region through which the lower part of the sector Fsa passes when the lower part of the sector Fsa is pivoted by ±β with respect to the reference axis ASb. This region is recognized as a sector Fsb in FIG. 3(c) and represents the spreading manner (the region through which the beam B can pass: the beam bundle FB) of the beam B. That is to say, the shape of the beam bundle FB having a series-of-scanners spread is a sector whose x-direction and y-direction curvature radiuses are different from each other. By considering the shape of the beam bundle FB having a series-of-scanners spread that is caused by enlarging an irradiation field by means of two scanning electromagnets 1a and 1b whose scanning directions are different from each other, as described above, the multi-leaf collimator 5 according to Embodiment 1 of the present invention is configured for the purpose of accurately forming a high-contrast irradiation field without undergoing the effect of a penumbra. In other words, in the multi-leaf collimator 5 according to Embodiment 1 of the present invention, each of the leaf plates 5L is configured in such a way that the substantial facing side PL facing the adjacent leaf plate in the thickness direction is formed of a plane including the scanning axis Asa, of the scanning electromagnet 1a, that is set at the reference point CPa on the beam axis XB of the charged particle beam B, and each of the leaf plates 5L is driven along a circumferential orbit with respect to the scanning axis Asb), of the scanning electromagnet 1b, that is set at the reference point CPb on the beam axis XB and is perpendicular to the beam axis XB and the scanning axis Asa. Detailed explanation will be made below with reference to FIGS. 4 and 5. FIG. 4 is a set of views for explaining the configurations of a multi-leaf collimator and leaf plates to be driven in the multi-leaf collimator, when the leaves are all closed; FIG. 4(a) is an appearance perspective view of all the leaf group of the multi-leaf collimator; FIG. 4(b) is a top perspective view of the multi-leaf collimator when viewed from P direction in FIG. 4(a); FIG. 4(c) is a front perspective view of the multi-leaf collimator when viewed from F direction in FIG. 4(a); FIG. 4(d) is a side perspective view of a row of leaves in the left-half portion of the multi-leaf collimator, when viewed from S direction in FIG. 4(a). FIG. 5 is a set of views illustrating the state where an irradiation field having a predetermined shape is formed; FIG. 5(a) is an appearance view of all the leaf group of a multi-leaf collimator; FIG. 5(b) is a top perspective view of the multi-leaf collimator when viewed from P direction in FIG. 5(a); FIG. 5(c) is a front perspective view of the multi-leaf collimator when viewed from F direction in FIG. 5(a); FIG. 5(d) is a side perspective view of a row of leaves in the left-half portion of the multi-leaf collimator, when viewed from S direction in FIG. 5(a). As illustrated in FIGS. 4 and 5, the multi-leaf collimator 5 is provided with a leaf group 5G that has two leaf rows (5c1 and 5c2: collectively referred to as 5c), in each of which a plurality of leaf plates 5L are arranged in the thickness direction (x direction) in such a way that the end faces EL thereof are trued up, and in which the leaf rows 5c1 and 5c2 are arranged in such a way that the respective end faces EL thereof face each other and with an unillustrated leaf plate drive mechanism that drives each of the leaf plates 5L in a direction in which that leaf plate approaches the opposed leaf plate or in a direction in which that leaf plate departs from the opposed leaf plate; as the shape of the leaf plate 5L, the substantial shape of the main face as the plate material of each leaf plate, i.e., the facing side PL facing the adjacent leaf plate is formed of a plane including the scanning axis Asa of the scanning electromagnet 1a that enlarges the charged particle beam B in the x direction. In other words, the main plane as the plate material is formed of two planes including the scanning axis Asa of the scanning electromagnet 1a; the cross-section of the leaf plate, obtained by cutting the leaf plate along the plane including the irradiation direction and the board-thickness direction, becomes thicker in a direction from the upstream side of the irradiation direction of the charged particle beam B to the downstream side thereof. The drive (in the yz-plane direction) of the leaf plates 5L is set to be an circumferential orbit OL corresponding to a distance Rsb from the scanning axis Asb of the downstream electromagnet 1b that enlarges the charged particle beam B in the y direction, and the shapes of an incident-side end face PI that is adjacent to the end face EL and an emitting-side end face PX, among the four end faces of the leaf plate 5L, are each formed of an arc whose center is the scanning axis Asb), i.e., each formed as if it is part of a ring whose center is the scanning axis Asb), so that even when the leaf plate 5L is driven along the circumferential orbit OL, the thickness dimension along the irradiation direction of the charged particle beam B does not change. Because of the foregoing configuration, in whichever position the leaf plate 5L is driven, for example, as illustrated in FIG. 5, the end face EL of the leaf plate 5L that forms the x-direction contour of the penetration shape PS is in parallel with the irradiation direction of the charged particle beam B that passes through the vicinity of the end face EL, whereby no penumbra is caused. The facing side PL of the leaf plate 5L that forms the y-direction contour of the penetration shape PS is in parallel with the irradiation direction of the charged particle beam B that passes through the vicinity of the facing side PL, whereby no penumbra is caused. In other words, no penumbra is caused in any contour portion of the penetration shape PS formed by the multi-leaf collimator 5; therefore, an accurate irradiation field suited to the shape of a diseased site can be formed. That is to say, it is only necessary that the thickness-direction shape and the driving orbit OL of the leaf plate 5L in the multi-leaf collimator 5 according to Embodiment 1 of the present invention form a shape the same as the spread of the beam bundle FB of the charged particle beam B. That is to say, the spread is the passable range at a time when the respective scanning angles of the couple of scanning electromagnets 1a and 1b are limited. Furthermore, the spread is the position of a charged particle beam at a time when the beam propagation distance from the beam source is within a given range. Because the multi-leaf collimator 5 is obtained by laminating the leaf plates 5L, the formed penetration shape PS is also the spread shape of the beam bundle FB of a charged particle beam. Moreover, because of the foregoing configuration, whatever the shape of the opening (contour) that forms the penetration shape PS is, the end face EL, of the leaf plate 5L, that is a wall face of the opening and faces the center of the irradiation field and the facing side PL facing the adjacent leaf plate are in parallel with the irradiation direction of a charged particle beam that passes through the vicinity of those faces. Accordingly, the problem of a penumbra, caused when a couple of scanning electromagnet 1a and 1b are utilized, can be solved. In the case where irradiation is implemented by use of a scatterer for the purpose of raising the flatness, a range in the distribution of the irradiation directions is caused by the foregoing series-of-scanners manner. Accordingly, because even in the case where the multi-leaf collimator 5 is utilized, some of charged particle beams hit the end face EL of the leaf plate or the facing side PL, the effect of suppressing a penumbra is reduced in comparison with the case where no scatterer is utilized; however, it is made possible to obtain a larger effect of suppressing a penumbra in comparison with a simple cone-shaped conventional multi-leaf collimator. In the multi-leaf collimator 5 according to Embodiment 1, the thickness-direction shape and the driving orbit OL are set based on the position of the upstream electromagnet 1a and the position of the downstream electromagnet 1b, respectively; however, the present invention is not limited thereto. They may be set on the opposite positions. Accordingly, there has been described that the upstream electromagnet 1a and the downstream electromagnet 1b scan a beam in the x direction and in the y direction, respectively; however, they may scan a beam in an opposite manner. Although the drawings illustrate that the angles, between the facing sides PL are uniform, that specify the thickness of the leaf plate 5L; however, the present invention is not limited thereto. Even when the angles are not uniform, it is made possible to obtain the effect of suppressing a penumbra. The reason why the expression “substantial” is utilized for the facing side is to mean that the facing side is a side for distinguishing it from the leaf that is substantially adjacent to it when the leaves are laminated in the thickness direction; for example, even when a groove or a recess for forming a driving rail is provided in the facing side, it is understood that the facing side is formed of a plane including the scanning axis Asa of the scanning electromagnet 1a set at the reference point CPa. The drawings illustrate the state where one of the leaves 5L of the leaf row 5c1 and one of the leaves 5L of the leaf row 5c2 make a pair; however, they do not necessarily need to make a pair. The number of the leaf rows does not need to be two; for example, even when the number of the leaf rows is one, it is only necessary that when the end face EL of the leaf plate becomes closest to the beam axis XB, the end face EL adheres to the fixed side so as to block the beam B. The number of the leaf rows may be more than two. As a method of enlarging an irradiation field, there has been explained a spiral Wobbling method in which a scanning locus becomes a spiral; however, as explained in the following embodiments, another spiral Wobbling method may be utilized, and the method may not be limited to a spiral Wobbling method. Moreover, the electromagnet that functions as an irradiation nozzle is not limited to the wobbler electromagnet 1; it is only necessary that the irradiation nozzle is to enlarge an irradiation field by means of two electromagnets whose scanning directions are different from each other. As described above, the multi-leaf collimator 5 according to Embodiment 1 is disposed in the charged particle beam B that is irradiated by use of the scanning electromagnet 1 so as to enlarge an irradiation field of that, in order to form the irradiation field so as to be conformed to the shape of a diseased site, which is an irradiation subject; the multi-leaf collimator 5 is provided with the leaf row 5c in which a plurality of leaf plates 5L are arranged in the thickness direction in such a way that the end faces EL thereof are trued up and with the leaf plate drive mechanism 5D that drives each of the leaf plates 5L in such a way that the end face EL thereof approaches or departs from the beam axis XB of the particle beam B or that drives each of the leaf plates 5L in a direction in which that leaf plate 5L approaches the opposed leaf plate or in a direction in which that leaf plate 5L departs from the opposed leaf plate. In each of the leaf plates 5L, the facing side PL facing a leaf plate that is adjacent to that leaf plate in the thickness direction (x direction) is formed of a plane Psa including the scanning axis Asa, which is a first axis perpendicular to the beam axis XB and is set at the reference point CPa that is a first position on the beam axis XB of the charged particle beam B; the leaf plate drive mechanism 5D drives the leaf plate 5L along the circumferential orbit OL around the scanning axis Asb, which is a second axis perpendicular to the beam axis XB and the first axis Asa, set at the reference point CPb that is a second position on the beam axis XB. As a result, the spreading manner of the beam bundle FB of the charged particle beam B and the directions of the facing side PL and the end face EL that form the contour of the penetration shape PS of the multi-leaf collimator 5 coincide with each other, so that the effect of a penumbra is suppressed and hence an accurate irradiation field conforming to the shape of an irradiation subject can be formed. Furthermore, the shapes of the end face PI at the incident side of the charged particle beam B and the end face PX at the emitting side thereof that are adjacent to the end face EL, among the main four end faces of the leaf plate 5L, are formed in the shape of an arc whose center is the scanning axis Asb), which is the second axis; therefore, the leaf plate 5L can readily be driven along the circumferential orbit OL. And whichever position the leaf plate 5L is driven, the depth dimension along the irradiation direction of the charged particle beam B does not change; therefore, the distance for shutting off the charged particle beam becomes constant. The particle beam therapy system 10 according to Embodiment 1 of the present invention is provided with the wobbler electromagnet 1, which is an irradiation nozzle that scans the charged particle beam B supplied from an accelerator, by use of two electromagnets 1a and 1b whose scanning directions are different from each other and irradiates the charged particle beam B in such a way as to enlarge an irradiation field, and the foregoing multi-leaf collimator 5 that is disposed in the charged particle beam B (the beam bundle FB thereof) irradiated from the irradiation nozzle 1; the multi-leaf collimator 5 is disposed in such a way that the first axis thereof coincides with the scanning axis (Asa or Asb) of one of the two electromagnets and the second axis thereof coincides with the scanning axis (Asb or Asa) of the other electromagnet. Therefore, the effect of a penumbra is suppressed and hence a charged particle beam can be irradiated with an accurate irradiation field conforming to the shape of an irradiation subject. In Embodiment 1, there has been described the application of a multi-leaf collimator according to the present invention to the spiral Wobbling method in which a beam is scanned in a spiral manner. However, the technical idea of the present invention is not limited to the foregoing scanning orbit shape (scanning locus) in the irradiation field of a beam; the effect of the present invention is demonstrated even in the case of other beam scanning loci, as long as the spreading manner is a series-of-scanners manner. Thus, in Embodiment 2, there will be described a case where a multi-leaf collimator according to the present invention is applied to an irradiation system having another typical beam scanning locus. At first, there will be explained a beam scanning locus produced through the spiral Wobbling method utilized in Embodiment 1. As disclosed in Patent Document 3, the spiral scanning locus is given by the equation (1) including the following three equalities. r ⁡ ( t ) = R max - R min π ⁢ ⁢ N ⁢ v 0 ⁢ t + R min 2 ⁢ ⁢ ω ⁡ ( t ) = v 0 R max - R min π ⁢ ⁢ N ⁢ v 0 ⁢ t + R min 2 ⁢ ∴ θ ⁡ ( t ) = θ ⁡ ( 0 ) + ∫ 0 t ⁢ ω ⁡ ( τ ) ⁢ ⁢ ⅆ τ ( 1 ) where Rmin is the radius at a time when the time t=0, Rmax is the radius at a time when the time t=T, and N is the scanning rotation speed. In addition, r(t) is the radial-direction coordinates, and θ(t) is the angle-direction coordinates; r(t) and θ(t) are represented through a polar coordinate system. The shape of the beam scanning locus given by the equation (1) is spiral; the shape is effective in obtaining a uniform dose distribution by scanning a beam within a circular region. However, it is not required that in order to obtain a uniform dose distribution, the beam scanning locus is limited to a spiral locus. It is conceivable that the beam scanning loci for obtaining a uniform dose distribution through scanning by two electromagnets can be categorized into a number of typical patterns. The Wobbling method is to form a uniform dose distribution by continuously scanning a beam. That is to say, it is desirable that the beam scanning locus in the Wobbling method is continuous and periodical. Thus, there has been studied a pattern in which a beam orbit is represented by a polar coordinate system and r(t) and θ(t) are continuously and periodically changed. <Typical Pattern-1> In the first pattern, r(t) and θ(t) are each defined as a function that changes continuously and periodically, as described below. r(t)=continuous and periodic function (period: T1) θ(t)=continuous and periodic function (period: T2) In this situation, the respective periods of r(t) and θ(t), which are different from each other, may be utilized. Attention should be drawn to the fact that as for the angle θ, 360° can be regarded as 0° as it rotates once. In other words, 360° continues to 0°. When represented in radian, 2π can be regarded as 0. Examples that realize the foregoing pattern include such a beam scanning locus as represented by the equation (2) including the following three equalities.r(τ)=r1+r2 sin(ωrτ+φr)θ(τ)=ωθττ=τ(t) (2)where τ(t) is the parameter of the equation (2) that is represented by utilizing a parameter, and is the function of the time. ωr is the angular velocity that determines r(t), and the period of r(t) is 2π/ωr. φr is the initial phase. ωθ is the angular velocity that determines θ(t), and the period of θ(t) is 2π/ωθ. FIG. 6 represents an example of beam scanning locus ST1 generated according to the equation (2). FIG. 6 represents a scanning locus on a given plane that is perpendicular to the beam axis; the abscissa denotes “x” and the ordinate denotes “y”; x and y are each normalized. The reason why in the equation (2), the parameter is not the time is that it is required to make the drawing speed changeable depending on the place. For example, in FIG. 6, beam scanning concentrates in the vicinity of the center of the beam axis represented as the coordinates (0, 0); thus, in a portion in the vicinity of the center portion where the locus concentrates, contrivance such as raising the scanning speed is made so that a uniform dose distribution is obtained. <Typical Pattern-2> In the second pattern, two or more functions for defining a drawing pattern are combined so that a beam scanning locus is formed. For example, a function for drawing a large circle is combined with a function for drawing a small circle. An example is represented by the equation (3) including the following three equalities.x(τ)=r1 cos(ω1τ+φ1)+r2 cos(ω2τ+φ2)y(τ)=r1 sin(ω1τ+φ1)+r2 sin(ω2τ+φ2)τ=τ(t) (3)where x(τ) and y(τ) are the x coordinate and the y coordinate, respectively, of a beam scanning locus; they are represented by use of an orthogonal coordinates system. FIG. 7 represents an example of beam scanning locus generated according to the equation (3). As is the case with FIG. 6, FIG. 7 represents a scanning locus on a given plane that is perpendicular to the beam axis; the abscissa denotes “x” and the ordinate denotes “y”; x and y are each normalized. Among toys, there exists a tool in which a gear-shaped disk is disposed in a circular hole inside of which teeth are formed; a geometrical pattern is drawn by inserting a pen tip into a small hole provided at a predetermined position in the disk and rolling the disk along the circular hole. A geometrical pattern generated with the tool also belongs to this category. A curve drawn with this tool is referred to as a hypotrochoid; geometrically, the curve is defined as a locus drawn by a fixed point that is lr away from the center of a circle of a radius r when the circle of a radius r rolls without sliding along the inner circumference of a circle of a radius kr. In many mixing devices, the curve is adopted as the driving pattern for a mixing unit. The reason why the parameter is not the time t is that it is required to make the drawing speed changeable depending on the place, as is the case with the above example. As described above, in the method in which through a wobbler electromagnet, a continuous and periodical pattern (line drawing) is drawn, the pattern is not necessarily a spiral. However, the idea in which by utilizing no scatterer but by contriving a beam orbit, large-area uniform irradiation is realized originates in the “spiral Wobbling method”; therefore, in some cases, each of these methods described in Embodiment 2 is also referred to as a broad-sense spiral Wobbling method. In addition, also in these broad-sense spiral Wobbling methods, a beam spreads not in a point-light-source manner but in a series-of-scanners manner. In other words, also in the particle beam therapy system having an irradiation system utilizing the broad-sense spiral Wobbling method according to Embodiment 2, by utilizing the multi-leaf collimator described in Embodiment 1, the thickness-direction shape of the leaf plate and the driving orbit can be made the same as the spread of the beam bundle FB of the charged particle beam B. Accordingly, the formed penetration shape PS becomes the same as the shape of the beam bundle FB of the charged particle beam B; thus, whatever the shape of the opening that forms the penetration shape PS is, the end face that is a wall face of the opening and faces the center of the irradiation field and the facing side facing the adjacent leaf plate coincide with the irradiation direction of a charged particle beam. Accordingly, the problem of a penumbra, caused when a couple of scanning electromagnets are utilized, can be solved. In each of Embodiments 1 and 2, there has been described a case where a multi-leaf collimator is applied to irradiation through the Wobbling method. However, as described above, the irradiation method itself is not essential and does not define the technical idea of the present invention. With regard to a particle beam therapy system, there has been proposed a spot-scanning method in which a charged particle beam is scanned by means of a couple of scanning electromagnets, and a spot is irradiated onto an irradiation subject in a pointillism manner. Also in the spot-scanning method, a beam spreads in a series-of-scanners manner. Therefore, in the case where a multi-leaf collimator is utilized in spot scanning, there is demonstrated an effect that a penumbra is suppressed and a high-contrast irradiation field is formed. In Embodiment 3, there has been described the application of a multi-leaf collimator according to the present invention to the spot-scanning method. There exists a raster-scanning method in which a charged particle beam is scanned by means of a couple of scanning electromagnets, as is the case with a spot-scanning method, and raster irradiation is performed onto an irradiation subject in a one-stroke writing manner. Also in the raster-scanning method, a beam spreads in a series-of-scanners manner. Therefore, in the case where a multi-leaf collimator is utilized in the raster-scanning method, the multi-leaf collimator 5 according to the foregoing embodiment demonstrates an effect. In other words, also in the case where an irradiation field is enlarged through a scanning method such as the spot-scanning method or the raster-scanning method, when the multi-leaf collimator 5 according to the embodiments of the present invention is utilized, there is demonstrated an effect that a penumbra is suppressed and a high-contrast irradiation field is formed. There has been proposed a particle beam therapy system in which, for example, as disclosed in Patent Document 4, one of two scanning electromagnets is omitted, by contriving control method for a deflection electromagnet. However, even in the case of such an irradiation system, a deflection electromagnet for changing the orbit direction (the direction of the beam axis) replaces the omitted scanning electromagnet that scans a charged particle beam; therefore, the beam bundle has a series-of-scanners spread, whereby the multi-leaf collimator according to each of the foregoing embodiments demonstrates an effect of suppressing a penumbra. FIG. 8 is a view illustrating an irradiation system including a multi-leaf collimator in a particle beam therapy system according to Embodiment 5. In FIG. 8, the beam axis of a charged particle beam B supplied in the horizontal direction (the x direction) is deflected to the vertical direction by a deflection electromagnet 201a and passes through a scanning electromagnet 201b; then, as is the case in Embodiment 1, the charged particle beam B is irradiated onto an irradiation subject, by way of a ridge filter 2, a range shifter 3, a ring collimator 4, a multi-leaf collimator 205, and the bolus 6. The configuration of a particle beam therapy system 210 according to Embodiment 5 is the same as that of Embodiment 1, excluding the fact that instead of the scanning electromagnet 1a in the particle beam therapy system 10 according to Embodiment 1, the deflection electromagnet 201a is provided and that the setting reference for the shape and the orbit of the leaf plate of the multi-leaf collimator 205 is different. In FIG. 8, inside the deflection electromagnet 201a, the charged particle beam B supplied in the horizontal direction is deflected in the z direction, while the beam axis PX draws an arc. In this situation, in the case of a normal deflection electromagnet, because control is performed in such a way that the magnetic field becomes constant, the beam bundle of the charged particle beam B does not spread; however, by periodically changing the magnetic field, the deflection electromagnet 201a scans the charged particle beam B in the x direction so that the beam bundle can spread in the x direction from PE1 to PE2. In other words, the deflection electromagnet 201a plays the role of the upstream scanning electromagnet 1a of Embodiment 1. The portion thereafter is basically the same as Embodiment 1; the scanning electromagnet 201b further spreads the beam bundle, which has been spread in the x direction, in the y direction. This beam spreading manner can be regarded as a spreading manner at a time when the scanning axis of the upstream scanning electromagnet 201a exists at an equivalent reference point EAS in FIG. 8 and a beam, irradiated from the upper side along the beam axis Ex, is scanned in the x direction (including the z-direction component) and spreads in the x direction from EE2 to EE2. Because inside the deflection electromagnet 201a, the beam axis is gradually deflected as the beam advances, the beam axes (=beam axis Ex) at the entrance side and at the exit side are different from each other; thus, a scanning axis EAs exists off the deflection electromagnet 201a. However, because the axis of a beam that enters the multi-leaf collimator 205 is the beam axis EX, the reference point CPa that specifies the position of the scanning axis EAs can be regarded as existing on the beam axis of the beam that enters the multi-leaf collimator 205, as a manner of thinking; therefore, the scanning axis EAs can also be regarded as being perpendicular to the beam axis of the beam that enters the multi-leaf collimator 205. Accordingly, also in an irradiation system in which one of the electromagnets that perform scanning also plays the role of a deflection electromagnet, it may be allowed that the equivalent scanning axis EAs is calculated based on the manner of beam spreading with respect to the beam axis of the beam that enters a multi-leaf collimator, and as is the case in Embodiment 1, the shape of the leaf plate of the multi-leaf collimator 205 and the orbit are set based on the equivalent scanning axis EAs and the scanning axis Asb (the reference point CPb). As can be seen from FIG. 8, in the case of an irradiation system in which one of the scanning electromagnets is omitted and instead of the omitted scanning electromagnet, the deflection electromagnet 201a that bends the orbit is utilized, the distance between the reference point CPb and the reference point (equivalent) CPa that specifies the equivalent scanning axis EAs is wide in comparison with an ordinary irradiation system in which scanning is performed by an electromagnet dedicated to scanning (e.g., 1a and 1b in Embodiment 1). Accordingly, in the case of a multi-leaf collimator in which a beam is assumed to spread in a point-light-source manner, there is more conspicuously posed a problem that a penumbra is caused. However, the shape and the orbit of the leaf plate of the multi-leaf collimator 205 according to Embodiment 5 of the present invention are set in such a way that whatever penetration shape is formed, the direction of the plane on which the contour of the penetration shape is formed is the same as the direction of the beam spread. Therefore, the problem of a penumbra, which is conspicuously caused with an irradiation system in which one of the scanning electromagnets is omitted, can readily be solved. As described above, in the particle beam therapy system 210 according to Embodiment 5, it is configured in such a way that scanning for one direction (x or y) out of the x-direction scanning and y-direction scanning is performed by the deflection electromagnet 201a that deflects the direction of a beam axis, and by regarding that the beam axis for setting the reference points CPa and CPb is the beam axis EX of the beam that enters the multi-leaf collimator 205, the configuration and the positioning of the multi-leaf collimator 205 are implemented; therefore, a penumbra can be suppressed and hence a high-contrast irradiation field can be formed. In each of Embodiments 1 through 5, there have been explained the configurations of a multi-leaf collimator and an irradiation system utilizing the multi-leaf collimator and the beam orbit in the irradiation system. In Embodiment 6, there will be explained a treatment planning apparatus in which the operation conditions of a multi-leaf collimator and a particle beam therapy system according to each of the foregoing embodiments of the present invention are set. Here, before explaining a treatment planning apparatus, there will be explained medical practice on which a treatment plan to be implemented by the treatment planning apparatus is based. In general, it is conceivable that medical practice is configured with a number of stages. FIG. 9 represents the stages (flow) of medical practice by a flowchart and describes one or more apparatuses utilized in each stage. With reference to FIG. 9, the flow of a medical practice will be explained. Specifically, medical practice may be roughly configured with a preventive diagnosis stage (MS1), a diagnosis stage (MS2), a treatment planning stage (MS3), a treatment stage (MS4), and a rehabilitation/follow-up stage (MS5). In particular, in a particle beam therapy or the like, the respective apparatuses utilized in the foregoing stages are those described in the right column of FIG. 9. For example, the apparatuses utilized in the diagnosis stage (MS2) are an X-ray image-capturing device, a CT (Computed Tomography), an MRI (Magnetic Resonance Imaging); the apparatus utilized in the treatment planning stage (MS3) is the one that is called a treatment planning apparatus. In addition, the apparatuses utilized in the treatment stage (MS4) are a radiation therapy system and a particle beam therapy system. Next, each of the stages will be explained. The preventive diagnosis stage (MS1) denotes a stage where a diagnosis is implemented preventively, regardless of whether or not there has been shown the onset of a disease. For example, a regular health check and a complete physical examination fall into this stage; with regard to a cancer, a method utilizing fluoroscopic imaging such as radiology, a method utilizing tomography such as PET (Positron Emission Tomography) or PET/CT, and a method utilizing a genetic test (immunological test) are known. The diagnosis stage (MS2) denotes a stage where a diagnosis to be followed by a treatment is implemented after the onset of a disease. In the case of particle beam therapy, in order to implement a treatment, three-dimensional information on the position and the shape of a diseased site is required. Accordingly, there are utilized various kinds of CT and MRI that are capable of obtaining three-dimensional data on a diseased site. The treatment planning stage (MS3) denotes a stage where a treatment plan is generated based on the result of the diagnosis. In the case of particle beam therapy, a treatment plan is generated, in this stage, by a treatment planning apparatus according to Embodiment 6. The treatment planning apparatus will be explained in detail later; here, the residual stage will be explained. The treatment stage (MS4) denotes a stage where an actual treatment is performed based on the result of the treatment plan. In the case of particle beam therapy, a particle beam therapy system is utilized in this stage. A multi-leaf collimator according to each of the foregoing embodiments is utilized for forming an irradiation field in the irradiation system of a particle beam therapy system. In addition, in some cases, the treatment stage is completed with a single irradiation; however, usually, there are implemented a plurality of irradiations, each irradiation of which is performed every certain period. The rehabilitation/follow-up stage (MS5) literally denotes a stage where rehabilitation is performed or there is performed a follow-up to check whether or not a disease has recurred. In the case of a cancer, in a follow-up of this stage, as is the case in the preventive diagnosis stage, a method utilizing fluoroscopic imaging such as radiology, a method utilizing tomography such as PET or PET/CT, or a method utilizing a genetic test (immunological test) is adopted. As described above, in medical practice, the treatment planning is a series of works performed after the diagnosis stage and before the treatment stage. In a particle beam therapy system, a charged particle beam is irradiated based on a treatment plan obtained through a treatment planning apparatus; therefore, a treatment planning apparatus in particle beam therapy is provided with units that approximately play the following roles. Role A: a unit for generating three-dimensional data, based on a plurality of image information items for an irradiation subject, which are preliminarily obtained. Role B: a unit for generating an optimum irradiation condition (treatment planning draft) under given requirements. Role C: a unit for simulating and displaying a final dose distribution for the optimum result (treatment planning draft). In other words, a treatment planning apparatus is provided with a role in which in response to the result of a diagnosis, irradiation condition required for treatment is set; furthermore, the treatment planning apparatus has a unit that plays a role D of generating control data for the particle beam therapy system and the like, based on the set condition. In order to play the foregoing roles, the treatment planning apparatus is specifically provided with the following functions. <Role A> Function a: a function for generating three-dimensional data based on a tomographic image obtained in the diagnosis stage. Function b: a function for displaying the generated three-dimensional data as seen from various viewing points, as is the case with a three-dimensional CAD. Function c: a function for distinguishing a diseased site from normal tissues and storing them in the generated three-dimensional data. <Role B> Function d: a function for setting parameters for a particle beam therapy system utilized in the treatment stage and for simulating irradiation. Function e: a function for optimizing irradiation under the requirements set by a user of the apparatus. <Role C> Function f: a function for displaying the optimized irradiation result in such a way as to be superimposed on the three-dimensional data. <Role D> Function g: a function for setting the shapes, of a multi-leaf collimator and a bolus, for realizing the optimized irradiation (this function is a one when broad-beam irradiation is anticipated, and includes a case of multi-port irradiation). Function h: a function for setting the beam irradiation orbit for realizing the optimized irradiation (when scanning irradiation is anticipated). Function i: a function for generating a driving code, for a particle beam therapy system, for realizing the beam irradiation orbit. <Others> Function j: a function for storing various kinds of data items generated in the apparatus. Function k: a function capable of reading various kinds of data items stored in the past and reusing past information. There will be explained the system configuration of a treatment planning apparatus for realizing the foregoing functions. In recent years, almost no manufacturer of a treatment planning apparatus has designed and manufactured dedicated hardware; the hardware is configured based on a commercially available Unix (registered trademark) workstation or a PC, and as peripheral devices, universal devices are utilized in many cases. That is to say, manufacturers of treatment planning apparatuses primarily develop, manufacture, and sell treatment planning software. In the treatment planning software, for example, there is prepared a module for realizing the functions a through k, as a subprogram to be called by main program. By omitting, as may be necessary, the flow between the function a and the function k or re-implementing it by changing the requirements, the user of a treatment planning apparatus can generate a treatment plan while calling necessary modules. Next, while advancing the explanation to the functions or the modules for realizing those functions, there will be explained a treatment planning apparatus according to Embodiment 6. Function a (module a) generates three-dimensional data based on a series of tomographic images obtained in the diagnosis stage. It is desirable that when a tomographic image is read, patient information such as a patient ID and scanning information (such as a slice interval, a slice thickness, FOV, and a tomographic condition) are also read in a corresponding manner. Here, the three-dimensional data denotes information required for virtually and three-dimensionally reproducing an imaging subject including a diseased site in a treatment planning apparatus. In general, there is utilized a method in which a virtual space is defined in a treatment planning apparatus, points are arranged within the virtual space in such a way as to be spaced evenly apart from one another and in a lattice-like manner, and the respective material information items, which are obtained from a tomographic image, are positioned at the corresponding points. The reason why Function a is required is that one of the biggest objects of a treatment planning apparatus is to simulate treatment, and for that purpose, it is necessary to reproduce a diseased site, which is an irradiation subject, and the peripheral tissues thereof. Function b (module b) displays the generated three-dimensional data as seen from various viewing points, as is the case with a three-dimensional CAD. Function c (module c) distinguishes a diseased site from normal tissues and stores them in the generated three-dimensional data. For example, it is assumed that a tomographic image is obtained through X-ray CT. In this case, the “material information” utilized in Function a corresponds to the radiolucency of an X-ray. That is to say, the three-dimensional model reproduced in the virtual space from this tomographic image represents the shape of a three-dimensional body formed of materials whose radiolucencies are different from one another. In the virtual space of a treatment planning apparatus, the “material information”, i.e., the X-ray radiolucency is rendered by changing the color and the brightness. Furthermore, this “material information” makes it possible to understand that this part of the three-dimensional model reproduced in the virtual space corresponds to a bone or that part corresponds to a tumor, and a diseased site is distinguished from normal tissues. The result of the distinction between a diseased site and normal tissues can be stored in a storage device (such as a hard disk) of the treatment planning apparatus. Function d (module d) sets parameters for a particle beam therapy system utilized in the treatment stage and simulates irradiation. The parameters for a particle beam therapy system denote geometric information on the particle beam therapy system and information on an irradiation field. The geometric information includes the position of the isocenter, the position of the bed, and the like. The information on an irradiation field includes the foregoing “coordinates of the reference point CPa and the coordinates of the reference point CPb” and the like. The parameters include the width (thickness) of the leaf plate 5L of the multi-leaf collimator 5 or 205 (hereinafter, only “5”, representing both, is expressed), the number of the leaf plates 5L, the traveling distance (angle) of the leaf plate 5L, and the like. Function e (module e) optimizes irradiation under the requirements set by a user of the treatment planning apparatus. Function f (module f) displays the optimized irradiation result in such a way as to be superimposed on the three-dimensional data. Function g (module g) sets the shapes, of the multi-leaf collimator 5 and the bolus 6, for realizing the optimized irradiation. This function is a one when broad-beam irradiation is anticipated, and includes a case of multi-port irradiation. Function h (module h) sets the beam irradiation orbit for realizing the optimized irradiation. This function is a one when scanning such as spot scanning or raster scanning is anticipated. Function I (module i) generates a driving code, for a particle beam therapy system, for realizing the beam irradiation orbit. In this situation, when as described later, a coordinate system conforming to a series-of-scanners spread is adopted, there can readily be generated a driving code for realizing an opening shape (penetration shape SP) corresponding to the obtained optimum irradiation plan for the multi-leaf collimator 5 according to each of Embodiments 1 through 5. Function j (module j) stores various kinds of data items set and generated in the apparatus. Function k (module k) can read various kinds of data items stored in the past and reuse past information. <Coordinate System Conforming Series-of-Scanning Spread> In a conventional treatment planning apparatus, the three-dimensional data utilized in Function a and functions following to Function a are represented by an orthogonal coordinate system (xyz coordinate system). In the case of a multi-leaf collimator whose total shape is a conventional rectangular parallelepiped, the leaf driving direction thereof is also represented by an orthogonal-coordinate direction (for example, the x direction and the y direction); therefore, it is convenient to represent the three-dimensional data by an orthogonal coordinate system. That is because leaf driving data and shape data for generating the shape of the opening portion in such a way as to coincide with the shape of a diseased site coincide with each other. On the other hand, in the case of the multi-leaf collimator 5 according to the present invention, it is desirable that because the drive of the leaf plate 5L is performed in a curvilinear manner, the command value for driving the leaf is given as an angle with respect to the reference point. That is to say, it is desired that the shape data for forming the shape of the opening portion in accordance with the shape of a diseased site includes an angle, with respect to the reference point, that is in the same format as the leaf driving command value of the present invention. Thus, the treatment planning apparatus according to Embodiment 6 of the present invention is configured in such a way that the three-dimensional data for a diseased site is represented by a special coordinate system. Specifically, it is a special coordinate system represented by the following definition (D1).[ψa,ψb,rb] (D1)where ψa is a beam deflection angle with respect to the reference axis (Asa) that is perpendicular to the beam axis XB and passes through the reference point CPa, ψb is a beam deflection angle with respect to the reference axis (Asb) that is perpendicular to the beam axis XB and the reference axis Asa and passes through the reference point CPb, and rb is a distance between the reference point CPb (or the reference axis Asb)) and the irradiation point. An arbitrary point in the three-dimensional space can uniquely be represented by the foregoing three information items. In this regard, however, it is required to preliminarily determine the reference points CPa and CPb in accordance with the arrangement of the scanning electromagnets 1a and 1b. Instead of rb, there may be utilized a beam propagation distance ra between the reference point CPa (or the reference axis (Asa)) and the irradiation point. Here, it is assumed that the isocenter, which is an irradiation reference, is utilized as the origin of the xyz coordinate system, and the xyz coordinates of the reference point CPa and the xyz coordinates of the reference point CPb are given as follows. reference point CPa: (0, 0, −1a) reference point CPb: (0, 0, −1b) Then, it is assumed that as illustrated in FIGS. 1 through 3, the upstream scanning electromagnets 1a and the downstream scanning electromagnet 1b are the x-direction scanning electromagnet and the y-direction scanning electromagnet, respectively. In this situation, when the coordinates of a certain point is given by [ψa, ψb, rb] represented by use of the special coordinate system described in the definition (D1), the x coordinate, the y coordinate, and the z coordinate of this point are given by the following equation (4). [ x y z ] = Rot x ⁡ ( φ b ) ⁢ { Rot y ⁡ ( φ a ) ⁡ [ 0 0 l a - l b + r b ] - [ 0 0 l a - l b ] } - [ 0 0 l b ] ( 4 ) Here, when Rotx(ψb), and Roty(ψa) in the equation (4) are defined as in (D2), the xyz coordinates of this certain point is obtained as in the equation (5). Rot x ⁡ ( φ b ) = [ 1 0 0 0 cos ⁢ ⁢ φ b - sin ⁢ ⁢ φ b 0 sin ⁢ ⁢ φ b cos ⁢ ⁢ φ b ] , ⁢ Rot y ⁡ ( φ a ) = [ cos ⁢ ⁢ φ a 0 sin ⁢ ⁢ φ a 0 1 0 - sin ⁢ ⁢ φ a 0 cos ⁢ ⁢ φ a ] ( D ⁢ ⁢ 2 ) [ x y z ] = ⁢ Rot x ⁡ ( φ b ) ⁢ { [ ( l a - l b + r b ) ⁢ sin ⁡ ( φ a ) 0 ( l a - l b + r b ) ⁢ cos ⁢ ⁢ ( φ a ) ] - [ 0 0 l a - l b ] } - [ 0 0 l b ] = ⁢ [ ( l a - l b + r b ) ⁢ sin ⁡ ( φ a ) - sin ⁡ ( φ b ) ⁢ { ( l a - l b + r b ) ⁢ cos ⁡ ( φ a ) - ( l a - l b ) } cos ⁡ ( φ b ) ⁢ { ( l a - l b + r b ) ⁢ cos ⁡ ( φ a ) - ( l a - l b ) } ] - [ 0 0 l b ] = ⁢ [ ( l a - l b + r b ) ⁢ sin ⁡ ( φ a ) - sin ⁡ ( φ b ) ⁢ { ( l a - l b + r b ) ⁢ cos ⁡ ( φ a ) - ( l a - l b ) } cos ⁡ ( φ b ) ⁢ { ( l a - l b + r b ) ⁢ cos ⁡ ( φ a ) - ( l a - l b ) } - l b ] ( 5 ) On the contrary, the method of obtaining the special coordinate system from the xyz coordinate system is described below. Because lb is a given value that is inherent to an irradiation system, ψb can be obtained, as in the equation (6), from the relationship between y and z in the equation (5). - y z + l b = sin ⁢ ⁢ φ b cos ⁢ ⁢ φ b = tan ⁢ ⁢ φ b ⁢ ∴ φ b = arctan ⁡ ( - y z + l b ) ( 6 ) Because being also a given value that is inherent to an irradiation system, la can be defined, as in the definition (D3), from the relationship between y and z in the equation (5); thus, from the relationship with z in the equation (5) and the definition (D3), ψa can be obtained from the equation (7). Λ := ⁢ y 2 + ( z + l b ) 2 + ( l a - l b ) = ⁢ ( l a - l b + r b ) ⁢ cos ⁢ ⁢ ψ a ( D ⁢ ⁢ 3 ) x Λ = sin ⁢ ⁢ φ a cos ⁢ ⁢ φ a = tan ⁢ ⁢ φ a ⁢ ∴ φ a = arctan ⁡ ( x Λ ) ( 7 ) Lastly, rb can be obtained from the equation (8).x2+Λ2=(la−lb+rb)2 ∴rb=√{square root over (x2+Λ2)}−(la−lb) (8) There is provided a coordinate transformation function in which the coordinate system [ψa, ψb, rb] conforming to the foregoing series-of-scanners spread is utilized already from the stage of Function a, i.e., as Function a or as an auxiliary function for implementing Function a, there is performed transformation to a special coordinate system, under the assumption of series-of-scanners. For example, FIG. 10 illustrates, with a block diagram, the characteristic parts in the roles (units) and the functions (modules) of a treatment planning apparatus according to Embodiment 6 of the present invention. In FIG. 10, a treatment planning apparatus 20 is provided with a three-dimensional data generation unit 21 for generating three-dimensional data from image data on a diseased site, which is an irradiation subject; the irradiation condition setting unit 22 for setting an irradiation condition, based on the generated three-dimensional data; and a control data generation unit 23 for generating control data for a particle beam therapy system, based on the set irradiation condition. As described above, these units and modules are formed in a computer by software; thus, these parts are not physically formed. The three-dimensional data generation unit 21 is provided with a three-dimensional data generation module 21M1 for, as Function a, generating three-dimensional data on a diseased site, a body shape, and the like; a coordinate transformation module 21M2 for transforming the generated three-dimensional data into data in the coordinate system [ψa, ψb, rb] represented through the definition (D1) under the assumption of series-of-scanners; a display data generation module 21M3 for, as Function b, generating display data, based on the transformed data; and an irradiation subject separation module 21M4 for distinguishing a diseased site, which is an irradiation subject, from normal tissues, based on the transformed data. As Role A, the three-dimensional data generation unit 21 generates, from image information, three-dimensional data in the coordinate system represented through the definition (D1). As Function B, the irradiation condition setting unit 22 sets, through the functions d and e, an optimum irradiation condition, based on three-dimensional data in the coordinate system represented through the definition (D1). The control data generation unit 23 is provided with a penetration shape setting module 23M1 that sets, as the function g, the penetration shape PS to be formed by the multi-leaf collimator 5 based on at least the set irradiation condition; and a driving code generation module 23M2 that generates, as the function i, a respective driving code for the leaf plates 5L of the multi-leaf collimator 5 based on the set penetration shape. As Role D, the control data generation unit 23 generates at least control data for the multi-leaf collimator 5 in the coordinate system represented through the definition (D1), based on the set irradiation condition. Accordingly, in the three-dimensional data generation unit 21 and the irradiation condition setting unit 22, three-dimensional data, in the coordinate system represented through the definition (D1), for determining the irradiation position is specified by use of a beam deflection angle with respect to the reference axis (Asa) that is perpendicular to at least the beam axis XB and passes through the reference point CPa, and a beam deflection angle with respect to the reference axis (Asb) that is perpendicular to the beam axis XB and the reference axis Asa and passes through the reference point CPb. Thus, the driving code for the multi-leaf collimator 5 generated by the control data generation unit 23 becomes such a driving code as realizes the opening shape (penetration shape PS) conforming to the optimum irradiation plan obtained in the irradiation condition setting unit 22. In other words, in the treatment planning apparatus 20 according to Embodiment 6 of the present invention, a function for conversion into a special coordinate system in which series-of-scanners is anticipated is provided in the functions (modules) for playing the roles of a treatment plan, and three-dimensional data is specified in the special coordinate system. As a result, the shape data for generating the shape of the opening portion in accordance with the shape of a diseased site and the leaf driving command value can be represented by a same format including angles with respect to the reference point (one of the angles is for selecting the leaf plate 5L, among the leaf rows 5c, that has a facing side PL whose angle is near to the angle). Accordingly, in an irradiation system in which a beam spreads in a series-of-scanners manner, a driving code for optimally controlling the multi-leaf collimator 5 can readily be generated. Therefore, in the treatment planning apparatus 20 according to Embodiment 6 of the present invention, for a particle beam therapy system utilizing the foregoing multi-leaf collimator 5 (or 205) capable of suppressing a penumbra for an irradiation system in which a particle beam spreads in a series-of-scanners manner, the leaf driving command value for forming the shape of the opening portion in accordance with the shape of a diseased site can be generated by directly utilizing the three-dimensional data inputted to and outputted from the treatment planning apparatus 20. As described above, the treatment planning apparatus 20 according to Embodiment 6 is configured in such a way as to include the three-dimensional data generation unit 21 for generating three-dimensional data from image data on a diseased site, which is an irradiation subject; the irradiation condition setting unit 22 for setting an irradiation condition, based on the generated three-dimensional data; and the control data generation unit 23 for generating at least the control data, among control data items for a particle beam therapy system, that is for the multi-leaf collimator 5 according to one of Embodiments 1 through 5, based on a set irradiation condition. In addition, the treatment planning apparatus 20 according to Embodiment 6 is configured in such a way that the three-dimensional data generation unit 21 generates the three-dimensional data through a coordinate system that is specified by the beam deflection angle ψa with respect to the reference axis Asa that is perpendicular to the beam axis XB and passes through the reference point CPa, the beam deflection angle ψb with respect to the reference axis Asb that is perpendicular to the beam axis XB and the reference axis Asa and passes through the reference point CPb, and the distance r from the reference axis Asa or Asb, or from the reference point CPa or CPb. As a result, the leaf driving command value for forming the shape of the opening portion in accordance with the shape of a diseased site can be generated by directly utilizing the three-dimensional data that is inputted to or outputted from the treatment planning apparatus 20. In other words, in the control data generation unit 23, the control data can be specified by two deflection angles ψa and ψb; therefore, in a particle beam therapy system that can suppress a penumbra in an irradiation system in which a particle beam spreads in a series-of-scanners manner and that can irradiate a high-contrast and an excellent beam, it is made possible to perform a high-contrast and high-accuracy irradiation. 1: wobbler electromagnet 1a: x-direction (upstream) scanning electromagnet 1b: y-direction (downstream) scanning electromagnet 2: ridge filter 3: range shifter 4: ring collimator 5: multi-leaf collimator 5L: leaf plate 5G: leaf group 5D: leaf driving unit 6: bolus 10: particle beam therapy system 20: treatment planning apparatus 21: three-dimensional data generation unit 22: irradiation condition setting unit 23: control data generation unit Asa: scanning axis (1st axis) of upstream scanning electromagnet (EAs: virtual axis) Asb: scanning axis (2nd axis) of downstream scanning electromagnet CPa: 1st reference point CPb: 2nd reference point EL: facing end face of leaf plate FB: beam bundle (spread) of particle beam OL: driving orbit of leaf plate PI: beam-incident-side end face (adjacent to EL) of leaf plate PL: thickness-direction facing side of leaf plate PS: penetration shape Px: beam-emission-side end face (adjacent to EL) of leaf plate ST: scanning locus of particle beam XB: beam axis of particle beam (EX: beam axis of beam entering multi-leaf collimator) Three-digit numbers each denote variant examples in Embodiments.
	summary
047626623	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a thermally activated trigger device. More particularly it relates to such a device for shutting down a nuclear reactor. Still more particularly it relates to a thermally activated trigger device for rendering a space nuclear reactor subcritical upon reentry to the earth's atmosphere. 2. Description of the Prior Art Temperature-responsive or thermally activated devices are used for a variety of control and safety cutoff functions involving many specialized applications. In several such devices a change in temperature triggers the movement of a piston in a cylinder to accomplish the desired result. Thus U.S. Pat. No. 3,638,733 shows a heat operated fire extinguisher in which a piston closing the mouth of an extinguishing agent pressure vessel is held in a sealing position by a plug of heat fuseable material. When the ambient temperature indicates the presence of a fire, the plug melts releasing the piston and discharging the pressurized vessel. U.S. Pat. No. 4,498,491 shows a thermo-electric valve for scuttling buoyant devices. In this valve a low power resistor holds a spring in compression. Passage of a low control circuit current through the resistor disintegrates it and allows the energy of the spring to expand against the piston, which by its movement permits a pressurized medium to become active. Similarly, in U.S. Pat. No. 4,164,953 a flood valve is shown in which a piston assembly is activated by an electrical heating element surrounding a thin-walled section which is melted to permit pressure of a contained fluid to displace the piston assembly. U.S. Pat. No. 4,422,503 shows a blow-out preventer for use in a control conduit for a downhole safety valve. A fusible link is used to hold a ball check valve in an open position. Melting of the fusible link upon the occurrence of abnormal temperature permits movement of the piston against the ball check valve to close the conduit and prevent fluid flow from the safety valve upwardly through the conduit. U.S. Pat. No. 4,263,839 shows a heat sensitive locking device. By melting of a fusible link, a spring wire is permitted to uncoil thereby permitting a piston to release locking segments for movement to their unlocked positions. U.S. Pat. No. 3,313,312 shows a thermally responsive actuator device in which a piston is held in place by a mass of solidified fusible material. When this fusible mass is melted and become liquid in response to an ambient temperature rise, the piston means assembly then becomes operative. Many control and safety devices have been proposed for use with nuclear reactors, whether of the gas-cooled or liquid-cooled types, including sodium reactors. Basically, in the event of a loss of coolant or other type accident, these control and safety devices act to render the reactor subcritical (incapable of sustaining a nuclear fission reaction) and also to prevent loss of radioactive material to the atmosphere. These devices generally rely on injection of a nuclear poison into the core containing the radioactive fissionable material. Such poisons consist of any material of high absorption cross-section that absorbs neutrons unproductively and hence removes them from the fission chain reaction in a reactor, decreasing its radioactivity. Representative nuclear poisons include boron, cadmium, and hafnium. These are conveniently incorporated into control safety rods, which are then generally inserted from above into the reactor core. U.S. Pat. No. 3,162,578 shows such a control rod having high neutron capture capabilities for use in a nuclear reactor and operating means to move the control rod at high speed when required. U.S. Pat. No. 3,432,387 shows a pressure-actuated control device for moving neutron-absorbing means into the reactor core. U.S. Pat. No. 3,855,060 shows a bottom actuated reactor control rod device. U.S. Pat. No. 3,933,581 shows another control rod drive for reactor shut-down utilizing a latching assembly. U.S. Pat. No. 4,076,584 shows a magnetically activated rodded shut-down system for a nuclear reactor. Spacecraft, whether manned or unmanned, require electric power for several purposes. Particularly for long space voyages requiring large power supplies, chemical forms of energy and solar power have many significant limitations. In situations where large amounts of power are needed over long periods of time, the best source of electricity is a nuclear reactor. Thus, the first SNAP (Systems for Nuclear Auxiliary Power) reactor power plant was launched into space and placed in orbit in 1965. It is necessary that nuclear space reactors be rendered subcritical after reentry into the earth's atmosphere. As indicated, many devices are available for rendering terrestrial nuclear reactors subcritical, such as by inserting poison rods, inserting poison sleeves, or moving reactor segments. Such devices rely on some type of mechanical motion or electrical signal to initiate the desired actuation. Electrical signals frequently tend to be unreliable and could be activated prematurely, shutting a reactor down when not desired, or else fail to operate when required. Accordingly, it is an object of this invention to provide a simple, passive thermally activated trigger device, free from the limitations of known trigger devices, for providing desired control or safety cutoff. More specifically, it is another object to provide such a thermally activated trigger device particularly suitable for rendering a nuclear reactor subcritical. It is still a further specific object of the invention to provide such a device that is effective for rendering a nuclear space reactor subcritical after reentry into the earth's atmosphere. SUMMARY OF THE INVENTION In its broadest aspects, the thermally activated trigger device of this invention comprises a closed vessel with a piston slideably mounted in the vessel to divide it into two compartments. An expandable pressurized fluid, suitably an inert gas such as argon, neon, helium or nitrogen, is present in each of the compartments at substantially the same pressure. A connecting rod or shaft is operatively connected to the piston, the other end of the rod being connected to actuator means. Communicating with one end of one of the compartments is a normally closed vent means for venting fluid from this compartment when the vent means are open. Suitable vent means consist of one or more closed-end tubes. The vent means are designed to open above a selected temperature so that fluid is vented from the compartment connected to the vent means. This results in a difference in pressure between the two compartments so that the higher pressure in the nonvented compartment moves the piston and thereby activates the actuator means. The movement of the piston and its connecting rod provides the desired motion so that the actuator means serve as a trigger or as a supplier of power to activate a safety device. The thermally activated trigger device of the present invention is capable of providing control and safety shutdown functions for use in many chemical process industries where an increase in temperature, whether desired or undesired, is the actuating factor. The present invention is particularly suitable for use as a reentry-activated trigger device for a nuclear space reactor. Accordingly, the present invention will be particularly described with respect to providing such a permanent end-of-life shutdown for a nuclear space reactor. The advantages of the present invention will become more readily apparent from consideration of the following detailed description of the preferred embodiment of the invention in conjunction with the attached drawing.
	claims	1. An x-ray diagnostic device comprising: base; a positioning plate, adapted to receive an examination subject, adjustably mounted at said base; a gallows frame adjustably mounted at said base so as to be selectively positionable relative to said positioning plate, said gallows frame having a first gallows arm, to which said solid state detector is mounted, said first gallows arm being substantially horizontally oriented, and a second gallows arm, oriented substantially vertically relative to said first gallows arm, with said first gallows arm having a longitudinal axis and being mounted to said second gallows arm so as to be adjustable along said longitudinal axis; and a solid state detector for x-rays adjustably mounted at said gallows frame. 2. An x-ray diagnostic device as claimed in claim 1 wherein said solid state detector is mounted at said gallows so as to be adjustable in three orthogonal axes. claim 1 3. An x-ray diagnostic device as claimed in claim 1 wherein said solid state detector is adjustable via said gallows frame from a position above said positioning plate into a position below said positioning plate. claim 1 4. An x-ray diagnostic device as claimed in claim 1 wherein said gallows frame is adjustably mounted at said base for movement along said positioning plate. claim 1 5. An x-ray diagnostic device as claimed in claim 1 wherein said first gallows arm additionally is adjustable along said second gallows arm. claim 1 6. An x-ray diagnostic device as claimed in claim 1 further comprising a common holder disposed at said base, to which both said positioning plate and said gallows frame are mounted. claim 1 7. An x-ray diagnostic device as claimed in claim 6 wherein said holder is adjustably mounted at said base. claim 6 8. An x-ray diagnostic device as claimed in claim 7 wherein said holder is adjustable relative to said base around a rotational axis. claim 7 9. An x-ray diagnostic device as claimed in claim 7 wherein said holder is adjustable in height relative to said base. claim 7 10. An x-ray diagnostic device as claimed in claim 6 wherein said solid state detector is also mounted at said holder, below said positioning plate. claim 6 11. An x-ray diagnostic device as claimed in claim 6 wherein said positioning plate has a longitudinal axis, and wherein said positioning plate is adjustable at said holder along said longitudinal axis. claim 6 12. An x-ray diagnostic device as claimed in claim 6 wherein said positioning plate has a positioning plate surface, and wherein said positioning plate is adjustable at said holder in a direction substantially vertical to said positioning plate surface. claim 6 13. An x-ray diagnostic device as claimed in claim 6 wherein said positioning plate has a longitudinal axis, and wherein said positioning plate is mounted at said holder so as to be adjustable in a direction substantially vertically to said longitudinal axis. claim 6 14. An x-ray diagnostic device as claimed in claim 13 wherein said positioning plate is also adjustable at said holder in a direction substantially along said longitudinal axis. claim 13
	abstract	A system attenuates scatter radiation during a radiological procedure. The radiological procedure using a radiation machine including an emitter, a receiver, a base, and a table for a patient. The base is supported by a floor. The system includes a barrier formed of a radiation attenuation material and positioned over an area on the floor. The barrier is comprised of an elastomeric material and disposed beneath the table. The barrier on the floor can reduce substantial amounts of scatter radiation.
042788927	summary	FIELD OF THE INVENTION The present invention relates to a shielding transport or storage receptacle for radioactive wastes and, more particularly, to a receptacle of the type used to store radioactive substances such as irradiated nuclear reactor fuel elements with a minimum of release of radiation into the environment. BACKGROUND OF THE INVENTION In the above-identified copending application and elsewhere, a transport and storage vessel for radioactive wastes, especially for irradiated nuclear reactor fuel elements, has been made known which comprises a chamber defined by upright walls, i.e. a receptacle shell, a receptacle bottom and a shielding cover of the plug type which fits within the mouth of the receptacle. The shell of the vessel and the bottom are formed unitarily of cast iron, especially spherolitic cast iron, or cast steel, and the shell or walls are provided with a shoulder or flange configuration which can engage an outwardly extending flange on the plug-type shielding cover. The cast alloy or body can be provided with cells in which radiation-absorbing materials can be received, these cells being, for example, so orientated and constructed that they block straight-line paths from the interior of the receptacle outwardly. The radiation-shielding effect, however, is primarily a result of the thickness of the vessel shell and bottom and the thickness or height of the cover which not only must take up the static stresses of transport and storage, but must be sufficient to effect the predominant shielding or adsorption of the radiation from the radioactive wastes whether this radiation is gamma radiation or neutrons. In the earlier transport or storage vessels for the aforedescribed purpose, the shielding cover is held in place by threaded bolts. This enables, prior to the insertion of the cover, the introduction of the radioactive wastes into the interior of the vessel. The sealing between the shielding cover and the vessel walls is effective for long periods, but only as long as any sealing agent remains effective or the sealing structure formed by the flange arrangement remains effective. While such systems have proved to be effective, they nevertheless do not provide a closure which is not dependent upon the sealing means between the shielding cover and the receptacle nor do they permit control of the sealing, i.e. ascertainment of a failure of the shielding-cover seal. OBJECT OF THE INVENTION It is an object of the present invention to provide, in a shielding and transport vessel of the type described in the aforementioned copending application, which is hereby included in its entirety by reference, a hermetic closure of the vessel which is independent of the seal between the shielding cover and the vessel walls and which, in addition, affords sealing control or monitoring as may be required. SUMMARY OF THE INVENTION This object is achieved, in accordance with the present invention, in a receptacle or vessel for the shielding transport or shielding storage of radioactive wastes, especially irradiated fuel elements, which apart from the improvement described below can be of the type fully described in the above-identified application. According to the invention, the receptacle shell is formed along its upper face with an upstanding continuous annular welding lip which defines an annular opening into which an additional or safety cover can be fitted. The safety cover can, in accordance with this invention, be formed with an annular counterlip which lies adjacent the first-mentioned lip and defines an annular welding crevice or junction in which a deposit weld is formed along the upper face of the container. The bead of weldment between these lips can be formed readily by any cast iron or cast steel deposit welding technique because the lips themselves are separated from the mass of the container shell and the mass of the safety cover, respectively, by annular upwardly open grooves. Heat conduction away from the welding site is thus minimized. The safety cover thus overlies the shielding cover and is hermetically sealed to the container wall by the weld seam. In a construction in which the container is provided with a conduit opening from the upper end face into the upper part of the interior of the vessel to allow a fluid to be introduced into the vessel as described in the aforementioned application, the conduit being cast in place or being formed by a space in the cast material, the invention provides that the mouth of this conduit also be closed by the safety cover, i.e. that the safety cover extend over this mouth and that the bead of weldment be deposited outwardly thereof. The system of the present invention has been found to be highly effective in that it affords a seal for the vessel which is not dependent upon the seal between the shielded cover and the body of the vessel. The resulting shielding transport and/or shielding storage receptacle for radioactive waste thus fulfills all of the requirements for such a container and fulfilled by conventional containers with the additional advantage that a greater degree of safety is afforded. While the safety cover is welded onto the vessel wall to provide the hermetic seal, the contents of the vessel remain accessible since the bead of weldment can simply be burned away and the safety cover removed, thereby affording access to the shielding cover. It has been found to be advantageous to provide the safety cover with a bore to which a suction duct can be connected to ascertain whether the seal between the vessel and the shielding cover remains effective. In the event of a failure of the latter seal, the high pressure gas usually provided within the vessel, e.g. helium, can penetrate into the space beneath the safety cover and can be drawn by the suction duct from this space.
059050144	claims	1. A radiation image storage panel having a support, an intermediate layer and a phosphor layer comprising a binder and a stimulable phosphor dispersed therein, said panel being colored with a colorant so that the mean reflectance of said panel in the wavelength region of the stimulating rays for said stimulating phosphor is lower than the mean reflectance of said panel in the wavelength region of the light emitted by said stimulable phosphor upon stimulation thereof, characterized in that said colorant is a triarylmethane dye having at least one aqueous alkaline soluble group and is present in at least one of said support, said phosphor layer or an intermediate layer between said support and said phosphor layer. 2. A radiation image storage panel according to claim 1, wherein the said colorant is a triphenylmethane dye according to the general formula (I), ##STR3## wherein R.sup.1 -R.sup.8 each independently represents hydrogen, alkyl or aryl, provided that at least one of R.sup.1 -R.sup.8 represents an acidic group. 3. A radiation image storage panel according to claim 1, wherein the said colorant has a structure corresponding to formula (II) ##STR4## 4. A radiation image storage panel according to claim 1, wherein the said support has a reflectance percentage of from 45 to 60% in the wavelength range from 350 to 600 nm. 5. A radiation image storage panel according to claim 1, wherein the said intermediate layer has a reflectance percentage of from 85 to 100% in the wavelength range from 350 to 600 nm. 6. A radiation image storage panel according to claim 1, wherein the said support is covered with an aluminum layer which is present between the said support and the said intermediate layer. 7. A radiation image storage panel according to claim 1, wherein the wavelength region of the stimulating rays is between 500 and 700 nm . 8. A radiation image storage panel according to claim 1, wherein the wavelength region of light emitted by said stimulable phosphor upon stimulation thereof is between 350 and 450 nm. 9. A radiation image storage panel according to claim 1, wherein the said binder is a rubbery binder substantially consisting of one or more block copolymers, having a saturated elastomeric midblock and a thermoplastic styrene endblock, as rubbery and/or elastomeric polymers. 10. A radiation image storage panel according to claim 1, wherein said stimulable phosphor has a composition according to the formula EQU Ba.sub.1-x-y-p-3q-z Sr.sub.x M.sub.y.sup.2+ M.sub.2p.sup.1+ M.sub.2q.sup.3+ F.sub.2-a-b Br.sub.a I.sub.b :zEu M.sup.1+ is at least one alkali metal selected from the group consisting of Li, Na, K, Rb and Cs; M.sup.2+ is at least one divalent metal selected from the group consisting of Ca, Mg and Pb; M.sup.3+ is at least one trivalent metal selected from the group consisting of Al, Ga, In, Tl, Sb, Bi, Y and a trivalent lanthanide; 0.ltoreq.x.ltoreq.0.30, 0.ltoreq.y.ltoreq.0.10, 10.sup.-6 .ltoreq.z.ltoreq.0.2, 0.ltoreq.p.ltoreq.0.3, 0.ltoreq.q.ltoreq.0.1, 0.05.ltoreq.a.ltoreq.0.76, 0.20.ltoreq.b.ltoreq.0.90 and a+b<1.00. M.sup.1+ is at least one alkali metal selected from the group consisting of Li, Na, K, Rb and Cs; M.sup.2+ is at least one divalent metal selected from the group consisting of Ca, Mg and Pb; M.sup.3+ is at least one trivalent metal selected from the group consisting of Al, Ga, In, Ti, Sb, Bi, Y and a trivalent lanthanide; 0.06.ltoreq.x.ltoreq.0.20, 0.ltoreq.y.ltoreq.0.10, 10.sup.-6 .ltoreq.z.ltoreq.0.2, .ltoreq. p.ltoreq.0.3, 0.ltoreq.q.ltoreq.0.1, 0.05.ltoreq.a.ltoreq.0.76, 0.20.ltoreq.b.ltoreq.0.90 and 0.85.ltoreq.a+b.ltoreq.0.96. M.sup.1+ is Rb or Cs; M.sup.2+ is Pb; M.sup.3+ is at least one trivalent metal selected from the group consisting of Al, Ga, In, Tl, Sb, Bi, Y and a trivalent lanthanide; 0.06.ltoreq.x.ltoreq.0.20, 10.sup.-4 .ltoreq.y.ltoreq.10.sup.-3, 10.sup.-4 .ltoreq.p.ltoreq.10.sup.-1, q=0, 10.sup.-6 .ltoreq.z.ltoreq.0.2, 0.05.ltoreq.a.ltoreq.0.76, 0.20.ltoreq.b.ltoreq.0.90, and 0.85.ltoreq.a+b.ltoreq.0.96. 11. A radiation image storage panel according to claim 10, wherein said trivalent lanthanide is selected from the group consisting of La, Ce, Pr, Nd, Sm, Gd, Tb, Dy, Ho, Er, Tm, Yb, and Lu. 12. A radiation image storage panel according to claim 1, wherein said stimulable phosphor has a composition according to the formula EQU Ba.sub.1-x-y-p-3q-z Sr.sub.x M.sub.y.sup.2+ M.sub.2p.sup.1+ M.sub.2q.sup.3+ F.sub.2-a-b Br.sub.a I.sub.b :zEu 13. A radiation image storage panel according to claim 1, wherein said stimulable phosphor has a composition according to the formula EQU Ba.sub.1-x-y-p-3q-z Sr.sub.x M.sub.y.sup.2+ M.sub.2p.sup.1+ M.sub.2q.sup.3+ F.sub.2-a-b Br.sub.a I.sub.b :zEu
043691617	description	DETAILED DESCRIPTION FIG. 1 shows the cylindrical housing 1 containing the apparatus for moving the control rods disposed with its vertical axis at the upper part of the vessel of the nuclear reactor, this very thick housing being pressure-resistant and communicating at its lower part with the interior of the vessel. The housing 1 is closed sealingly at its upper part not represented in FIG. 1. The control shaft 2 disposed along the axis of the housing 1, i.e., disposed with its axis vertical, is connected at its lower part (not shown in FIG. 1) to the upper part of the control rod which is solid with the control shaft in its movements in the vertical direction. A rack 3 is machined on one part of the control shaft in the longitudinal direction of this control rod. A pinion 5 whose axis is perpendicular to the axis of the control shaft is in engagement with the rack. The rack occupies a length of the control shaft which is such that the pinion 5 is able to move the control shaft upwards with the aid of the rack with which it is in engagement to the position of the control shaft corresponding to the position of maximum extraction of the control rod. The rack is also machined towards the top of the control shaft over a length sufficient for the pinion to remain engaged with this rack when the control rod is in its position of maximum insertion, for example at the time of an emergency shutdown. The pinion 5 remains constantly engaged with the rack 3, whatever the vertical position of the control shaft 2. As FIGS. 1 and 2 show, the pinion 5 is freely mounted on a shaft 7 itself mounted to rotate inside the support 8 of the apparatus for moving the control rod by two bearings 9 and 10. The pinion 5 is freely mounted on the shaft 7 by bearings 11 and 12 whose inner rings are solid with the shaft 7. In this way, the pinion 5 remains in a fixed position with respect to the rack 3, the shaft 7 being fixed in axial position by a stop 14 and by the back part of a conical pinion 15, both solid with the shaft 7. The conical pinion 15 itself engages with a conical pinion 16 solid with a shaft 18 mounted to rotate in the support 8 and itself solid at its end opposite the conical pinion 16 with a pinion 19. The pinion 19 is itself solid in rotation with a ring 20 connected to the output of a reducer 21 whose input is connected to the driven part of a magnetic coupling (not represented in FIG. 1). The motor apparatus for driving the shaft 7, through the magnetic coupling of the reducer and the set of pinions described, is constituted by a three-phase motor disposed outside the magnetic casing 1 which causes a locking in position when it is not supplied. The three-phase driving motor transmits its rotary movement to the driven part of the magnetic coupling through the casing 1 which therefore remains totally sealed and isolated from the external environment. When the three-phase motor is set in rotation, this rotary movement is then transmitted to the shaft 7 which continues to move as long as the three-phase motor is supplied and which locks in position when the three-phase motor is no longer supplied. As FIG. 2 shows, the shaft 7 has grooves 24 over one part of its length while a clutch device made in the form of a pinion 25 has corresponding grooves over its inner bore. The pinion 25 is engaged on the shaft 7 so as to be solid in rotation with the shaft 7 but movable in translation in the axial direction of this shaft. The pinion 25 is engaged on the shaft 7 so that its front face 26 comes into contact, at the forward translatory movement of the pinion 25, with the back face 27 of the pinion 5 whose axial position on the shaft 7 is fixed. The front face 26 of the pinion 25 and the back face 27 of the pinion 5 have corresponding teeth of the "jaw" type which come into engagement at the end of movement. In this way, the shaft 7, the clutch ring 25 and the driving pinion 5 are solid in rotation when the clutch pinion 25 is in its front position as represented in FIG. 2. A push-rod 30 is also mounted articulatedly via a shaft 31 on the support 8 of the apparatus for moving the control rods, one of the ends of this push-rod 30 being connected articulatedly to the lower part of an actuating sleeve 32 coaxial with the control shaft 2 and extending to the upper part of the control apparatus. The sleeve 32 is kept in its high position as represented in FIG. 1 by a magnetic coil, disposed at the upper part of the apparatus and not represented, acting on a magnetic part solid with the upper part of the sleeve 32. At its base, the sleeve 32 is connected to an actuating ring 34 on which is articulated, at 35, one of the arms of the bent push-rod 30 whose other end is constituted by an actuating fork 36 in engagement with the clutch pinion 25 via two wheels 37 and 38. In its high position represented in FIG. 1, the actuating sleeve holds, via the bent push-rod 30, the clutch pinion 25 in an advanced position as represented in FIGS. 1 and 2. In this position the teeth of the front face of the clutch 25 are engaged in the teeth of the back face of the pinion 5 for driving the rack. As long as the coil for holding the sleeve 32 in the high position is supplied, the driving pinion 5 is consequently in mechanical connection with the pinion 25 and the shaft 7 which is itself in mechanical connection with the rotary driving means described. It is therefore possible to move the control rod in this position, by causing a controlled rotation of the driving pinion 5 by means of the three-phase motor and the whole kinematic chain previously described. Similarly, it is possible to hold the control rod in position, since the three-phase motor is locked against rotation when it is not supplied. It is therefore possible to obtain all movement of the control rod and any holding of position by control of the supply of the three-phase motor. If the supply to the coil for holding the sleeve 32 in the high position is interrupted, this sleeve falls under the action of its own weight and drives the push-rod 30 in rotation so that the fork 36 drives the clutch pinion in translation backwards, the clutch teeth 25 and 5 being then disengaged. The pinion 5 freely mounted for rotation on the shaft 7 then no longer holds the rack 3 and the control rod 2 in position and the control rod is able to fall under the action of its own weight until this control rod, connected to the control shaft 2, arrives at its position of maximum insertion in the reactor core. Emergency shutdown of the reactor can therefore be obtained by cutting off the supply to the coil for holding the sleeve 32 in position, this being particularly simple and quick. Throughout its movements, the control shaft 2 is guided by wheels such as 40 and 41 freely mounted on shafts solid with the support 8 of the control apparatus. Clearly, the principal advantages of the invention are that it allows the use of a rack and pinion apparatus to move a control rod in a reactor during control of this reactor, and holding of the control rod in position, while allowing tripping of the emergency shutdown in an extremely reliable and very simple way, by releasing an actuating sleeve which falls under its own weight. In this way, all the advantages of the pinion-rack apparatus which allows movements of the control rod in a continuous and shockless motion with precise stopping at any level of the stroke of the control rod are obtained, this mechanism also taking up less room inside the housing over the reactor vessel. In addition, the mounting of the rack and pinion control apparatus inside the housing over the reactor vessel is particularly simple and does not require a precise positioning in rotation of the apparatus with respect to the control rod. In addition, after an emergency shutdown, the driving pinion 5 remains engaged with the rack 3, so that all that is required to return the apparatus for moving the control rod to service is to cause the sleeve 32 to return to the high position by supplying the coil for holding this sleeve and to supply the three-phase motor for raising the rod and any other subsequent movement for controlling the reactor. In addition, the emergency shutdown can be easily and reliably achieved and, to further increase the reliability of the apparatus, a return spring can be associated with the sleeve 32 to reinforce the action of the weight so as to bring the sleeve 32 into the low position. Furthermore, in the case of the use of a motor for driving the pinion apparatus in rotation, the motor being locked against rotation by cutting off of its supply, holding in position of the control rod is assured without consumption of energy other than the energy supplied to the coil for holding the actuating sleeve in the high position. In addition, when a magnetic coupling is used to traverse the wall of the housing of the mechanism, for example a permanent magnet apparatus which allows the motor torque to be transmitted through the sealed casing 1 synchronously and with a high output, an automatic limitation of the stresses exerted on the control rod is thus assured since, in the case of excessive stress, the driving and driven parts of the magnetic coupling become uncoupled. The invention is not limited to the embodiment just described; it includes all the variants thereof and points of detail can be changed without exceeding the scope of the invention. Thus, instead of a three-phase motor effecting a locking in position by lack of current associated with a magnetic coupling for traversing the sealed housing, any type of braking motor can be used, associated with a rotary mechanical traversing allowing the movement to be transmitted into the sealed housing. Use of a stepping motor with a shaded rotor (or with a polegap sleeve) is also possible to assure motorization of the apparatus for driving the control rod. The invention applies to all nuclear reactors in which power control and emergency shutdown is effected by movement in the vertical direction of a unit absorbing neutrons which can be inserted to a greater or lesser depth in the reactor core.
054897378	abstract	A processing system for radioactive waste is composed of an adjusting tank having a sampling port, a solidification processing system, and a package inspection apparatus, and a package, of which inventory per a package has been exactly grasped, is prepared by solidification of the waste with the processing system after determining radioactivity of the waste by measurement before the solidifying process.. In accordance with the present invention, data on radioactivity before and after preparation of package of waste become clear, and management of each package at transportation and intermediate storage of the packages is facilitated.
062884018	description	DETAILED DESCRIPTION FIG. 2 shows a schematic cross sectional view of a field emission source 100, with an electron emitter 102 and an extraction electrode 120 and including centering electrodes 130 to electrostatically align an electron beam with the optical axis 101 in accordance with an embodiment of the present invention. The electron emitter 102 is a Schottky emitter with an etched single crystal tungsten tip 104, approximately 50-100 .mu.m in diameter, that is spot-welded on a filament 106 such as a tungsten wire, approximately 50-100 .mu.m in diameter. The filament 106 is mounted on a support structure, which includes a base 108, two rods 110, and a suppressor cap 112. The filament 106 is connected to the rods 110, which is supported by the base 106. Electron emitter 102 may also be a cold-field emitter as is well known in the art. The electron emitter 102 is mounted in front (upstream) of the extraction electrode 120. The extraction electrode 120 defines a center aperture 122, which is approximately 1-2 .mu.m diameter. Following extraction electrode 120 are the conventional lens structures of the microcolumn, which for the sake of simplicity are shown as a single lens electrode 140 defined by a lens aperture 142. The optical axis 101 is centered on the extraction electrode aperture 122 and the lens aperture 142. The field emission source 100 electrostatically corrects any misalignment between the electron emitter 102 and the optical axis 101. Thus, the electron emitter 102 may be rigidly mounted with respect to optical axis 101 and only a coarse physical prealignment of the electron emitter 102 with the extraction electrode 120 is necessary. The prealignment is mechanically performed, for example, using a conventional flexure stage or inertial walker during assembly. Advantageously, the electrostatic alignment in accordance with the present invention aligns the electron beam with the optical axis with the same or greater precision as with the conventional mechanical alignment. Thus, the necessity of extremely precise mechanical alignment is obviated. In accordance with one embodiment of the present invention, the electrostatic alignment is achieved by electrostatic centering electrodes 130 positioned between the electron emitter 102 and the extraction electrode 120. FIG. 3 shows a top (plan) view of the extraction electrode 120 and electrostatic centering electrodes 130. As shown in FIG. 3, the centering electrodes 130 are in a quadrupole arrangement with electrode elements 130a, 130b, 130c, and 130d and approximately centered on optical axis 101. It should be understood that centering electrodes 130 may be a higher number multipole arrangement, e.g., an octopole or dodecapole. The centering electrodes 130 are fabricated using the same micromachining technology used to fabricate lens components in a microcolumn, as is well understood by those of ordinary skill in the art. An electrically insulating layer 132 is deposited over the extraction electrode 120. The insulating layer 132 is for example silicon dioxide, pyrex, or a similar material and is 0.5 to 20 .mu.m thick. A conductive layer, such as aluminum, gold, silicon (that is heavily n doped), copper, platinum, or other conductive material, is then deposited over the insulating layer 132 to a thickness of 1-100 .mu.m. The conductive layer is then lithographically patterned and etched to form the desired centering electrodes 130. The deposition, patterning and etching of a conductive layer is well understood by those of ordinary skill in the art. To cause the emission of electrons, a voltage Vc is applied to the rods 110 of the electron emitter 102, while a voltage Vs is applied to the suppressor cap 112, and a voltage Ve is applied to the extraction electrode 120. The difference in potentials between the electron emitter 102 and the extraction electrode 120 (Vc-Ve) creates a strong electric field in the area of the tip 104, causing the emission of electrons. The temperature of the tip 104 is regulated to approximately 1700 to 1800 degrees K by a current passing through the filament 106, and the average power is 1.5-2.0 W. Potentials are applied to the individual centering electrode elements 130a, 130b, 130c, and 130d to form a deflection field near the optical axis 101. The deflection field approximately centers the emitted electron beam with respect to the optical axis, i.e., the axis of the electron beam passes through the center of the next lens down stream. Potentials of equal amplitude and opposite polarity are applied to opposite electrodes. Thus, for example, electrode element 130a will be at a voltage Vdx while electrode element 130c will be at a voltage -Vdx. Similarly, electrode element 130b will be at a voltage Vdy while electrode element 130d will be at a voltage -Vdy. The typical voltages used on the electrode elements range from a few tens of volts to a few hundred volts. If the electron emitter 102 is properly aligned with optical axis 101 and thus no centering potential is necessary, a uniform bias potential Vb may be applied to all individual electrode elements so that a uniform extraction field is preserved. FIG. 4 shows a schematic cross sectional view of a misaligned field emission source 100 producing an electron beam 103 while centering electrodes 130 electrostatically align the electron beam 103 with the optical axis 101. As shown in FIG. 4, without the centering potential produced by centering electrodes 130, an electron beam would be misaligned with the optical axis (as indicated by the broken lines 103a). By application of centering potential on centering electrodes 130, an electrostatic deflection field is generated (as indicated by arrow 131), which deflects the electron beam 103 so that it is in approximate alignment with the optical axis 101, i.e., the axis of the electron beam passes through the center of the next lens down stream (not shown in FIG. 4). The centering process may result in a small tilt of the electron beam 103 with respect to the optical axis 101, as shown in FIG. 4. The centering systems in the lenses that follow the extraction electrode 120, e.g., lens 140 shown in FIG. 2, may compensate for any residual tilt. FIG. 5 shows a schematic cross sectional view of a field emission source 200 in accordance with another embodiment of the present invention. Field emission source 200 is similar to field emission source 100, shown in FIG. 2, like designated elements being the same, however, field emission source 200 includes a second set of electrostatic centering electrodes 210 follow centering electrodes 130. The second set of centering electrodes 210 are similar in fabrication and operation to centering electrodes 130. The second set of centering electrodes 210 are used to allow simultaneous beam translation and parallelism to the optical axis thereby removing the residual tilt generated by centering electrodes 130 (which is illustrated in FIG. 4). Centering electrodes 210 are fabricated in a manner similar to centering electrodes 130. An insulating layer 212 of approximately 0.5 to 20 .mu.m is deposited over the extraction electrode 120. A conductive layer that forms the second set of centering electrodes 210 is deposited over the insulating layer 212. Another insulating layer 130, similar to insulating layer 212 is then deposited followed by another conductive layer that forms the first set of centering electrodes 130. The stack of conductive layers and insulating layers is then lithographically patterned and etched to define the desired centering electrodes 130 and second set of centering electrodes 210. Of course, if desired additional sets of centering electrodes may be produced in a similar manner. FIG. 6 shows a schematic cross sectional view of a misaligned field emission source 200 producing an electron beam 203 while centering electrodes 130 and a second set of centering electrodes 210 electrostatically align the electron beam 203 with the optical axis 101. As shown in FIG. 6, by application of centering potential on centering electrodes 130, a first electrostatic deflection field is generated (as indicated by arrow 231), which deflects the electron beam 203 so that it is in approximate alignment with the optical axis 101, i.e., the axis of the electron beam 203 passes through the center of the centering electrodes 210. The applied centering potentials are of equal amplitude and opposite polarity for opposite electrodes, i.e., .+-.Vdx1 and .+-.Vdy1 (which is applied to the centering electrode elements not shown in the cross sectional view of FIGS. 5 and 6). By application of a second centering potential on the second set of centering electrodes 210, a second electrostatic deflection field is generated (as indicated by arrow 232), which deflects the electron beam 203 in a direction opposite to the direction that the electron beam 203 was deflected by centering electrodes 130. The second set of centering potentials are applied to opposite electrodes of the second set of centering electrodes 210, i.e., .+-.Vdx2 and .+-.Vdy2 (which is applied to the centering electrode elements not shown in the cross sectional view of FIGS. 5 and 6). As shown in FIG. 6, the orientations of the deflection fields generated by the two sets of deflection electrodes 130 and 210 are opposite in direction. The second set of potentials applied to centering electrodes 210 removes residual tilt created by centering electrodes 130, thereby deflecting the electron beam 203 to be approximately parallel with the optical axis 101, e.g., within 3 milliradians. A bias potential Vb may be applied to one or both sets of centering electrodes 130 and 210 so that a uniform extraction field is preserved if no electrostatic alignment is necessary. FIG. 7 shows a schematic cross sectional view of a field emission source 300 in accordance with another embodiment of the present invention. Field emission source 300 is similar to field emission source 100, shown in FIG. 2, like designated elements being the same, however, the extraction electrode 120 and the centering electrodes 130 are replaced with a centering extraction electrode 310. FIG. 8 shows a top view of the centering extraction electrode 310. As shown in FIG. 8, the centering extraction electrodes 310 is an extraction electrode split into a quadrupole arrangement having electrode elements 310a, 310b, 310c, and 310d. Of course, centering extraction electrode 310 may have a higher multipole arrangement if desired. The centering extraction electrodes 310 operate as both the extraction electrode and the centering electrode. As shown in FIGS. 6 and 7, the extraction potential Ve and the centering potentials .+-.Vdx and .+-.Vdy are superimposed on the individual elements of the centering extraction electrodes 310. The centering extraction electrodes 310 are fabricated using the same micromachining silicon technology used to fabricate lens components in a microcolumn, as is well understood by those skilled in the art. If desired, centering extraction electrodes 310 may be fabricated on a substrate (not shown), such as a silicon substrate, which may aid in the prevention of warping or mechanical breakdown of the centering extraction electrodes 310. While the present invention has been described in connection with specific embodiments, variations of these embodiments will be obvious to those of ordinary skill in the art in light of the present disclosure. Thus, for example, while the present disclosure describes a field emission source in accordance with the present invention as including an electron emitter, it should be understood that any charged particle, including positive ions may be emitted and electrostatically aligned in accordance with the present invention. Therefore, the spirit and scope of the appended claims should not be limited to the foregoing description.
	abstract	A nail lamp for curing UV-curable nail gel uses light emitting diodes (LEDs) that emit ultraviolet light and are relatively lower power. The nail lamp is powered from an exterior power source, such as a wall socket, or by a rechargeable battery pack. A battery compartment of the nail lamp holds the battery pack, which is removable without disassembling the nail lamp. The nail lamp is easily transportable to different locations and can be used even when a wall socket is unavailable. A curing time of the nail lamp is user-selectable. The nail lamp can also include detection sensors to detect a person's hand or foot in a treatment chamber and automatically turn on or off the LEDs.
042749220	abstract	An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.
	summary
053496248	abstract	Samples of particulate soil contaminated with solid particles having higher -ray absorption coefficients than the soil, are irradiated with a microfocused X-ray beam within an analysis zone to produce X-ray images of such samples. By controlling relative scanning and magnifying movements between the samples within the analysis zone and the X-ray beam, image features corresponding to the solid particles in the soil are rendered detectable for measurement purposes and to provide data from which size and distribution of contaminants in the soil are calculated.
	description	The present invention relates to an installation method of a water-chamber working apparatus that can facilitate an installing process of a manipulator in a configuration in which the manipulator is suspended from a ceiling of a water chamber and arranged therein. A water-chamber working apparatus is carried into and installed in a water chamber in a steam generator to perform a work in the water chamber by remote control. In recent years, a water-chamber working apparatus provided with a manipulator having a tool for a work in a water chamber attached to an end thereof has been proposed. As a conventional water-chamber working apparatus employing such a configuration, a technique described in Patent Literature 1 is known. Patent Literature 1: Japanese Patent Application Laid-open No. 2007-183278 An object of the present invention is to provide an installation method of a water-chamber working apparatus that can facilitate an installing process of a manipulator in a configuration in which the manipulator is suspended from a ceiling of a water chamber and arranged therein. According to an aspect of the present invention, in an installation method of a water-chamber working apparatus that is suspended from a tube plate surface of a water chamber in a steam generator and driven by remote control to perform a work in the water chamber, the water-chamber working apparatus includes a base that holds heat transfer tubes on the tube plate surface and is fixed to the tube plate surface, and a manipulator that is coupled with the base, suspended in the water chamber and arranged therein, and has a separable configuration. The installation method includes a base installing step of installing the base on the tube plate surface, and a manipulator coupling step of carrying the separated manipulator into the water chamber sequentially and individually and coupling the manipulator with the base. According to the installation method of a water-chamber working apparatus, an installing process of the manipulator is facilitated in a configuration in which the manipulator is suspended from the ceiling of the water chamber and arranged therein. For example, in a configuration in which the manipulator is carried into the water chamber in an integrated state and then installed, the weight of the manipulator becomes heavy, and thus a work for hoisting the manipulator to the ceiling of the water chamber is not easy. Advantageously, in the installation method of a water-chamber working, at the manipulator coupling step, an attaching jig is spanned from the base on the tube plate surface to a maintenance hatch of the water chamber and then installed, and the separated manipulator is coupled with the base, while being guided by the attaching jig. According to the installation method of a water-chamber working apparatus, the attaching jig is used as a guide, and the separated manipulator is carried from the maintenance hatch of the water chamber to a position of the base. With this configuration, a heavy rear stage can be easily carried from the maintenance hatch to the base on the tube plate surface. Advantageously, in the installation method of a water-chamber working, at the base installing step, the base and a base carrying and attaching jig installed on the tube plate surface are connected with each other via a wire, and the base is pulled up to the base carrying and attaching jig by the wire. According to the installation method of a water-chamber working apparatus, the heavy base can be easily pulled up to the tube plate surface of the water chamber and installed therein. Advantageously, in the installation method of a water-chamber working, replacement of a tool attached to an apical end of the manipulator is performed in a state where the apical end of the manipulator is projected from the maintenance hatch of the water chamber to outside thereof while in a state of being suspended in the water chamber. According to the installation method of a water-chamber working apparatus, a replacement work of the tool can be performed outside of the water chamber, in a state where the water-chamber working apparatus is installed in the water chamber. With this configuration, the replacement work of the tool can be facilitated. Advantageously, in the installation method of a water-chamber working, at a time of attaching or detaching the tool, a joint portion of the manipulator is made flexible with respect to an external force. According to the installation method of a water-chamber working apparatus, the moment acting on the base can be reduced, and thus it is possible to prevent detaching of clamping of the base at the time of performing an attachment work and a replacement work of the tool. According to the installation method of a water-chamber working apparatus, an installing process of a manipulator is facilitated in a configuration in which the manipulator is suspended from a ceiling of a water chamber and arranged therein. The present invention is explained below in detail with reference to the accompanying drawings. The present invention is not limited to the following embodiment. Furthermore, constituent elements in the embodiment include elements that can be easily replaced or obviously replaceable while maintaining the unity of invention. In addition, a plurality of modifications described in the following embodiment can be arbitrarily combined within a scope obvious to persons skilled in the art. [Steam Generator in Nuclear Plant] A nuclear plant 100 includes, for example, a pressurized light-water reactor nuclear power plant (see FIG. 34). In the nuclear plant 100, a reactor vessel 110, a pressurizer 120, a steam generator 130, and a pump 140 are sequentially coupled by a primary coolant pipe 150 so as to form a circulation pathway of a primary coolant (a primary circulation pathway). A circulation pathway of a secondary coolant (a secondary circulation pathway) is also formed between the steam generator 130 and a turbine (not shown). In the nuclear plant 100, the primary coolant is heated in the reactor vessel 110 to become a high-temperature and high-pressure primary coolant, which is supplied to the steam generator 130 via the primary coolant pipe 150 while being pressurized by the pressurizer 120 to maintain the pressure constant. In the steam generator 130, the primary coolant flows into an inlet-side water chamber 131, and is supplied from the inlet-side water chamber 131 to a plurality of U-shaped heat transfer tubes 132. Heat exchange is performed between the primary coolant and the secondary coolant in the heat transfer tubes 132, and the secondary coolant is evaporated to generate steam. The turbine is driven by supplying the secondary coolant as the steam to the turbine so as to generate power. The primary coolant having passed through the heat transfer tubes 132 is recovered at a side of the pump 140 from an outlet-side water chamber 133 via the primary coolant pipe 150. In the steam generator 130, an inlet nozzle 135 is provided in the inlet-side water chamber 131, and the inlet-side primary coolant pipe 150 is welded and connected to the inlet nozzle 135 (see FIG. 35). An outlet nozzle 136 is provided in the outlet-side water chamber 133, and the outlet-side primary coolant pipe 150 is welded and connected to the outlet nozzle 136. The inlet-side water chamber 131 and the outlet-side water chamber 133 are divided by a partition plate 134. A tube plate 137 is installed in the steam generator 130. The tube plate 137 supports lower ends of the heat transfer tubes 132, and divides an upper part of the steam generator 130 and the water chambers 131 and 133 to constitute a ceiling of the water chambers 131 and 133. A maintenance hatch 138 from which workers enter into and exit from the water chambers 131 and 133 is provided in the inlet-side water chamber 131 and the outlet-side water chamber 133 (see FIG. 36). [Water-Chamber Working Apparatus] FIG. 1 is a perspective view of an installed state of a water-chamber working apparatus according to an embodiment of the present invention. FIG. 2 is a perspective view of the water-chamber working apparatus shown in FIG. 1. FIG. 3 is an explanatory diagram of an assembly of a base and a coupling link of the water-chamber working apparatus shown in FIG. 1. A water-chamber working apparatus 1 is carried into and installed in the water chambers 131 and 133 in the steam generator 130, and is remote-controlled to perform a work in the water chamber (see FIG. 1). The water-chamber working apparatus 1 includes a base 2, a coupling link 3, a manipulator 4, and a tool 5 (see FIG. 2). The base 2 is a part for suspending the water-chamber working apparatus 1 from a tube plate surface 137a of the water chambers 131 and 133, and includes a base body 21, a pair of wings 22a and 22b, and a plurality of clampers 23a and 23b. The base body 21 is a framed casing. The wings 22a and 22b are inserted into and installed in the base body 21. These wings 22a and 22b are driven by an extendable ladder mechanism, and can be slidably displaced with respect to an installed position of the base body 21 (see FIG. 3). The wings 22a and 22b can be slidably displaced in mutually different directions. Furthermore, the wings 22a and 22b are driven mutually independently. The clampers 23a and 23b have a clamping mechanism for clamping the heat transfer tubes 132. For example, a configuration in which a claw-like apical end is brought into frictional contact with an inner periphery of the heat transfer tube by inserting the claw-like apical end into the heat transfer tube and widening a toe part, thereby clamping the heat transfer tube can be adopted as the clamping mechanism. For example, in the present embodiment, the base body 21 is formed of a frame-like member having a substantially cubic shape, and the pair of wings 22a and 22b having an extension mechanism are respectively inserted into and installed in the base body 21 (see FIGS. 2 and 3). Furthermore, ends of the respective wings 22a and 22b can be slidably displaced in a width direction of the base body (in a planar direction of the tube plate surface 137a in the installed state) by driving the extension mechanism. Pairs of the wings 22a and 22b are arranged so that the wings can be slidably displaced in directions orthogonal to each other. A plurality of the clampers 23a (23b) in a set are brought into line and arranged, matched with an installation interval of the heat transfer tubes 132. The set of the clampers 23a (23b) are respectively arranged at the front and back of the wing 22a (22b). With this configuration, in a state where the base 2 is installed on the tube plate surface 137a, the clampers 23a (23b) are respectively arranged at four quarters of the base 2, and the wing 22a (22b) extends or retracts to slidably displace the end thereof, thereby enabling to slidably displace these clampers 23a (23b) in the planar direction of the tube plate surface 137a. The coupling link 3 is a unit that couples the base 2 with the manipulator 4. The coupling link 3 is rotatably coupled with the base body 21 of the base 2, designating a height direction of the base 2 as a rotation axis I. Furthermore, the coupling link 3 includes a mounting surface 31, which is inclined with respect to the rotation axis I. The manipulator 4 is coupled with the mounting surface 31. The manipulator 4 is a multi-axis manipulator. The manipulator 4 is coupled with the coupling link 3, with a reference axis m of a basic orientation thereof (an upright state) being inclined at a predetermined angle θ with respect to the rotation axis I of the coupling link 3. Furthermore, the manipulator 4 has a separable configuration that can be separated into a front stage 41 and a rear stage 42. For example, in the present embodiment, a seven-axis manipulator is adopted, and the manipulator is formed to be separable into four axes on the end side of the manipulator 4 as the front stage 41 and three axes on the rear stage side as the rear stage 42 (see FIGS. 14 and 15). The front stage 41 and the rear stage 42 have a connection structure, which can be detached by a single touch by a telescopic clamping mechanism. Specifically, a rod of the front stage 41 is inserted into a coupling hole in the rear stage 42, and the front stage 41 opens a claw by an air cylinder (not shown) and is clamped to the rear stage 42. With this configuration, the front stage 41 and the rear stage 42 can be coupled easily by remote control. The tool 5 is a tool that corresponds to a predetermined work in a water chamber and is attached to an end of the manipulator 4. For example, the tool 5 is a maintenance work tool used for a maintenance work in the water chamber, and is constituted by a testing tool, a cutting tool, a welding tool and the like. Specifically, the tool 5 for testing or repairing the inlet nozzle 135, the outlet nozzle 136, the heat transfer tubes 132, a welded part between the partition plate 134 and the tube plate 137, and a welded part between the partition plate 134 and a water chamber mirror is prepared. In the present embodiment, a plurality of types of the tools 5 are prepared corresponding to various works in the water chamber. The tool 5 has a detachable configuration with respect to the manipulator 4 so that these tools 5 can be replaced. [Work in Water Chamber by Water-Chamber Working Apparatus] At the time of performing a work in a water chamber, the water-chamber working apparatus 1 is hung in a suspended state from the tube plate surface 137a and installed in the water chambers 131 and 133 (see FIG. 1). In this installed state, the clampers 23a and 23b of the base 2 clamp and hold the heat transfer tubes 132, thereby fixing the base 2 to the tube plate surface 137a. The manipulator 4 is coupled with the base 2 via the coupling link 3. Accordingly, in the installed state, the manipulator 4 is suspended from the ceiling (the tube plate surface 137a) of the water chambers 131 and 133 and held. The tool 5 corresponding to a work in the water chamber is attached to the end of the manipulator 4. An installing process of the water-chamber working apparatus 1 is described later. In this example, a worker performs a work in a water chamber by remote-controlling the water-chamber working apparatus 1 from a safe area outside of the water chambers 131 and 133. Accordingly, the work in the water chamber is performed without requiring the worker to enter into the water chambers 131 and 133. In the installed state of the water-chamber working apparatus 1, the manipulator 4 is suspended from the ceiling of the water chambers 131 and 133 (see FIG. 1). Consequently, the manipulator 4 is turned to change the orientation thereof, thereby realizing the work in the water chamber in a wide area, with the base 2 as a point of origin (see FIGS. 4 and 5). Specifically, the water-chamber working apparatus 1 is suspended from the tube plate surface 137a with the base 2 as a point of origin and is then installed. When the coupling link 3 is driven by remote control, the manipulator 4 turns around the rotation axis I of the coupling link 3, and the direction thereof can be changed in a circumferential direction of the water chambers 131 and 133. Furthermore, the manipulator 4 can move the tool 5 at the end to an arbitrary position in the water chambers 131 and 133 by changing its own orientation. With this configuration, because the tool 5 can be moved into every corner of the water chambers 131 and 133, the water-chamber working apparatus 1 can handle various works in the water chamber flexibly. For example, FIG. 4 depicts a state of a testing work of a welded part between the partition plate 134 and the tube plate 137, and FIG. 5 depicts a state of the testing work of the heat transfer tubes 132. As shown in FIGS. 4 and 5, it is understood that even if the base 2 is fixed in a certain position, by bending and deforming the manipulator 4 by turning, the manipulator 4 can move the tool 5 into every corner of the water chambers 131 and 133. In the steam generator 130, because the floor surface of the water chambers 131 and 133 has a hemispherical shape, it is not easy to install the water-chamber working apparatus on the floor surface. In this respect, because the water-chamber working apparatus 1 is installed by being suspended from the ceiling (the tube plate surface 137a) of the water chambers 131 and 133 (see FIG. 1), an installation work on the floor surface is not required, which is preferable. For example, in a configuration in which a manipulator is supported by a pillared turning support unit and installed in a water chamber (see Patent Literature 1), workers need to enter into the water chamber to install the turning support unit, which is not preferable. In the water-chamber working apparatus 1, in the installed state thereof, the coupling link 3 is coupled with the base 2 with the rotation axis I thereof facing downward from the tube plate surface 137a (see FIGS. 2 and 3). Therefore, the manipulator 4 can turn around a normal direction of the tube plate surface 137a as the rotation axis I in a state where the manipulator 4 is suspended with the base 2 as a point of origin (see FIGS. 4 and 5). Therefore, when the water chambers 131 and 133 have a quarter spherical internal shape with the tube plate surface 137a as the ceiling, a direction of the manipulator 4 can be changed in a circumferential direction of the water chambers 131 and 133. With this configuration, the tool 5 can be moved into every corner of the water chambers 131 and 133, thereby improving the workability of the work in the water chamber. In the present embodiment, the coupling link 3 is installed with the rotation axis I thereof being directed in the normal direction of the tube plate surface 137a (see FIGS. 2 and 3). However, the present invention is not limited thereto, and the rotation axis I of the coupling link 3 needs only to be downward from the normal direction of the tube plate surface 137a, and for example, the coupling link 3 can be arranged by inclining the rotation axis I by a predetermined angle with respect to the normal direction of the tube plate surface 137a. Furthermore, in the water-chamber working apparatus 1, in the installed state thereof, the manipulator 4 is coupled with the coupling link 3, with the reference axis m of the basic position thereof being inclined at the predetermined angle θ with respect to the rotation axis I of the coupling link 3 (see FIGS. 2 and 3). In this configuration, the reference axis m of the basic position of the manipulator 4 is inclined with respect to the normal direction of the tube plate surface 137a, in a state where the manipulator 4 is suspended with the base 2 as a point of origin. Accordingly, when the water chambers 131 and 133 respectively have a quarter spherical internal shape with the tube plate surface 137a as the ceiling, the direction of the manipulator 4 can be easily changed with respect to the floor surface and the wall surface of the water chambers 131 and 133. With this configuration, the tool 5 can be moved into every corner of the water chambers 131 and 133, thereby improving the workability of the work in the water chamber. Further, the water-chamber working apparatus 1 can be moved in the water chambers 131 and 133 within a predetermined area by moving the base 2 on the tube plate surface 137a (see FIG. 6). Specifically, the base 2 moves on the tube plate surface 137a by moving a clamping position with respect to the heat transfer tubes 132, and a position as a point of origin of the manipulator 4 (a fixed position of the base 2) can be moved on the tube plate surface 137a. With this configuration, the work in the water chamber can be performed using different positions in the water chambers 131 and 133 as a point of origin, and a work area of the work in the water chamber is expanded, thereby improving the workability of the work in the water chamber. Particularly, in the steam generator 130, because the floor surface of the water chambers 131 and 133 has the hemispherical shape, it is not easy to move the water-chamber working apparatus on the floor surface. In this respect, the water-chamber working apparatus 1 is installed on the flat tube plate surface 137a of the water chambers 131 and 133 by being suspended from the tube plate surface 137a. With this configuration, the movement of the water-chamber working apparatus 1 in the water chambers 131 and 133 is facilitated. The movement of the water-chamber working apparatus 1 (the base 2) is performed, for example, in the following manner. First, at the time of performing a work in the water chamber, the base 2 is fixed in a certain position by inserting the apical ends of the both clampers 23a and 23b into the heat transfer tubes 132 to clamp and hold the heat transfer tubes 132. At the time of moving the water-chamber working apparatus 1, while one of the clampers 23a (23b) clamps and holds the heat transfer tubes 132, the other one of the clampers 23b (23a) is pulled out of the heat transfer tubes 132 to release clamping to the heat transfer tubes 132. Next, the wing 22b (22a) extends (or retracts) to slidably displace its end, thereby to move the other clamper 23b (23a) along the tube plate surface 137a. Subsequently, the other clamper 23b (23a) inserts the apical ends again into the heat transfer tubes 132 to clamp and hold the heat transfer tubes 132. With this configuration, the clamping position of the other clamper 23b (23a) is moved. The one clamper 23a (23b) then moves its clamping position in the same manner, while the other clamper 23b (23a) keeps clamping and holding the heat transfer tubes 132. Because the both clampers 23a and 23b alternately move the clamping position, the base 2 can walk and move on the tube plate surface 137a. At the time of replacing the tool 5, the manipulator 4 is remote-controlled so that an apical end thereof protrudes from the maintenance hatch 138 of the water chamber 131 or 133 toward outside of the water chamber 131 or 133, still in a state where the water-chamber working apparatus 1 is installed in the water chamber 131 or 133 (see FIG. 1). In this state, the tool 5 attached to the apical end of the manipulator 4 is replaced. Accordingly, replacement of the tool 5 can be performed outside of the water chamber 131 or 133, while the water-chamber working apparatus 1 is kept to be installed in the water chamber 131 or 133. With this configuration, replacement of the tool 5 can be facilitated. [Installing Process of Water-Chamber Working Apparatus] FIGS. 7 to 17 are, respectively, a flowchart (FIG. 7) and explanatory diagrams (FIGS. 8 to 17) of an installing process of the water-chamber working apparatus shown in FIG. 1. In the installing process of the water-chamber working apparatus, the water-chamber working apparatus 1 is installed in the water chamber 131 or 133 in the following manner. A case where the water-chamber working apparatus 1 is installed in the inlet-side water chamber 131 is explained below. A rod-like jig 10 is inserted into the inlet-side water chamber 131 from the maintenance hatch 138, and the jig 10 is used to install a base carrying and attaching jig 11 on the tube plate surface 137a (Step ST1) (see FIG. 8). The base carrying and attaching jig 11 is a jig for attaching the base 2 to the tube plate surface 137a, and is inserted into the heat transfer tubes 132 and fixed to the tube plate surface 137a. A wire 12 for hoisting the base 2 is tied to the base carrying and attaching jig 11 (see FIG. 9). The base 2 and the coupling link 3 are carried into the inlet-side water chamber 131 and installed on the tube plate surface 137a (base installing step ST2) (see FIG. 10). At this time, the base 2 and the coupling link 3 are coupled with each other in advance and carried into the inlet-side water chamber 131. Furthermore, the base 2 is mounted with a winch 24 (see FIG. 11), and the wire 12 is wound by the winch 24 to hoist the base 2 from the maintenance hatch 138 to the tube plate surface 137a in the inlet-side water chamber 131. With this configuration, the heavy base 2 can be easily pulled up to the tube plate surface 137a of the inlet-side water chamber 131. The base 2 is then fixed to the tube plate surface 137a by inserting the apical ends of the clampers 23a and 23b into the heat transfer tubes 132 to clamp and hold the heat transfer tubes 132 (see FIG. 3). In this case, a worker pushes up the base 2 from below by using the rod-like jig 10, or pulls up the base 2 by using another rope (not shown), so that the base 2 is effectively hoisted up to the tube plate surface 137a. An attaching jig 13 is installed below the base 2 (Step ST3) (see FIG. 12). The attaching jig 13 is formed of a long plate-like member curved in a circular arc shape, and is used as a jig for coupling the manipulator 4 with the coupling link 3. An upper end of the attaching jig 13 is attached to a lower part of the base 2, and a lower end thereof is fixed to an inlet of the maintenance hatch 138. Accordingly, the attaching jig 13 is spanned from the lower part of the base 2 to the inlet of the maintenance hatch 138, thereby forming a slide-like guide. The rear stage 42 of the manipulator 4 is coupled with the coupling link 3 (Step ST4) (see FIG. 13). In this case, a rod-like jig 14 is inserted into the rear stage 42 and attached thereto. The rear stage 42 is then lifted on the attaching jig 13, pushed up by the jig 14 while being guided, and coupled with the coupling link 3. With this configuration, the heavy rear stage 42 can be easily carried from the maintenance hatch 138 to the base 2 on the tube plate surface 137a. Furthermore, the rear stage 42 can be easily guided to the coupling link 3 in the inlet-side water chamber 131 by the slide-like attaching jig 13. The jig 14 is detached from the rear stage 42 after coupling the rear stage 42 and the coupling link 3 with each other (see FIG. 14). Next, the front stage 41 of the manipulator 4 is coupled with the rear stage 42 of the manipulator 4, and the tool 5 is attached to the front stage 41 (Step ST5) (see FIG. 15). At this time, the front stage 41 is lifted on the attaching jig 13, pushed up by the jig while being guided, and coupled with the rear stage 42. With this configuration, the heavy front stage 41 can be easily carried from the maintenance hatch 138 to the rear stage 42. An upper end of the front stage 41 can be easily guided to a lower end of the rear stage 42 by the slide-like attaching jig 13. This work is performed by a worker from outside of the inlet-side water chamber 131 over the maintenance hatch 138. Subsequently, the attaching jig 13 is detached from the coupling link 3 and removed (Step ST6) (see FIG. 1). Accordingly, an installing step of the water-chamber working apparatus is complete. In the present embodiment, the base 2 is mounted with the winch 24, the base carrying and attaching jig 11 is installed on the tube plate surface 137a (Step ST1), and the wire 12 attached to the base carrying and attaching jig 11 is wound by the winch 24, thereby installing the base 2 by hoisting the base 2 up to the tube plate surface 137a in the inlet-side water chamber 131 (Step ST2) (see FIGS. 8 to 11). However, the present invention is not limited thereto, and for example, the base 2 can be installed on the tube plate surface 137a by using a small crane device 16 that can be fixed in the maintenance hatch 138 (see FIGS. 16 and 17). In the present embodiment, the winch 24 to be used at the time of carrying the base 2 into the inlet-side water chamber 131 (Step ST2) is mounted on the base 2 (see FIG. 11). However, the present invention is not limited thereto, and the winch 24 can be mounted on the base carrying and attaching jig 11 installed on a side of the tube plate surface 137a (this arrangement is not shown). In the present embodiment, the base carrying and attaching jig 11 is installed on the tube plate surface 137a by using the rod-like jig 10 (Step ST1) (see FIGS. 8 and 9). However, the present invention is not limited thereto, and for example, the base carrying and attaching jig 11 can be installed on the tube plate surface 137a by using the small crane device 16 that can be fixed in the maintenance hatch 138 (see FIGS. 16 and 17). [Specific Example of Base] FIG. 18 is a perspective view of an Example of the base of the water-chamber working apparatus shown in FIG. 1. FIG. 18 depicts an assembly of the base 2 and the intermediate link 3, and also depicts a state where the wings 22a and 22b of the base 2 are opened. FIGS. 19 to 21 are, respectively, a front view (FIG. 19), a plan view (FIG. 20), and a right side view (FIG. 21) of the base shown in FIG. 18. These drawings depict a state where the base 2 clamps the heat transfer tubes 132 on the tube plate surface 137a. FIGS. 22 and 23 are, respectively, a perspective view (FIG. 22) and a right side view (FIG. 23) of the base shown in FIG. 18. These drawings depict a state where the wings 22a and 22b of the base 2 are closed. In this Example, the base body 21 is formed of a substantially cubic frame member (see FIGS. 18 to 23). The base 2 has two sets of the wings 22a, 22a and 22b, 22b, designating a pair of the wings 22a and 22a, and 22b and 22b as a set. These wings 22a and 22b are respectively inserted from the four sides of the base body 21 and arranged. These wings 22a and 22b are arranged slidably (to be able to move forward and backward) with respect to the base body 21, so that the wings 22a and 22b can protrude and can be accommodated from the sides of the base body 21 mutually independently. Furthermore, the wings 22a and 22b are driven mutually independently by an actuator accommodated in the base body 21. In this configuration, when the base 2 opens the wings 22a and 22b, the wings 22a and 22b are slidably displaced and protrude from the sides of the base body 21. Furthermore, when the base 2 closes the wings 22a and 22b, the wings 22a and 22b are accommodated in the base body 21. Three clampers 23a and 23b as a set are brought into line and arranged on the wings 22a and 22b, matched with an installation interval of the heat transfer tubes 132 (see FIGS. 18 to 23). The clampers 23a and 23b include a clamping mechanism 231, a grip cylinder mechanism 232, and a main cylinder mechanism 233 (not shown. See FIGS. 24 to 33). The clamping mechanism 231 is arranged at the apical ends of the clampers 23a and 23b, and is inserted into the heat transfer tube 132 to enlarge or reduce a diameter thereof, thereby clamping the heat transfer tube 132. Specifically, the clamping mechanism 231 is formed of a tapered rod and a cotter. When the tapered rod is fitted to the cotter to open the cotter, the clamping mechanism 231 enlarges the diameter thereof to clamp the heat transfer tube 132 (clamping state: ON). Furthermore, when the tapered rod is pulled out from the cotter, the clamping mechanism 231 reduces the diameter thereof to release clamping to the heat transfer tube 132 (clamping state: OFF). The grip cylinder mechanism 232 drives the tapered rod of the clamping mechanism 231 to switch ON/OFF of the clamping state of the clamping mechanism 231 (enlarging and reducing of the diameter). Specifically, the grip cylinder mechanism 232 is formed of a cylinder, which uses the tapered rod of the clamping mechanism 231 as a piston. In a state where the clamping mechanism 231 is inserted into the heat transfer tube 132, when the grip cylinder mechanism 232 pulls in the tapered rod of the clamping mechanism 231 from a side of the heat transfer tube 132, the clamping state of the clamping mechanism 231 becomes ON. When the grip cylinder mechanism 232 pushes the tapered rod of the clamping mechanism 231 toward the heat transfer tube 132, the clamping state of the clamping mechanism 231 becomes OFF. The main cylinder mechanism 233 displaces the grip cylinder mechanism 232 forward and backward so that the clamping mechanism 231 is inserted into or pulled out of the heat transfer tube 132. Specifically, in a state where the base 2 is installed on the tube plate surface 137a, when the main cylinder mechanism 233 pushes up the grip cylinder mechanism 232, the grip cylinder mechanism 232 abuts on the tube plate surface 137a, and the clamping mechanism 231 is inserted into the heat transfer tube 132. Furthermore, when the main cylinder mechanism 233 pulls down the grip cylinder mechanism 232, the grip cylinder mechanism 232 is separated from the tube plate surface 137a and the clamping mechanism 231 is pulled out from the heat transfer tube 132. [Walking Logic of Base] FIGS. 24 to 33 are explanatory diagrams of a walking logic of the base. These drawings depict an Example of a basic operation of the wings 22a and 22b and the clampers 23a and 23b, when the base 2 walks on the tube plate surface 137a. The walking logic of the base 2 is not limited to the Example. In this Example, the base 2 moves along the tube plate surface 137a by sequentially moving the clamping positions of the clampers 23a and 23b with respect to the heat transfer tubes 132, while slidably displacing the wings 22a and 22b that are orthogonal to each other alternately. Furthermore, the pair of wings 22a, 22a and 22b, 22b facing each other are driven simultaneously. In this Example, a case where the base 2 moves from left to right in the drawings is explained (see FIGS. 24 to 33). In a suspended state of the base 2, the base 2 inserts the apical ends of all the clampers 23a and 23b into the heat transfer tubes 132 to clamp the heat transfer tubes 132 (see FIGS. 19, 21, and 23). At this time, at each of the clampers 23a and 23b, the main cylinder mechanism 233 pushes up the grip cylinder mechanism 232, and the grip cylinder mechanism 232 pulls in the tapered rod of the clamping mechanism 231 from the heat transfer tube 132, and thus the clamping state of the clamping mechanism 231 becomes ON (see FIG. 24). In this state, the base 2 is firmly fixed to the tube plate surface 137a. When the base 2 moves, first, the grip cylinder mechanism 232 pushes the tapered rod of the clamping mechanism 231 into the heat transfer tube 132 at the clampers 23b of the wings 22b that can be slidably displaced in the moving direction of the base 2 (see FIG. 25). The clamping state of the clamping mechanism 231 then becomes OFF. Subsequently, the main cylinder mechanism 233 pulls down the grip cylinder mechanism 232 at the clampers 23b (see FIG. 26). In this state, the base body 21 is supported by the clampers 23a (clamping state: ON) of the remaining wings 22a (not shown). The wings 22b having the clampers 23b with the clamping state being OFF are then slidably displaced in the moving direction of the base 2 (see FIG. 27). Subsequently, the main cylinder mechanism 233 pushes up the grip cylinder mechanism 232 at these moved clampers 23b so as to abut on the tube plate surface 137a (see FIG. 28). When the grip cylinder mechanism 232 pulls in the tapered rod of the clamping mechanism 231 from the heat transfer tube 132, the clamping state of the clamping mechanism 231 becomes ON. With this configuration, the clamping state of all the clampers 23a and 23b becomes ON. Subsequently, the grip cylinder mechanism 232 pushes the tapered rod of the clamping mechanism 231 toward the heat transfer tube 132 at the clampers 23a of the wings 22a (not shown) in a direction orthogonal to the moving direction of the base 2 (see FIG. 29). The clamping state of the clamping mechanism 231 then becomes OFF. The main cylinder mechanism 233 then pulls down the grip cylinder mechanism 232 at these clampers 23a (see FIG. 30). In this state, the base body 21 is supported by the clampers 23b of the remaining wings 22b in the moving direction. Subsequently, the base body 21 and the wings 22a with the clamping state being OFF (not shown) are slidably displaced in the moving direction of the base (see FIG. 31). Specifically, the wings 22b with the clamping state being ON is driven and slidably displaced in a direction opposite to the moving direction of the base 2, thereby mutually displacing the base body 21 and the wings 22a (not shown) with respect to the wings 22b with the clamping state being OFF. With this configuration, the base 2 moves with respect to the tube plate surface 137a (see FIG. 32). The main cylinder mechanism 233 then pushes up the grip cylinder mechanism 232 at the clampers 23a having moved together with the base body 21 to abut on the tube plate surface 137a (see FIG. 33). When the grip cylinder mechanism 232 pulls in the tapered rod of the clamping mechanism 231 from a side of the heat transfer tube 132, the clamping state of the clamping mechanism 231 becomes ON (refer back to FIG. 24). With this configuration, the clamping state of all the clampers 23a and 23b becomes ON, and the base 2 returns to the initial suspended state. By repeating the above operation, the base 2 can move an arbitrary distance along the tube plate surface 137a. Furthermore, by using the wings 22a and 22b that are orthogonal to each other, the base 2 can move in an arbitrary direction on the tube plate surface 137a (see FIG. 3). [Specific Example of Carrying Process of Water-Chamber Working Apparatus] In a carrying process of the water-chamber working apparatus 1 (see FIG. 7), at the base installing step ST2 (see FIG. 7 and FIGS. 10 to 12), it is preferable that the base 2 (an assembled body of the base 2 and the intermediate link 3) is carried into the water chamber 134 in a state where all the wings 22a and 22b are closed (see FIG. 22). The assembled body 2, 3 is then made more compact than a case where the base 2 is carried in with a state where the wings 22a and 22b are opened (see FIG. 18), and thus a carrying work into the water chamber 134 and an installation work on the tube plate surface 137a of the base are facilitated. Furthermore, at the time of performing an attachment work of the rear stage 42 and the front stage 41 of the manipulator 4 and the tool 5 (Steps ST4 and ST5) (see FIG. 7 and FIGS. 14 and 15), it is preferable that the base is installed on the tube plate surface 137a with all the wings 22a and 22b being opened (see FIGS. 19 to 21). Particularly, at the time of performing the attachment work of the front stage 41 of the manipulator 4 and the tool 5, the weight of the manipulator 4 acts on the clampers 23a and 23b of the base. Therefore, at the time of performing the attachment work, the base 2 is in a state where all the wings 22a and 22b are opened, thereby enabling to reduce the moment acting on the base 2. With this configuration, it is possible to prevent detaching of clamping of the base 2 at the time of performing the attachment work of the manipulator 4. Similarly, at the time of performing the replacement work of the tool 5, it is preferable that the base is installed on the tube plate surface 137a with all the wings 22a and 22b being opened (see FIGS. 19 to 21). With this configuration, the moment acting on the base 2 can be reduced, and thus it is possible to prevent detaching of clamping of the base 2 at the time of performing the replacement work of the tool 5. Furthermore, at the time of performing the attachment work (Step ST5) and the replacement work of the tool 5, it is preferable to cancel energization of the manipulator 4. That is, it is preferable to perform the attachment work and the replacement work of the tool 5 in a state where a joint portion of the manipulator 4 is made flexible with respect to an external force. Accordingly, the moment acting on the base 2 can be reduced, and thus it is possible to prevent detaching of clamping of the base 2 at the time of performing the attachment work and the replacement work of the tool 5. Further, at the time of performing a work in the water chamber (see FIG. 5), it is preferable that the base is installed on the tube plate surface 137a with all the wings 22a and 22b being opened (see FIGS. 19 to 21). Accordingly, the moment acting on the base 2 can be reduced, and it is possible to prevent detaching of clamping of the base 2 at the time of performing the work in the water chamber. Furthermore, at the time of moving the water-chamber working apparatus 1 (see FIG. 6), the clamping state of the base 2 becomes OFF at the wings 22a or 22b in one direction (for example, see FIG. 26 and FIG. 27). At this time, it is preferable that the barycenter of the manipulator 4 is arranged immediately below the base body 21 in a vertical direction (this arrangement is not shown). For example, at the time of moving the water-chamber working apparatus 1, the manipulator 4 first folds the front stage 41 (this state is not shown), so that the barycenter of the manipulator 4 becomes immediately below the base body 21 in a vertical direction, in a state where the base 2 is installed on the tube plate surface 137a with all the wings 22a and 22b being opened (see FIGS. 19 to 21). In this state, the base 2 is moved along the tube plate surface 137a (see FIGS. 24 to 33). With this configuration, the moment acting on the base 2 can be reduced, and thus it is possible to prevent detaching of clamping of the base 2 at the time of moving the water-chamber working apparatus 1. [Effect] As described above, according to the installation method of the water-chamber working apparatus 1, the water-chamber working apparatus 1 includes the base 2 that holds the heat transfer tubes 132 on the tube plate surface 137a and is fixed to the tube plate surface 137a, and the manipulator 4 that is coupled with the base 2, suspended in the water chamber 130 and arranged therein, and has a separable configuration. In this case, the base installing step ST2 of installing the base 2 on the tube plate surface 137a, and Steps ST4 and ST5 (manipulator coupling step) of carrying the separated manipulator 4 (the front stage 41 and the rear stage 42) into the water chamber 131, 133 sequentially and individually and coupling the manipulator 4 with the base 2 (the coupling link 3) are performed (see FIG. 7 and FIGS. 12 to 15). According to this configuration, in a configuration in which the manipulator 4 is suspended from the ceiling of the water chamber 131, 133 and arranged therein (FIG. 1), the installing process of the manipulator 4 is facilitated. For example, in a configuration in which a manipulator is carried into a water chamber in an integrated state and then installed, the weight of the manipulator becomes heavy, and thus a work for hoisting the manipulator to the ceiling of the water chamber is not easy. According to the installation method of the water-chamber working apparatus 1, at the time of coupling the separated manipulator 4 (the rear stage 42) with the base 2 (the coupling link 3) (manipulator coupling step ST4), the attaching jig 13 is spanned from the base 2 on the tube plate surface 137a to the maintenance hatch 138 of the water chamber 131 and then installed (Step ST3), and the separated manipulator 4 is coupled with the base 2 (the coupling link 3), while being guided by the attaching jig 13 (see FIG. 7 and FIGS. 12 to 14). With this configuration, the attaching jig 13 is used as a guide, and the separated manipulator 4 is carried from the maintenance hatch 138 of the water chamber 131 to the position of the base 2. Accordingly, the heavy rear stage 42 can be easily carried from the maintenance hatch 138 to the base on the tube plate surface 137a. According to the installation method of the water-chamber working apparatus 1, when the base 2 (and the coupling link 3) is installed on the tube plate surface 137a (base installing step ST2), the base 2 and the base carrying and attaching jig 11 installed on the tube plate surface 137a are connected with each other via the wire 12, and the base 2 is pulled up to the base carrying and attaching jig 11 by the wire 12 (FIGS. 9 to 11). With this configuration, the heavy base 2 can be easily pulled up to the tube plate surface 137a of the water chamber 131 and then installed. According to the installation method of the water-chamber working apparatus, at the time of replacing the tool 5, the water-chamber working apparatus 1 projects the apical end from the maintenance hatch 138 of the water chamber 131, 133 to outside of the water chamber 131, 133, in a state being installed in the water chamber 131, 133 (see FIG. 1). In this state, the tool 5 attached to the apical end of the water-chamber working apparatus 1 (the manipulator 4) is replaced. Therefore, the replacement work of the tool 5 can be performed outside of the water chamber 131, 133, while the water-chamber working apparatus 1 is installed in the water chamber 131, 133. With this configuration, the replacement work of the tool 5 is facilitated. According to the installation method of the water-chamber working apparatus, at the time of attaching or detaching the tool 5 (at the time of performing the attachment work (Step ST5), the time of performing a replacement work and the like), the joint portion of the manipulator 4 is made flexible with respect to an external force (this state is not shown). With this configuration, the moment acting on the base 2 can be reduced, and thus it is possible to prevent detaching of clamping of the base 2 at the time of performing an attachment work and a replacement work of the tool 5. As described above, the installation method of a water-chamber working apparatus according to the present invention is useful such that an installing process of a manipulator is facilitated in a configuration in which the manipulator is suspended from a ceiling of a water chamber and arranged therein. 1 water-chamber working apparatus 2 base 21 base body 22a, 22b wing 23a, 23b clamper 231 clamping mechanism 232 grip cylinder mechanism 233 main cylinder mechanism 24 winch 3 coupling link 31 mounting surface 4 manipulator 41 front stage 42 rear stage 5 tool 10 jig 11 base carrying and attaching jig 12 wire 13 attaching jig 14 jig 16 crane device 100 nuclear plant 110 reactor vessel 120 pressurizer 130 steam generator 131 inlet-side water chamber 132 heat transfer tube 133 outlet-side water chamber 134 partition plate 135 inlet nozzle 136 outlet nozzle 137 tube plate 137a tube plate surface 138 maintenance hatch 140 pump 150 primary coolant pipe
	summary
	claims	1. A method for portioning high radiation intensity waste, comprising:step S11: using a hanging device to move a net basket into a water container storing the high radiation intensity waste and using a pump to pump the high radiation intensity waste into the net basket;step S12: lifting the net basket and the high radiation intensity waste collected in the net basket to a space above the water container and putting the waste in a lead shield during lifting the net basket such that the high radiation intensity waste in the net basket can be dried;step S13: moving the net basket and the dried high radiation intensity waste collected in the net basket to a shielding container for temporary storage; andstep S15: sealing the net basket and the dried high radiation intensity waste collected in the net basket in the shielding container for temporary storage. 2. The method of claim 1, wherein before the step S15, further comprising step S14 of determining if the shielding container for temporary storage is suitable for containing another net basket, comprising confirming if in the shielding container for temporary storage is enough space suitable for accommodating another net basket and performing step S13 of moving another net basket and the high radiation intensity waste collected in the another net basket to the shielding container for temporary storage in response to confirming that in the shielding container for temporary storage is enough space suitable for accommodating another net basket.
052934128	description	DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows the vessel 1 of a pressurized-water nuclear reactor, mounted inside a vessel well 2 made within a concrete structure 3 constituting part of the reactor building of the nuclear power station. The vessel 1, which is of general cylindrical shape, is arranged in the vessel well 2 with its axis vertical, and has its lower part closed by a domed bottom and its upper part by a cover 1a, likewise of domed shape. Above the cover 1a of the vessel is arranged the set 4 of mechanisms for controlling the bars adjusting the reactivity of the core of the reactor, which consists of juxtaposed fuel assemblies placed inside the vessel 1. The vessel 1 is connected by means of connection pieces 5 to the pipelines 6 of the various loops of the primary circuit of the reactor, in which circulates the pressurized water coming into contact with the core assemblies within the vessel 1 and ensuring the heating and evaporation of feed water inside the steam generators of the power station. The concrete structure 3 forms, above the vessel well 2, a pool 8 which can be filled with water up to the vicinity of its upper level 8a, to make it possible to execute handling and maintenance operations on the inside of the vessel of the nuclear reactor during reactor shutdowns and after removal of the control set 4 and of the cover 1a of the vessel. The pool 8 comprises a part 9 which is placed laterally of the actual reactor pool located vertically in line with the vessel and in which the internal equipment of the reactor vessel can be arranged in order to carry out underwater maintenance or repair operations. The bottom of the vessel 1 has passing through it instrumentation conduits 10 which are connected to an instrumentation room located laterally of the vessel well 2. FIG. 2 illustrates the vessel 1 of a pressurized-water nuclear reactor during a dismantling operation executed by the use of the process according to the invention. The process according to the invention is put into practice after a permanent shutdown of the nuclear reactor and after unloading of the core assemblies and of the internal equipment of the nuclear reactor. After the shutdown and cooling of the nuclear reactor, the pool 8 is filled with water and the vessel cover is removed. The unloading of the core assemblies and the dismounting and disposal of the internal equipment of the vessel are then carried out under water. The fuel assemblies can be placed in containers to ensure their transport and disposal towards a reprocessing factory. The generally highly-irradiated internal equipment can be stored temporarily, before being dismantled under water and disposed of in transport containers. It is also possible at least partially to carry out the underwater dismantling of the internal equipment of the vessel on the inside of the latter. After the unloading of the vessel, the disposal of the internal equipment and the emptying of the pool, there is installed above the upper level 8a of the reactor pool a supporting structure 11 which consists of beams and on which rests the upper part of a lifting device 12, comprising particularly a mast of great length 13 which is arranged vertically along the axis of the vessel 1 and whose lower part is connected to the bottom of the vessel 1. The mast 13 is arranged within a tubular structure 14 placed vertically along the axis of the vessel 1 and having its upper part connected to the supporting structure 11. Arms 15 for centering and retaining the device 12 on the inside of the vessel, each having a jack 16 at its end, are fastened to the lower part of the tubular structure 14 and are arranged in the form of a star around this tubular structure. The jacks 16, which come to bear with their end part on the inner surface of the vessel 1, make it possible to carry out the centering and retention of the vessel 1 in relation to the tubular structure 14 and to the mast 13. The mast 13 comprises a toothing 13a over a substantial part of its length, the toothing 13a interacting with pawls 18 of a vessel-lifting mechanism 20 resting on the supporting structure 11 by means of a rotary thrust bearing 19. The rotating part of the bearing 19 can be driven in rotation about the vertical axis common to the vessel well 2 and to the vessel 1 by means of a motor 21. The lower part of supporting structure 11 carries a circular rail 22 on which are mounted movably in terms of rotation about the axis of the vessel well, by means of carriages, two monorails 23 and 23', shown in FIG. 5, allowing the displacement of hoists 24 in the entire zone located above the upper edge of the vessel 1 and in the storage pool 9 for the internal equipment as a result of the presence of fixed rails 25 and 25', in the extension of which the rotationally movable rails 23 and 23' can be placed. As will be explained later, the cutting of blocks 26 of irradiated material from the wall of the vessel is carried out substantially level with the bottom 9a of the pool 9 for the internal equipment, i.e., at the upper level of the vessel well. When a block 26 has been cut from the wall of the vessel 1, a hoist 24 can ensure that this block is picked up in any position and the block 26 transported into the pool 9 for the internal equipment, in which is arranged a container 27 for the storage and transport of the blocks 26 of irradiated material. The hoist 24 makes it possible to transport the blocks 26 between their cutting zone and their storage zone within the container 27. Zone containment walls 28 are placed at the upper level of the pool, below the supporting structure 11, in order to isolate the zone in which the cutting of the blocks 26 is carried out during the dismantling of the vessel 1, from the zone located above the pool, from which the control of the various operations put into effect for the dismantling is executed. Likewise, walls 29 make it possible to separate the pool for storing the internal equipment of the reactor from the zone 8 located vertically in line with the vessel, although passages are provided for the hoists 24 for transporting the blocks 26. Finally, the inner volume of the vessel 1 is isolated from- the reactor pool 8 by means of walls 30, in order to limit the radiation in the zone located above the vessel well 2. FIGS. 3 and 4 illustrate the lower part of the mast 13 of the lifting device 12 for the vessel 1. This lower part consists of a platen 32 which can be fastened to the lower part of the mast 13 in its axial direction by means of a threaded part 32a engaged in an internally-threaded hole at the end of the mast 13. The platen 32 comprises four orifices 33 and a centering stud 34 intended for ensuring the fastening and positioning of the end of the mast 13 on the bottom of the vessel 1. After unloading of the vessel, the connection pieces joining this vessel to the primary circuit and all the auxiliary pipework as well as the instrumentation tubes 10 of the vessel bottom are severed and then closed off. Four passage holes through the vessel bottom are machined or remachined in arrangements corresponding to the arrangements of the passage holes 33 of the platen 32 of the mast 13. It is also possible to fasten the vessel to the mast 13 by the use of a number of passage holes through the vessel bottom and a number larger than four of ties fastened in these holes, so as to employ ties and to machine holes of smaller diameter. The mast 13 can be installed by introducing the centering stud 34 into an instrumentation passage hole and by bringing the holes 33 into coincidence with the orifices of the vessel bottom which have been machined or remachined. All the passage orifices of the instrumentation tubes, with the exception of the orifices which have been remachined as appropriate, are closed off, and the fastening of the mast 13 is ensured by means of threaded rods 37 fastened to the platen 32 by nuts 35. A fastening plate 36 (see FIG. 2) having orifices in positions corresponding to the orifices 33 of the platen 32 is placed-under the vessel bottom in such a way that the threaded rods 37 engage into the orifices of this fastening plate 36. The fastening of the mast 13 is completed by nuts engaged on the rods 37 and coming to bear with a clamping effect against the lower face of the plate 36. The vessel 1 is thus firmly fixed at the end of the mast 13 which is mounted movably in the vertical direction along the axis of the tubular structure 14 and on the inside of this structure. Devices for wedging in the radial directions are also interposed between the tubular structure 14 and the mast 13, so as to ensure the guidance and retention of the mast 13 during its displacements in the vertical direction. Inflatable gaskets are likewise interposed between the mast 13 and the structure 14, so as to ensure the isolation or containment of the inner volume of the vessel 1 during the dismantling operations. Finally, as mentioned above, the vessel is retained by the arms 15 and jacks 16 in a position such that its axis is aligned with the axis of the mast 13 and of the tubular structure 14. As can be seen in FIG. 5, the supporting structure 11 comprises two parallel main beams 11a and 11b and four lateral beams 11c, 11d, 11e and 11f arranged in the form of a star around the axis of the vessel well of the reactor. The ends of the beams 11a to 11f rest on the concrete structure of the reactor, for example on the bearing surfaces of the anti-missile slab arranged vertically in line with the vessel well and at the upper level 8a of the reactor pool. FIG. 6 illustrates the entire apparatus for dismantling the vessel, during an operation for cutting the wall of the vessel. The corresponding elements in FIGS. 2 and 6 bear the same references, the apparatus, as illustrated in FIG. 6, comprising, in addition to the means for lifting the vessel and for handling the cut blocks 26, a horizontal cutting unit 40 and a vertical cutting unit 70 which are mounted on the tubular structure 14. The horizontal cutting unit 40 consists of a band saw 41 mounted on a support 42 fastened to the tubular structure 14 by means of a pivot bearing 43. The saw support 42 can be displaced, for the purpose of executing the cutting of the wall of the vessel, in the way which will be described in detail hereinbelow. The vertical cutting unit 70 consisting of a second band saw 71 makes it possible to separate the segment of the vessel wall cut by the saw of substantially horizontal displacement into blocks of irradiated materials 26 which are transported by the hoists 24 into the storage pool 9 for the internal equipment and deposited in a storage and disposal container 27. The cutting of the wall of the vessel over a particular height is carried out after the vessel 1 has been raised some distance in the vertical direction by means of the mast 13 and the lifting unit 20. The lifting unit 20 consists of a pawl device which will be described below. An appliance 46 for the suction and filtration of the gases in the storage pool 9 for the internal equipment is arranged in an isolated zone of this pool, in order to clear away the gases contaminated by radioactive materials present in the dismantling zone and in the storage zone for the irradiated material. An access orifice making it possible to dispose of the container 27 containing the blocks of irradiated material is provided in the biological containment wall 28, this orifice being closed during the dismantling operations by a slab 48 of radiation-absorbing material. As can be seen in FIG. 7, the pawl-type lifting device 20 comprises a support 50 resting on the supporting structure 11 by means of the pivoting bearing 19, the axis of which is the axis of the mast 13 coinciding with the axis of the vessel well 2 and the axis of the vessel 1. The bearing 19 is a roller bearing, the rollers 51 of which are inclined inwards and downwards so as to ensure perfect alignment of the axis of the mast 13 with the axis of the vessel well. The support 50 of the lifting device 20 has an annular shape and carries four fixed pawls, such as the pawl 18a, arranged at 90.degree. relative to one another about the axis of the mast 13, and mounted pivotably on the support 50 about horizontal axes, such as the axis 52a. The upper part of the support 50 constitutes a jack body 54 which is level with each of the fixed pawls 18a and in which is mounted a jack rod 55 of large cross-section, carrying at its upper end a support 56 in which a movable pawl 18b is mounted pivotably about a horizontal axis 52b. The pawls 18a and 18b comprise a profiled end part, the shape of which corresponds to the shape of the space delimited between two successive teeth of the toothing 13a of the mast 13. The pawls 18a and 18b are capable of pivoting through a particular angle of low amplitude between their position shown in solid lines in FIG. 7 and their position shown in broken lines. In the position shown in solid lines, the pawls are in engagement with the toothing 13a of the mast 13, and in their position shown in broken lines, they are in a position disengaged from the toothing 13a. FIGS. 8A to 8F show schematically the pawls 18a and 18b, the mast 13 and the actuating jack 54 of the movable pawls 18b in successive positions during a displacement phase in the vertical direction and towards the top of the mast 13, to the lower part of which the vessel 1 is fastened. In FIG. 8A, the mast 13 bears on the fixed pawl 18a in its engagement position within the toothing 13a. The jack rod 55 is in the low position. To execute the lifting of the mast 13 and of the vessel 1, the chamber of the jack 54 is fed in such a way as to displace the piston 55 and the support 56 upwards, as shown in FIG. 8B. The bearing pawl 18a, which has a ramp corresponding to the slope of the toothing 13a, comes into the disengaged position as a result of the sliding of its ramp on the toothing. The mast 13 rests on the movable pawl 18b which ensures that it is lifted by means of the jack 54. During the lifting of the mast 13, as shown in FIGS. 8C and 8D, the fixed pawl 18a disengages completely from the toothing as a result of an upward pivoting, and then escapes at the tip of the tooth with which it was in contact, when the tip of the tooth comes level with the end of the pawl 18a. The pawl 18a is then released and falls by pivoting back into the space located below the tip of the tooth, its inclined surface coming into contact with the slope of the toothing 13a. As illustrated in FIG. 8E, the double-action jack 54 is fed in such a way as to cause the descent of the rod 55 and of the movable pawl 18b which disengages from the toothing 13a, the mast 13 coming to rest on the fixed pawl 18a. As can be seen in FIG. 8F, the pawl 18b comes into position again in a space between two teeth located below the space in which this pawl 18b was engaged before the displacement of the mast 13, as shown in FIG. 8A. The pawls 18a and 18b are in identical positions in FIGS. 8A and 8F, the mast 13 having been displaced by one pitch of the rack 13a. The displacement of the jack 55 is equal to the pitch of the rack plus some play necessary for bringing about the engagement and disengagement of the pawls 18a and 18b. To carry out the lifting of a vessel of a pressurized-water nuclear reactor, four sets of pawls 18a and 18b and four jacks arranged at 90.degree. relative to one another about the axis of the mast 13 have been used. Each of the jacks has a lifting force of 100 tons, so that the total lifting capacity is 400 tons. The jacks have a stroke of 60 mm and the pitch of the toothing 13a of the mast 13 is 50 mm. The progressive raising of the mast 13 and of the vessel 1 is carried out in complete safety by means of the pawls, with which are associated devices for monitoring the correct engagement of the pawls in the toothing 13a. The lifting of the vessel is executed over a vertical distance corresponding to a particular number of displacement pitches of the rack, so as to provide above the level of the bottom of the pool for the internal equipment some wall height of the vessel 1, on which the cutting of blocks of material is carried out in a manner to be described below. FIGS. 9 and 10 illustrate in more detail the cutting machine 40 which consists of a band saw shown in FIG. 6. The band 41 of the saw is mounted on pulleys 44a and 44b driven in rotation by a motor means. The cutting of the wall of the vessel 1 is performed at a level located just above the level of the bottom 9a of the pool for the internal equipment. A guiding and centering device 60 is placed on the upper rim of the vessel well 2, level with the bottom 9a of the storage pool 9 for the internal equipment. The device 60 comprises bearing abutments 61 making it possible to carry out the centering of the vessel 1 and the alignment of its axis with the axis common to the well 2 and to the tubular structure 14, to which the cutting device 40 is fastened by means of the bearing 43. The guiding device 60 comprises a helical groove 62 of the axis of which corresponds to the axis of the vessel well 2. The cutting machine 40 has a guide roller 64 which moves along within the groove 62 during the cutting of the vessel. The groove 62 has an angular amplitude determining the rotational displacement of the saw blade 41 about the axis of the vessel, of the order of 30.degree.. The support 42 of the cutting machine, which is mounted rotatably on the tubular structure 14 by means of the pivot bearing 43, is displaced in rotation about the axis of the tubular structure 14 coinciding with the axis of the vessel, so as to describe an angle of 30.degree. about this axis, at the same time making a cut in part of the wall of the vessel 1 along a helix, the shape of which is homothetic with the helix formed by the guide groove 62. During this displacement, the support 42 of the cutting machine is also capable of pivoting in a vertical direction as a result of the construction of the bearing 43 in the form of a ball joint. The saw blade 41 taking the form of a band is driven in rotation by a drive motor. After the wall of the vessel 1 has been cut along a cylindrical sector of an amplitude of 30.degree. and along a helix the axis of which is the axis of the vessel, the cutting machine 40 is returned to its initial position, and the vessel is rotated oppositely to the cutting direction by means of the device 21 for setting the mast 13 in rotation, while at the same time it is raised over a height corresponding to one displacement pitch of the mast 13, so as to return the cutting blade 41 to the end of the helical incision previously made. A new cut of an amplitude of 30.degree. and of helical shape is made in the wall of the vessel as a result of the rotational displacement of the cutting machine 40 about the axis of the vessel. A cut of helical shape can thus be made over all or part of the periphery of the vessel by means of successive rotational displacements of the cutting machine 40 and translational and rotational displacements of the vessel 1. The total height H of the segment of the wall cut in the course of a complete revolution of the cutting machine is equal to the displacement pitch P of the mast 13 multiplied by the number of rotational displacements of the machine in the direction in which cutting is being carried out. For a rotational displacement of the machine of 30.degree., the number of displacements in the course of one revolution is 12, hence H=12P where the pitch P is 50 mm and the height H cut during each revolution is 600 mm. FIG. 13 shows a developed view of the helical cuts 65a, 65b, 65c, slightly inclined relative to the horizontal plane, which are made in the wall of the vessel 1 by the cutting machine illustrated in FIGS. 9 and 10. FIGS. 11 and 12 show the cutting machine 70 allowing straight cuts to be made in a direction forming a small angle relative to the vertical, so as to execute a sectioning of the vessel wall, in which one or more cuts, such as the cuts 65a, 65b, 65c shown in FIG. 13, have been made in a substantially horizontal direction. The cutting machine 70 illustrated in FIGS. 11 and 12 allows successive cuts 66 (FIG. 13) to be made in the wall of the vessel 1, in order to form blocks 26 of irradiated material which are delimited by the horizontal and vertical cuts. As explained above, the blocks 26 are picked up by a hoist 24 which makes it possible to transport these blocks into a storage container 27 arranged in the pool for the internal equipment. The machine 70 for cutting in the vertical direction comprises a support 72 mounted pivotably about a horizontal axis 73 on a second support 74 itself fixed to the rotating part 75 of a bearing mounted rotatably about the tubular structure 14. An actuating jack 76, of which the body is fixed to the tubular structure 14 and the rod is connected to the support 72 of the cutting machine 70, makes it possible to pivot the support 72 about the axis 73. FIG. 11 illustrates a first position, shown in solid lines, of the support 72 and two positions 72' and 72", shown by broken lines, which are obtained during the upward pivoting of the support 72 from its low position shown in solid lines. The actual cutting tool consists of a band saw mounted on the lower part of the support 72. The tensioning and driving of the band 71 of the saw are ensured by two pulleys 77a and 77b mounted loosely on the support 72, and by a driving pulley 78. The pivoting axis 73 of the support 72 in relation to the support 74 can be inclined slightly relative to the horizontal plane, so that the pivoting of the support 72 and of the saw band 71 under the effect of the jack 76 takes place in a plane slightly inclined relative to the vertical. This provides cuts, such as the cuts 66, inclined slightly in relation to the vertical direction. By rotating the support 72 about the axis of the vessel by means of the bearing 75, the cutting tool 70 can be placed in such successive positions that the band 71 executes the cutting of blocks 26 in the wall of the vessel 1, after horizontal cuts, such as the cuts 65a, 65b, 65c shown in FIG. 13, have been made. The centering of the vessel 1 and the alignment of its axis with the axis of the tubular structure 14 are ensured by the centering arms 15 and the jacks 16 and by the external centering devices 61. As can be seen in FIG. 13, the first cut 65a in the circumferential direction of the vessel, inclined slightly in relation to the horizontal plane, allows the horizontal cutting saw to penetrate into the metal of the vessel wall at a small angle and permits a progressive advance in the axial direction of the vessel. The first ring of metal delimited by a helical cut is sectioned by the vertical cutting saw according to the cuts 66, in order to form a first series of blocks 26 which can be disposed of and stored in a container placed in the pool for the internal equipment. The succeeding rings delimited by helical cuts and of substantially constant height are likewise sectioned by the vertical cutting saw, to form blocks 26 of substantially rectangular or square shape which are disposed of in sequence. The dismantling of the vessel is effected by a successive execution of substantially horizontal cuts and of substantially vertical cuts delimiting blocks 26 which are disposed of in sequence. During the cutting of the blocks for the purpose of dismantling of the vessel, the vessel can be filled with water up to a level below the part which is being cut or, if appropriate, can be empty of water. The water level in the vessel can be lowered during the progress of the cutting in the direction of the vessel bottom, before each operation of lifting the vessel between two successive series of cutting operations. The cutting operations are conducted at a substantially constant level located slightly above the upper level of the vessel well. This avoids the need to carry out the cutting on the inside of the vessel well and from the inner surface of the vessel, thus limiting the pollution of the concrete structures delimiting the vessel well by radioactive products. Moreover, the tools used for cutting are more easily accessible and it likewise becomes easier to control and guide them. FIGS. 14 and 15 and FIGS. 16 and 17 illustrate alternative embodiments of the horizontal cutting device and of the vertical cutting device making it possible to dismantle the vessel 1 by cutting blocks from its wall. The horizontal cutting device 80 illustrated in FIGS. 14 and 15 and the vertical cutting device 90 illustrated in FIGS. 16 and 17 consist of a respective circular saw 81 and 91 mounted movably in a radial direction in relation to the vessel 1, on a respective gantry 82 and 92 placed in a transverse direction above the vessel well. Where the device 80 for cutting in a horizontal direction is concerned, the disk 83 of the circular saw 81 is placed in a horizontal plane and mounted rotatably about a vertical axis. The advancing movement of the circular saw 81 in the direction of the arrow 84 allows a horizontal cut to be made in the wall of the vessel 1 and over its entire thickness, slightly above the vessel well and the bottom 9a of the pool for the internal equipment. If a cutting device comprising a circular saw is used, it is possible to make a perfectly horizontal cut, the penetration into the metal of the vessel wall being effected from inside the vessel and in a cross-sectional plane thereof. The circular saw for vertical cutting 91 comprises a saw disk 93 arranged in a vertical plane and mounted rotatably about a horizontal axis. The penetration into the metal of the vessel wall is effected from inside the vessel and in an axial plane. The cuts can be perfectly vertical and perpendicular to the horizontal cuts made previously. This provides blocks 26 of irradiated material of rectangular or square shape, delimited by the horizontal cuts and the vertical cuts. The cutting of the vessel wall is executed by rotating the vessel through a particular angle between two cutting operations involving successively the device 80 for cutting in the horizontal direction and the device 90 for cutting in the vertical direction. The cutting tools are controlled remotely, and the cutting operations are in all cases carried out in a zone making it possible to avoid major contamination of the reactor structures by radioactive products. FIG. 18 shows a vessel 101 of a water-cooled nuclear reactor during a preparatory phase prior to its dismantling. The vessel 101 is arranged inside a vessel well 102 within the concrete structure 103 of the nuclear reactor. The vessel well 102 opens out in its upper part into the pool 104 of the reactor. To dismantle the components of the reactor and particularly the vessel 101, the reactor is cooled after its permanent shutdown and the pool 104 is filled with water. The cover of the vessel is then dismounted and the core assemblies and of the internal equipment arranged in the vessel are unloaded underwater. The pool of the reactor is subsequently emptied and the vessel decontaminated, for example by the circulation of a chemical reagent in contact with its inner surface. The vessel is emptied and a device for the containment of the vessel well is installed. A scaffolding 107 is erected in the extension of the vessel well, underneath the hemispherical vessel bottom 101a. Cutting tool equipment is introduced into the vessel well so as to carry out the cutting of the pipework connecting the vessel to the reactor circuit, in the region of the connection pieces 105, 105' and 106, 106'. The cutting of the guide tubes or instrumentation tubes 108 passing through the bottom 101a of the vessel is also executed. This operation is conducted from the upper part of the scaffolding 107. A support 110, which can be seen particularly in FIG. 19, is put in place under the bottom 1a of the vessel. The support 110 comprises a bearing plate 10a which is fastened under the vessel bottom by means of rods 111 engaged in guide tubes or instrumentation tubes passing through the vessel bottom 101a, depending on the type of nuclear-reactor vessel for which the dismantling process according to the invention is used. The rods 111 have a threaded end which is engaged into an orifice passing through the plate 110a and onto which a nut is screwed. The nuts screwed onto the threaded end parts of the rods 111 make it possible to ensure the fastening of the plate 110a which carries abutments 112 coming to bear on the vessel bottom 101a during the tightening of the nuts. Before the displacement of the vessel 101 in successive steps in the vertical direction is executed to allow it to be cut in a zone located in the vicinity of the upper part of the vessel well 102, on the inside of the reactor pool 104, there are installed around the upper part of the vessel 101 an inflatable gasket 113 for closing the upper part of the vessel well 102 and guide jacks 114 for centering and guidance of the vessel 101 during its displacements in the vertical direction. Likewise installed in the pool 104 and in a room 104' arranged laterally of pool 4 are cutting and handling means which can be similar to the means described above and which enable cutting of blocks from the wall of the vessel 101 and the disposal of the cut blocks in storage containers. The reactor vessel 101 for which the dismantling process according to the invention is used, rests by means of supporting feet 116 on a supporting ring 115 fastened to the concrete structure 103 of the reactor at the upper level of the vessel well 102. In FIG. 21, the supporting ring 115 has been shown in a plan view, the upper surface of the ring 115 comprising eighteen successive zones 117 in the circumferential direction, the angular amplitude of each of these zones being 20.degree.. Fifteen zones 117 are intended for receiving the bearing surface of a supporting foot 116 of the vessel 101. The three remaining zones 117a, 117b and 117c, which are arranged vertically in line with the connection pieces joining the vessel to the reactor circuit, such as the connection pieces 105 and 105', do not receive supporting feet of the vessel 101 coming to bear on the ring 115. As can be seen in FIGS. 19 and 20, an initial lifting of the vessel can be carried out by means of jacks 120 which are interposed between the supporting ring 115 and some of the supporting feet 116. The jacks 120 are arranged within cutouts 121 of the supporting ring 115 and are brought to bear on wedging pieces 122. The height of the cutouts 121 is sufficient to ensure that a jack 120 bearing on the wedges 122 can be placed underneath a supporting foot 116 at the initial moment of prior lifting of the vessel 101. As can be seen in FIG. 21, the cutouts 121 are made in three zones distributed at 120.degree. around the ring 115 and corresponding to two successive zones 117 allowing the bearing of a supporting foot 116. Three sets of two jacks 120 are placed in the cutouts 121, each made in two successive zones 117 of the ring 115. The vessel is lifted passes by the simultaneous action of three jacks 120, each arranged in one of the three cutouts 121 distributed over the periphery of the vessel. After the vessel has been lifted over the height of a pass by the use of three jacks each located in a cutout 121, a wedging piece of a height corresponding to the height of the pass is placed underneath each of the jacks which have not been used for the lifting and which are arranged in the vicinity of the jacks which have executed the lifting, in the same cutout 121 of the ring 115. The next lifting pass is executed by using the jacks, the wedging of which has just been carried out, thus making it possible to raise the vessel an additional step. The wedging of the first set of three jacks which executed the lifting of the vessel is then carried out. This ensures the lifting of the vessel in successive passes by the placing of the wedging elements 23 (see FIG. 20) under each of the jacks 120. During the successive steps of the lifting of the vessel, wedging pieces are placed under all or some of the supporting feet 116 of the vessel which are not being used for lifting of the vessel as a result of interaction with a jack 120. At the end of the operation for the initial lifting of the vessel, there are placed underneath the supporting feet, in two zones 125 and 125', wedging pieces of sufficient height to maintain the vessel in the high position reached at the end of the initial lifting. The supporting ring 115 of the vessel is then cut is then performed, allow the passage of two parallel sections or girders 127, 127' intended for constituting part of the stationary support of the vessel during its subsequent displacement in successive steps in the vertical direction. The sections 127, 127' have the same height as the ring 115 and come to rest on the concrete structure 103 of the reactor in a lateral orifice 131, as can be seen in FIG. 19. The wedging pieces 128 make it possible to ensure good stability of the sections 127 and 127' which, together with the ring 115, constitute a stationary support on which the vessel rests during its lifting in successive steps and its cutting. As can be seen in FIGS. 22 and 23, at the end of the operation for the initial lifting of the vessel by the use of the jacks 120 and the wedging pieces 123, the vessel bottom 101a and the support 110 are at a particular height above the upper surface of the stationary support consisting of the ring 115 and of the sections 127. The vertical spacing present between the upper surface of the sections 127 and the lower bearing surface of the support 110 makes it possible to introduce between these elements a lifting module 130 which will be described below. The lifting module 130 is introduced through the lateral orifice 131 made in the concrete structure of the reactor, at a level located in the vicinity of the vessel bottom 101a. The rails 127 and 127' are arranged over the length of the orifice 131 and form a transfer track for the modular lifting element 130 when it is being put in place underneath the support 110 fixed to the vessel bottom 101a. The lifting element 130, which will now be described with reference to FIGS. 22, 23 and 24, comprise a raising device. The lifting element 130 comprises a raising device 132 and a modular supporting element 133 which are assembled together by means of keys 134. The raising device 132 takes the form of a frame comprising two parallel uprights 135a and 135b assembled together by means of spacers 136. The uprights and the spacers consist of metal plates assembled by welding. Fastened to the ends of the uprights 135a and 135b are jack boxes, such as 137a and 137b, inside each of which is placed a hydraulic jack, the body of which bears on the bottom of the corresponding jack box. As can be seen in FIG. 24, when the lifting element 130 is in vertical alignment with the vessel bottom 101a as shown in FIGS. 22 and 23, the jack boxes 137a and 137b of the raising device 132 are in vertical alignment with the supporting ring 115. The rods of the jacks 138a and 138b (FIG. 23) arranged inside the jack boxes 137a and 137b come to bear on the upper surface of the supporting ring 115. By feeding the jacks, such as 138a and 138b, of the raising device 132 in the direction bringing about the extension of the jack rods, the frame of the device 132 is raised in a direction perpendicular to the frame by means of the jack body coming to bear on the bottom walls of the corresponding jack boxes. By means of the frame of the raising device 132, the modular supporting element 133 fastened to the frame of the raising device 132 by means of the keys 134 is raised. The modular supporting element 133 takes the form of a frame of square cross-section, the faces 139 of which are connected at each of their ends to columns 140 in the region of the corners of the frame. The columns 140 are diametrically penetrated by orifices allowing the passage of the assembly keys 134 and having male or female frustoconical ends allowing a stable stacking of identical modular elements. The dimensions of the modular supporting element 133 are such that this modular element can come into place within the frame of the raising device 132 delimited by the uprights 135a and 135b and the spacers 136. In FIG. 24, the raising device 132 and the modular supporting element 133 are shown in their assembly position, the uprights 135a and 135b having through-orifices in alignment with the orifices of the columns 140 of the modular supporting element 133. In this position, the keys 134 can be introduced into the aligned orifices of the uprights 135a and 135b and of the columns 140. The columns 140 of the modular element 133 are arranged vertically in line with the supporting sections 127 and 127' when the lifting module 130 is in its operating position beneath the vessel bottom 101a. Feeding the jacks, such as 138a and 138b of the raising device 132 causes, by the extraction of the jack rods, the frame of the device 132 and of the modular supporting element 133 which is fastened thereto to to be raised. The upper part of the modular supporting element 133 taking the form of a turntable 141 (see FIG. 22) comes into contact with the lower surface of the plate 110a of the support 110 fixed to the vessel bottom 101a. The vessel 1 resting by means of the support 110 on the modular supporting element 133 can thereby be raised over a particular height corresponding to the amount of vertical displacement of the raising device 132. As can be seen in FIG. 25, when the lifting element 130 is in the high position obtained as a result of the extension of the jacks, such as 138a and 138b, a second modular supporting element 133' identical to the element 133 can be introduced underneath the element 133 raised by the device 132. The element 133 is displaced by transfer along the track consisting of the sections 127 and 127'. The amount of raising of the device 132 corresponds to the height of a modular lifting element, such as 133 or 133', plus a clearance allowing the passage of the element 133' underneath the frame of the device 132 and the modular supporting element 133 fastened within the frame of the device 132. The device 133' is arranged so as to be in exact vertical alignment with the modular element 133. The jacks, such as 138a and 138b of the raising device 132 are fed oppositely to the raising direction, in such a way that the element 133 comes to rest on the element 133', itself bearing on the sections 127 and 127', by means of frustoconical bearing surfaces of the columns 140. The assembly keys 134 making the connection between the frame of the raising device 132 and the modular supporting element 133 are then removed. The descending movement of the device 132 is then continued by feeding the jacks in the desired direction, up to the moment when the frame of the device 132 has returned to its initial position. The modular supporting element 133' is then in the position of the modular element 133 shown in FIG. 24. The modular supporting element 133' and of the raising device 132 can be assembled by introducing keys 134 into the aligned orifices of the uprights of the device 132 and of the columns of the modular supporting element 133'. A lifting element identical to the lifting element 130 and consisting of the raising device 132 to which the modular supporting element 133' is fastened is then placed underneath the modular supporting element 133 on which the vessel rests by means of the support 110. The vessel 101 is raised inside the well 102, in such a way that its upper part, consisting particularly of the vessel flange 101b, can be cut on the inside of the reactor pool 104 and in the vicinity of the upper part of the vessel well 102. It should be noted that, during the cutting at the end of the vertical displacement of the vessel by the agency of the raising device 132, the vessel 101, while it is being raised, rests by means of its bottom 101a, the support 110 and the modular supporting elements 133 and 133' on the rails 127 and 127' constituting elements of the stationary support of the vessel 101. The vessel 101 is therefore not suspended inside the vessel well 102 but rests, during the cutting operations, by means of its bottom on supporting elements bearing on the fixed structure of the reactor. The operations of cutting and handling the blocks cut from the wall of the vessel 101 can be conducted in the way described above. At the end of the cutting operation conducted on the part of the vessel located above the upper level of the vessel well after the vertical displacement of the vessel by means of the displacement device 132, the latter can execute a new vertical displacement of the vessel 101 which rests on the element 133', assembled together with the frame of the raising device 132, by means of the support 110 and the modular element 133. The vessel is raised by an amount slightly greater than the height of a modular supporting element, such as 133 and 133'. A third modular supporting element 133" identical to the modular supporting elements 133 and 133' is displaced by shifting on the sections 127 and 127' and is vertically aligned with the element 133' fixed to the frame of the raising device 132 and placed in the high position by this raising device. The jacks of the raising device 132 are subsequently fed oppositely to the raising direction, in such a way as to bring the element 133', on which the vessel rests by means of the element 133 and the support 110, to rest on the modular supporting element 133". The vessel 101 is now in a new lifting position in the vertical direction which allows a new segment of the vessel wall to be cut on the inside of the pool 104 above the upper level of the vessel well 102. The cutting of the vessel wall is thus executed in successive segments after each of the unit lifts of the vessel making it possible to place a new modular supporting element underneath the element, which is raised by means of the device 132, and to bring the vessel to rest, by means of the stacked modular elements, on this new element resting on the stationary support of the vessel formed by the rails 127. As can be seen in FIG. 26, the raising of the vessel in successive steps makes it possible to execute its cutting as far as the level of the domed bottom 101a. Successive supporting elements 133, 133', 133", . . . 133n have been interposed between the support 110 fixed to the vessel bottom and the stationary support of the vessel formed by the rails 127 and 127'. It is thus also possible to cut the vessel bottom 101a in the vicinity of the upper level of the vessel well 102 by the use of a specially adapted cutting tool outfit. It has been possible to execute the cutting of the vessel in the course of successive operations, during each of which the vessel rests, by means of a stack of modular supporting elements, on a stationary structure, itself bearing on the vessel bottom. The successive lifts of the vessel are of identical amount and are obtained from the same raising device which interacts successively with each of the modular supporting elements bearing on the stationary support. The process and apparatus according to the embodiment just described make it possible to obtain a vertical displacement of the vessel in successive steps, simply and in such a way that the vessel has a stable bearing during each of the cutting operations following a displacement in the vertical direction. The lifting of the vessel can be executed by a pull or a push on the vessel bottom by the use of means different from those described. Where the vessel is lifted by a push on the bottom, the initial displacement of the vessel in the vertical direction, making it possible to install the lifting element underneath the vessel bottom, can be effected by any means allowing the vessel to be raised by a push on its lower part. The raising device and the modular supporting element of the lifting unit employed for executing a unit lift of the vessel can have forms and structures different from those described. The push on the lower bottom of the vessel or, more generally, of the component being dismantled can be exerted by means of an intermediate support, as described, or directly on one or more push surfaces formed on the lower part of the component. The tools for cutting sections of irradiated material from the wall of the vessel can be different from a band saw or a circular saw. These cutting means can be non-mechanical, for example, an oxygen cutting torch, although thermal cutting processes give rise to the formation of vapor and of fine particles containing radioactive products, the trapping and filtration of which can be difficult to carry out. The cutting tools can comprise means for displacement and guidance over a complete revolution about the axis of the vessel. In this case, the dismantling of the vessel can be executed without the need to rotate the vessel about its axis. The disposal and storage of the sections of irradiated material can be carried out by means different from those described. The sections disposed of can be processed on the site of the reactor before their storage at a deactivation site or, on the contrary, transported to a processing factory and conditioned there for long-term storage. The cutting of the domed bottom of the vessel can be carried out by using the tools for cutting the cylindrical wall of the vessel as a result of accessory means for handling the domed bottom or, on the contrary, by using special tool outfits. Finally, the process according to the invention can be used for dismantling the vessel of any water-cooled nuclear reactor of the PWR or BWR type or for dismantling the internal equipment of such vessels. More generally, the process according to the invention can be used for carrying out the dismantling of any irradiated component of a nuclear reactor comprising at least one part of tubular shape arranged with its axis vertical.
052710459	summary	A portion of the disclosure of this patent document contains material which is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by anyone of the patent document or the patent disclosure, as it appears in the Patent and Trademark Office patent file or records, but otherwise reserves all copyright rights whatsoever. BACKGROUND OF THE INVENTION The present invention relates to apparatus and methods for monitoring and controlling the operation of commercial nuclear power plants. Conventionally, commercial nuclear power plants have a central control room containing equipment by which the operator collects, detects, reads, compares, copies, computes, compiles, analyzes, confirms, monitors, and/or verifies many bits of information from multiple indicators and alarms. Conventionally, the major operational systems in the control room have been installed and operate somewhat independently. These include the monitoring function, by which the components and the various processes in the plant are monitored; control, by which the components and the processes are intentionally altered or adjusted, and protection, by which a threat to the safety of the plant is identified and corrective measures immediately taken. The result of such conventional control room arrangement and functionality can sometimes be information overload or stimulus overload on the operator. That is, the amount of information and the variety and complexity of the equipment available to the operator for taking action based on such extensive information, can exceed the operator's cognitive limits, resulting in errors. The most famous example of the inability of operators to assimilate and act correctly based on the tremendous volume of information stimuli in the control room, particularly durinq unexpected or unusual plant transients, is the accident that occurred in 1978 at the Three Mile Island nuclear power plant. Since that event, the industry has focused considerable attention to increasing plant operability through improving control room operator performance. A key aspect of that improvement process is the use of human engineering design principles. Advances in computer technology since 1978 have enabled nuclear engineers and control room designers to display more information, in a greater variety of ways, but this can be counterproductive, because part of the problem is the overload of information. Improving "user friendliness" while maintaining the quantity and type of information at the operator's disposal has posed a formidable engineering challenge. SUMMARY OF THE INVENTION It is thus an object of the present invention to provide apparatus and method for nuclear power plant control and monitoring operations having the characteristics of concise information processing and display, reliable architecture and hardware, and easily maintainable components, while eliminating operator information overload. This objective should be accomplished while achieving enhanced reliability, ease of operation, and overall cost effectiveness of the control room complex. The solution to the problem is accomplished with the present invention by providing a number of features which are novel both individually and as integrated together in a control complex. The complex includes six major systems: (1) the control center panels, (2) the data processing system (DPS), (3) the discrete indication and alarm system (DIAS), (4) the component control system consisting of the engineered safeguard function component controls (ESFC) and the process component controls (PCC), (5) the plant protection system (PPS), and (6) the power control system (PCS). These six systems collect data from the plant, efficiently present the required information to the operator, perform all automatic functions and provide for direct manual control of the plant components. The control complex in accordance with the invention provides a top-down integrated information display and alarm approach that supports rapid assessment of high level critical plant safety and power production functions; provides guidance to the operator regarding the location of information to further diagnose high level assessments; and significantly reduces the number of display devices relative to conventional nuclear control complexes. The complex also significantly reduces the amount of data the operator must process at any one time; significantly reduces the operational impact of display equipment failures; provides fixed locations for important information; and eliminates display system equipment used only for off normal plant conditions. It is known that the nuclear steam supply system can be kept in a safe, stable state by maintaining a limited set of critical safety functions. The present invention extends the concept of the critical plant safety functions to include critical plant power production functions, in essence integrating the two functions so that the information presentation to the operator supports all high level critical plant functions necessary for power production as well as safety. The information display hierarchy in accordance with the invention includes a "big board" integrated process status overview screen (IPSO) at the apex, which provides a single dedicated location for rapid assessment of key information indicative of critical plant power production and safety functions. Further detail on the sources and trends of normal or abnormal parameter changes are provided by the DIAS. Both IPSO and the DIAS provide direct access and guidance to additional system and component status information contained on a hierarchy of CRT display pages which are driven by the DPS. The IPSO continually displays spatially dedicated information that provides the status of the plant's critical safety and power production functions. This information is presented using a small number of easily understood symbolic representations that are the results of highly processed data. This relieves the operator of the burden of correlating large quantities of individual parameter data, systems or component status, and alarms to ascertain the plant functional conditions. The IPSO presents the operator with high level effects of lower level component problems. The IPSO relies primarily on parameter trend direction, e.g., higher, lower, an alarm symbol color and shape, to convey key information. These are supplemented by values for selected parameters. The IPSO presents consolidated, simplified information to the operator in relatively small quantities of easily recognized and understood information. Furthermore, the IPSO compensates for the disadvantage inherent in recent industry trends towards presenting all information serially on CRTs, by enabling the operator to obtain an overview, or "feel" of the plant condition. Display of plant level overview on a large-format dedicated display addresses two additional operational concerns. First, operator tasks often require detailed diagnostics in very limited process areas. However, maintaining concurrent awareness of plant-wide performance is also necessary. Rather than relying on multiple operators in the control room to monitor respective indicators and the like on spatially separated panels, the IPSO can be viewed from anywhere in the control room and thus provides an operator a continuous indication of plant performance regardless of the detailed nature of the task that may be requiring the majority of his attention. In the preferred implementation, IPSO supports the assessment of the power and safety critical functions by providing for each function, key process parameters that indicate the functional status. For each function, key success paths are selected with the status of that success path displayed. The IPSO clearly relates functions to physical things in the plant. The critical functions are applied to power production, normal post trip actions, and optimal functional recovery procedures. The second level in the display information hierarchy in accordance with the present invention is the presentation of plant alarms from the DIAS. A limited number of fixed, discrete tiles are used with three levels of alarm priorities. Dynamic alarm processing uses information about the plant state (e.g., reactor power, reactor trip, refueling, shut-down, etc.) and information about system and equipment status to eliminate unnecessary and redundant alarms that would otherwise contribute to operator information overload. The alarm system provides a supplementary level of easily understood cueing into further information in the discrete indicators, CRTs and controls. Alarms are based on validated data, so that the alarms identify real plant process problems, not instrumentation and control system failures. The alarm features include providing a detailed message through a window to the operator upon the acknowledgment of an alarm and the ability to group the alarms without losing the individual messages. The tiles can dynamically display different priorities to the operator. The acknowledgment sequence insures that all alarms are acknowledged while at the same time reducing the operator task loading by providing momentary tones, then continuous alarm, followed by reminder tones to insure that the alarms are not forgotten. The operator has the ability to stop temporarily alarm flashing to avoid visual overload, and resume the flashing to insure that the alarm will eventually be acknowledged. The discrete indicators in the DIAS provide the third level of display in the hierarchy of the present invention. The flat panel displays compress many signal sources into a limited set of read-outs for frequently monitored key plant data. Signal validation and automatic selection of sensors with the most accurate signal ranges are also employed to reduce the number of control panel indicators. Information read-outs are by touch-screen to enhance operator interaction and include numeric parameter values, a bar form of analog display, and a plot trend. Various multi-range indicators are available on one display with automatic sensor selection and range display. The automatic calculation of a valid process representation parameter value, with the availability of individual sensor readings at the same display, avoids the need for separate backup displays, or different displays for normal operation versus accident or post-accident operation. Moreover, in another preferred feature of the invention, the parameter verification automatically distinguishes failed or multiple failed sensors, while allowing continued operation and accident mitigation information to the operator even if the CRT display is not available. Furthermore, the normal display information can be correlated to a qualified sensor, such as that used for post-accident monitoring purposes. At the information display level associated with control of specific components, dynamic "soft" controllers are provided with component status and control signal information necessary for operator control of these components. For the ESFC system, this information includes status lamp, on-off controls, modulation controls, open-closed controls, and logic controls. For the PCCS, the information includes confirm load, set points, operating range, process values, and control signal outputs. In the fourth level of the information hierarchy, dynamic CRT display pages are complementary to all levels of spatially dedicated control and information and can be accessed from any CRT location in the control room, technical support center, or emergency operations facility. These displays are grouped into a three level hierarchy that includes general monitoring (level 1), plant component and systems control (level 2), and component/process diagnostics (level 3). Display implementation is driven by the DPS and duplicates and verifies all discrete alarm and indicator processing performed in the DIAS. In the preferred implementation of the invention, the indicator, alarm, and control functions for a given major functional system of the plant are grouped together in a single, modularized panel. The panel can be made with cutouts that are spatially dedicated to each of the displays for the indicators, alarms, controls, and CRT, independent of the major plant functional system. This permits delivery, installation, and preliminary testing of the panels before finalization of the plant specific logic and algorithms, which can be software modified late in the plant construction schedule. This modularization is achievable because the space required on the panel is essentially independent of the major plant functional system to which the plant is dedicated. Both the alarms and indicators can be easily modified in software. The number of indicators and alarm tiles that can be displayed to the operator are not significantly limited by the available area of the panel, so that standardization of panel size and cutout locations for the display windows is possible.
	claims	1. An apparatus for measuring an aerial image, the apparatus comprising:a movable unit adapted to move a reflective extreme ultra-violet (EUV) mask disposed thereon in an x-axis and/or y-axis direction;an X-ray mirror arranged on the movable unit, the X-ray mirror being adapted to selectively reflect a coherent EUV light having a selected wavelength;a zoneplate lens that is located between the movable unit and the X-ray mirror, the zoneplate lens being adapted to focus the coherent EUV light on a portion of the reflective EUV mask; anda detector arranged on the movable unit, the detector being adapted to sense energy of the reflected coherent EUV light when the focused coherent EUV light is reflected by the portion of the reflective EUV mask,wherein NAzoneplate=NAscanner/4 and NAdetector=NAscanner/4σ, where NAzoneplate denotes a numerical aperture (NA) of the zoneplate lens, NAdetector denotes a NA of the detector, and NAscanner denotes a NA of a scanner, and σ denotes an off-axis degree of the scanner. 2. The apparatus as claimed in claim 1, further comprising an aperture between the reflective EUV mask and the detector. 3. The apparatus as claimed in claim 1, wherein the X-ray mirror comprises a multi-layer structure including at least one molybdenum layer and at least one silicon layer, which are alternately arranged. 4. The apparatus as claimed in claim 1, further comprising a EUV light generator, the EUV light generator comprising:a high power femtosecond laser adapted to output a high power femtosecond laser beam;a gas cell adapted to generate the coherent EUV light having a selected wavelength from the high power femtosecond laser; anda lens adapted to focus the high power femtosecond laser beam on the gas cell. 5. The apparatus as claimed in claim 4, wherein the gas cell is filled with a neon gas so as to optimize a production efficiency of a coherent EUV light having a wavelength of 13.5 nm. 6. The apparatus as claimed in claim 4, wherein the X-ray mirror is adapted to reflect the coherent EUV light emitted from the EUV light generator toward the portion of the reflective EUV mask at an angle of about 4° to about 8° with respect to a normal line of the reflective EUV mask. 7. The apparatus as claimed in claim 1, wherein the zoneplate lens is adapted to focus the reflected coherent EUV light on the portion of the reflective EUV mask at an angle of about 4° to about 8° with respect to a normal line of the reflective EUV mask. 8. The apparatus as claimed in claim 1, further comprising a computing unit adapted to reconstruct an image of the reflective EUV mask based on energy sensed by the detector. 9. An apparatus for measuring an aerial image of a pattern corresponding to a semiconductor pattern to be formed by scanning the pattern using a scanner, the apparatus comprising:a zoneplate lens arranged on a first side of an extreme ultra-violet (EUV) mask including the pattern, the zoneplate lens adapted to focus EUV light on a portion of the EUV mask at a same angle as an angle at which the scanner will be disposed with respect to a normal line of the EUV mask; anda detector arranged on a second side of the EUV mask and adapted to sense energy of the EUV light from the EUV mask,wherein NAzoneplate=NAscanner/n and NAdetector=NAscanner/nσ, where NAzoneplate denotes a NA of the zoneplate lens, NAdetector denotes a NA of the detector, and NAscanner denotes a NA of the scanner, σ denotes an off-axis degree of the scanner, and n denotes a reduction magnification of the scanner. 10. The apparatus as claimed in claim 9, further comprising a movable unit on which the EUV mask is arranged, the movable unit being adapted to move the EUV mask in an x-axis direction and/or an y-axis direction. 11. The apparatus as claimed in claim 9, wherein the EUV mask is a reflective EUV mask including a reflective material. 12. The apparatus as claimed in claim 11, wherein the detector is adapted to sense energy of reflected EUV light that is reflected from the reflective EUV mask. 13. The apparatus as claimed in claim 9, further comprising an EUV light generator and an X-ray mirror adapted to selectively reflect the EUV light from the EUV light generator. 14. The apparatus as claimed in claim 13, wherein the EUV light generator includes a high power femtosecond laser. 15. The apparatus as claimed in claim 9, wherein the EUV mask is a transmissive EUV mask. 16. The apparatus as claimed in claim 15, wherein the detector is adapted to sense energy of transmitted coherent EUV light that is transmitted through the transmissive EUV mask.
	abstract	A nuclear reactor includes at least: a pressure vessel including an upper vessel section and a lower vessel section connected by a mid-flange and containing primary coolant; a nuclear reactor core disposed in the lower vessel section and immersed in the primary coolant; and upper internals suspended from the mid-flange of the pressure vessel. The upper internals include at least internal CRDMs immersed in the primary coolant and control rod guide frames. To refuel, the nuclear reactor is depressurized. The upper vessel section is disconnected and removed while leaving the mid-flange in place with the upper internals remaining suspended from the mid-flange. The mid-flange is then removed with the upper internals remaining suspended from the mid-flange. The fuel is replaced, the mid-flange is placed back onto the lower vessel section with the upper internals remaining suspended from the mid-flange, and the upper vessel section is placed back and re-connected.
	claims	1. An x-ray mask tool for providing regions of reduced x-ray flux transmission, comprising: a silicon substrate having a thickness and substantially parallel first and second surfaces; a first x-ray attenuating layer supported by said substrate and applied to said first surface, said first x-ray attenuating layer comprising a first image forming pattern comprising a plurality of openings, wherein each of said openings comprise an area; and a second x-ray attenuating layer supported by said substrate and applied to said second surface, said second x-ray attenuating layer comprising a second image forming pattern, wherein some or all of said second x-ray attenuating layer eclipses some or all of said first x-ray attenuating layer and some or all of said areas. 2. The x-ray mask tool of claim 1 , wherein said first and second x-ray attenuating layers comprise a metal layer. claim 1 3. The x-ray mask tool of claim 2 , wherein said metal layer is selected from the group consisting of the, Transition series of metals listed in New IUPAC Group Numbers 4-12 of the Period Table of elements, aluminum, tin, and alloys thereof. claim 2 4. The x-ray mask tool of claim 2 , wherein said metal layer consists essentially of a layer of gold. claim 2 5. The x-ray mask tool of claim 2 , wherein said metal layers have a thickness such that a source of x-ray radiation is attenuated in intensity from about 50% to about 100%. claim 2 6. The x-ray mask tool of claim 2 , wherein said metal layer is applied by a deposition process. claim 2 7. The x-ray mask tool of claim 6 , wherein said metal deposit is deposited by electroless deposition. claim 6 8. The x-ray mask tool of claim 6 , wherein said metal is deposited by a process selected from the list consisting of thermal or particle vapor deposition, chemical vapor deposition, sputter deposition, molecular beam epitaxy. claim 6 9. The x-ray mask tool of claim 2 , wherein said metal layer is applied by a plating process. claim 2 10. The x-ray mask tool of claim 9 , wherein said metal deposit is deposited by electroplating. claim 9 11. An x-ray mask tool for providing regions of reduced x-ray flux transmission, comprising: a silicon substrate having a thickness and substantially parallel first and second surfaces: a first x-ray attenuating layer supported by said substrate and embedded into said substrate thickness, said first x-ray attenuating layer comprising a first image forming pattern comprising a plurality of openings, wherein each of said openings comprise an area, wherein said first image forming pattern is about flush with one of said first or second surface; and a second x-ray attenuating layer supported by said substrate and applied to either of said first or second surfaces, said second x-ray attenuating layer comprising a second image forming pattern, wherein some or all of said second x-ray attenuating layer eclipses some or all of said first x-ray attenuating layer and some or all of said areas.
	description	The present invention relates to molten salt reactors. More particularly, some implementations of the described invention relate to systems and methods for providing a thorium molten salt reactor. In this regard, some implementations of the reactor are configured to rotate a reactor core to vary a flow rate of fissionable fuel through the reactor. Moreover, in some implementations, the reactor core houses two or more fuel wedges that each define at least one fuel channel that extends through the wedges. In some implementations, one or more of the wedges, components of the reactor core, and/or reflectors surrounding the core are configured to be replaced relatively easily. The need for electrical energy across the world appears to be ever growing. In this regard, electricity for power grids across the world is generated through a wide variety of methods. In one example, coal, natural gas, petroleum, another fossil fuel, wood, waste, and/or one or more other fuel sources are burned to create heat, which is then used to turn a turbine (e.g., via pressure applied to the turbine by steam that is created, and/or air that is expanded, by the heat) and ultimately to turn an electrical generator. In another example, wind or water is used to create electricity as such media move past (or otherwise interact with) a generator. For instance, water passing through a hydroelectric dam, water passing a water wheel, air passing a wind turbine, and tidal water passing a tidal energy converter have each been found to be effective methods for generating electricity. In still other examples, sunlight (e.g., via solar cells, solar thermal energy generators) and/or geothermal energy (e.g., via vapor-dominated reservoirs, liquid-dominated reservoirs, enhanced geothermal systems, geothermal heat pumps, etc.) are used to generate electricity. Moreover, in another example, nuclear energy is used to generate electricity. In this regard, uranium or another fissionable material is typically used to generate heat that converts water to steam, which, in turn, rotates one or more turbines that are coupled to one or more electric generators. Although many conventional methods for generating electricity have proven to be very useful, such methods are not necessarily without their shortcomings. For instance, some methods that generate electricity by burning fossil fuels, also produce relatively large amounts of pollution and carbon dioxide gas, while depleting the Earth's limited natural resources. Additionally, some methods for generating electricity via solar-power and/or wind-power systems are only able to generate electricity when they are exposed to a sufficient amount of sunlight and/or wind—factors that are not necessarily available 24 hours a day and 365 days a year. Moreover, as some geothermal and hydroelectric power systems rely upon, and are limited by, the natural conditions on which such systems rely, many such systems are optimally (and sometimes only) placed in specific locations (e.g., at tectonic plate boundaries, rivers, reservoirs, coast lines, etc.) that have the requisite conditions. Furthermore, some nuclear power plants also have shortcomings, which can include potential environmental damage associated with potential meltdowns, accident, uranium mining, and nuclear waste generated by the power plants. Thus, while systems and methods currently exist that are used to generate electricity, challenges still exist, including those listed above. Accordingly, it would be an improvement in the art to augment or even replace current techniques with other techniques. The present invention relates to molten salt reactors. More particularly, some implementations of the described invention relate to systems and methods for providing a thorium molten salt reactor. In this regard, some implementations of the reactor are configured to rotate a reactor core to vary a flow rate of fissionable fuel through the reactor. Moreover, in some implementations, the reactor core houses two or more fuel wedges that each define at least one fuel channel that extends through the wedges. In some implementations, one or more of the wedges, components of the reactor core, and/or reflectors surrounding the core are configured to be replaced relatively easily. Some implementations include a molten salt reactor that includes a reactor core comprising graphite (e.g., a substantially pure and/or other suitable graphite) and defining an internal space containing multiple fuel wedges that each define a fuel channel that is configured to allow a thermonuclear or fissionable fuel to flow from a first end to a second end of each of the wedges. Some implementations further include a molten salt reactor that includes a reactor core that is disposed in a reactor housing and that comprises graphite and defines multiple fuel channels that run between a first end and a second end of the reactor core. In some cases, the reactor core comprises one or more fuel ingress ports (or inlets) and egress ports (or outlets), and the reactor core is rotatably received within the reactor housing such that the fuel ingress and egress ports are configured to become at least one or more occluded and less occluded as the reactor core rotates within the housing. Additionally, some implementations include a molten salt reactor that includes a reactor core that is disposed in a reactor housing and that comprises graphite and defines an internal space with multiple fuel wedges being received within the internal space, wherein the fuel wedges each define a fuel channel that is configured to allow a fissionable fuel to flow from a first end to a second end of each of the wedges. In some cases, a fuel pin rod is disposed between at least two of the wedges, with the fuel pin rod defining an internal fuel conduit. Additionally, in some cases, the reactor core further comprises a fuel ingress port and a fuel egress port, and the reactor core is rotatably received within the reactor housing such that the fuel ingress and egress ports are configured to become at least one of (i) more occluded and (ii) less occluded as the reactor core rotates within the housing. While the methods and processes of the present invention may be particularly useful for generation of electricity, those skilled in the art will appreciate that the described systems and methods can be used in a variety of different applications and in a variety of different areas of manufacture. For instance, instead of comprising a generator, some implementations of the described systems and methods are configured to provide heat to one or more buildings, stadiums, neighborhoods, and/or other structures and facilities. In some other cases, the described systems are configured for desalination and/or to distill water (e.g., to create drinking (or relatively clean) water from salt water or another non-potable and/or polluted water source). In still other cases, the described systems and methods are configured to provide energy for use in” oil shale and oil sand production, molten pool thermal electric sterling motors, onshore and offshore power plants, automobiles, trains, ships, submarines, airplanes, helicopters, space shuttles, off-planet applications (e.g., on the moon), the production of hydrogen fuels, the production of bio gas applications, in locations where portable power stations are useful (e.g., by attaching the molten salt reactor to a trailer, a skid, a vehicle, etc.), providing geothermal liquid enhancers, heating water for aqua culture, and/or for a wide variety of other suitable purposes. These and other features and advantages of the present invention will be set forth or will become more fully apparent in the description that follows and in the appended claims. The features and advantages may be realized and obtained by means of the instruments and combinations particularly pointed out in the appended claims. Furthermore, the features and advantages of the invention may be learned by the practice of the invention or will be obvious from the description, as set forth hereinafter. The present invention relates to molten salt reactors. More particularly, some implementations of the described invention relate to systems and methods for providing a thorium molten salt reactor. In this regard, some implementations of the reactor are configured to rotate a reactor core to vary a flow rate of fissionable fuel through the reactor. Moreover, in some implementations, the reactor core houses two or more fuel wedges that each define at least one fuel channel that extends through the wedges. In some implementations, one or more of the wedges, components of the reactor core, and/or reflectors surrounding the core are configured to be replaced relatively easily. The following disclosure is grouped into two subheadings, namely “MOLTEN SALT REACTOR” and “REPRESENTATIVE OPERATING ENVIRONMENT.” The utilization of the subheadings is for convenience of the reader only and is not to be construed as being limiting in any sense. While the described systems can comprise any suitable component, FIG. 1A shows a representative embodiment in which the described molten salt reactor system 10 comprises one or more heaters 15, reactors 20, heat exchangers 25, steam generators 30, and/or electric generators 35. Additionally, while the described systems can function in any suitable manner, FIGS. 1A-1C show that, in some embodiments, the heater 15 (not shown in FIGS. 1B-1C) is configured to heat one or more fissionable fuel sources (not shown) and/or carrier mediums (not shown) (collectively, the “fuel”) into a molten state and to pass the molten fuel to the reactor 20. In some embodiments, the reactor 20 is configured to function as a neutron moderator that is designed to reduce the speed of fast neutrons in the molten fuel and to convert such neutrons into thermal neutrons that allow the fuel to sustain a nuclear chain reaction (or to be in a critical state), which further heats the fuel. In accordance with some embodiments, FIGS. 1A-1C show that heated fuel (not shown) is cycled in a first fluid line 40 between the heat exchanger 25 and the reactor 20 such that as the fuel passes through the heat exchanger, heat from the heated fuel is passed to a heat transfer medium (not shown) running through a second fluid line 45 that is separate from the first fluid line. In some embodiments (as shown in FIGS. 1A-1C), the second fluid line 45 extends between the heat exchanger 25 and the steam generator 30. In some such embodiments, the system is configured move the heat transfer medium from the heat exchanger 25 (where the medium is heated), through the steam generator 30 (where heat from the heat transfer medium causes water in the steam generator to turn into steam), and the heat transfer medium is then returned to the heat exchanger (where the transfer medium is reheated). In accordance with some embodiments, FIG. 1A shows that steam from the steam generator 30 is optionally directed to the electric generator 35 (e.g., via a third line 50 or otherwise), where the steam is used to turn one or more turbines to generate electricity. To provide a better understanding of the described system 10, each of the aforementioned components of the system is described below in more detail. With respect to the heater 15, the heater can comprise any suitable component that allows it to heat the fissionable fuel to a molten state and to then pass the molten fuel to the reactor 20. Indeed, in accordance with some embodiments, FIG. 1A shows the heater comprises a container 55, which is configured to hold the fuel, and a heat source 60 that is configured to heat the fuel. The container 55 can have any suitable characteristic that allows the heater 15 to function as intended. For instance, the container can: be any suitable size (e.g., hold a volume of fuel that is larger than, smaller than, and/or approximately equal in volume to an internal volume of a reactor core in the reactor 20), be made of any suitable materials (e.g., comprise one or more nickel alloys, low-chromium nickel-molybdenum alloys (such as HASTELLOY-N™), metals, cements, ceramics, synthetic materials, and/or any other suitable materials), and have any suitable component (e.g., one or more drains that are configured to drain molten fuel to the reactor and/or another container, pumps that are configured to force the fuel to the reactor and/or another suitable container, mixers that are configured mix various components of the fissionable fuel, vents, valves, lids, seals, thermostats, sensors, fans, and/or other suitable components) that allow the heater to function as intended. Indeed, in some embodiments, the container comprises one or more agitators, shakers, orbital mixers, and/or other mixers that are capable of mixing the various components of the fuel as it is cracked. With regards to the heat source 60, the heat source can comprise any suitable heat source that is capable of converting (or cracking) one or more components of the fuel to a molten state. Some examples of suitable heat sources include, but are not limited to, one or more burners, heating coils, heating elements, ovens, fires, solar heaters, and/or other suitable heat sources that are capable of liquefying the fuel. The heat source may also use any suitable energy source to heat the container 55 to a desired temperature. Some non-limiting examples of such energy sources include fossil fuels, coal, electricity, wood, biomass, biofuel, and/or any other suitable source. Once the fuel has been cracked, the fuel can be moved from the heater 15 to the reactor 20 in any suitable manner. In one example, the fuel is pumped (e.g., via one or more pumps 22, as shown in FIG. 1A) from the heater 15 to the reactor 20. In another example, the fuel is allowed to drain into the reactor via gravity. In still another example, a reactor core comprises a vacuum that is configured to draw the fuel into the core (e.g., once a valve is opened). In some embodiments, once the heater 15 has cracked the fuel and the fuel has gone critical in the reactor 20, the heater is no longer needed to maintain the fuel in a molten state. Accordingly, while the heater 15 can have any suitable relationship with the reactor 20, in some embodiments, once the fuel has been cracked by the heater and been introduced into the reactor, the heater is disconnected from the reactor, a valve between the heater and the reactor is closed, and/or the system 10 is otherwise modified such that fuel in the reactor does not flow back into the heater until desired. Thus, in some embodiments, the heater is simply used to start and to restart the system (e.g., when the system is started for the first time and/or after the system has been shut down for maintenance and/or any other reason). With respect to the fuel, the fuel can comprise any suitable ingredient or ingredients that allow the fuel to be heated into a molten state and to go critical in the reactor 20. Indeed, as mentioned above, in some embodiments, the fuel comprises a fissionable fuel source and a carrier medium. Some examples of suitable fissionable fuel sources include, but are not limited to, U-233, thorium U-232, U-235, Th-232, Th-228, Th-230, Th-234, nuclear waste from a nuclear reactor (e.g., one or more light water, and/or other nuclear reactors), fuel un-cladded nuclear spent fuel rods, nuclear spent fuel rod pellets, Pu-239, UF4-LiF, PuF3, and/or any other suitable fissionable material and/or precursor to a suitable fissionable material. Indeed, in some embodiments, the fissionable fuel source comprises U-232, U-233, and U-235. Additionally, in some embodiments, the fuel comprises one or more other atomic elements that are configured to be mixed (e.g., homogeneously or otherwise) into the fuel. The various components of the fissionable fuel source can be present in the fuel at any suitable concentrations. Indeed, in some embodiments in which the fuel comprises U-232 and U-233, the two components are respectively used at a molar ratio between about 100:1 and 1:100, or at any suitable subrange thereof. Indeed, in some embodiments, when the fuel is initially added to the reactor, the fuel respectively comprises U-232 and U-233 at a molar ratio between about 6:1 and about 2:1 (e.g., at a ratio of about 4:1) (though other materials (e.g., atomic elements and/or other suitable materials) can also be mixed therein). With respect to the carrier medium, the fuel can comprise any suitable carrier medium that allows the fuel to go critical in, and that is safe for use with, the reactor 20. Some examples of such carrier mediums include, but are not limited to, KNO3 (potassium nitrate), NaNO3 (sodium nitrate), ThF4 (thorium fluoride), LiF (lithium fluoride), BeF2 (beryllium fluoride), FLiBe (a molten mixture of lithium fluoride and beryllium fluoride), FLiNaK (a metal salt mixture of LiF, NaF (sodium fluoride), and KF (potassium fluoride)), and/or any other suitable salt or salts. Indeed, in some embodiments, the carrier medium comprises potassium nitrate and/or sodium nitrate. In some other embodiments, the carrier medium comprises potassium fluoride and/or sodium fluoride along with one or more other high thermal salts that can become a homogenous atomic element blend in the fuel. Where the carrier medium comprises more than one ingredient, the various ingredients can be present at any suitable concentration in the fuel. Indeed, in some embodiments, the two components (e.g., potassium nitrate and sodium nitrate, potassium fluoride and sodium fluoride, etc.) are respectively used at a molar ratio between about 100:1 and 1:100, or at any suitable subrange thereof. In this regard, in some embodiments, the carrier medium respectively comprises potassium nitrate and sodium nitrate at a molar ratio between about 6:1 and about 0.5:1 (e.g., at a ratio of about 1.5:1). In some embodiments, the fuel includes a mixer of 60% potassium nitrate to 40% sodium nitrate, along with one or more other homogenous salt blends Turning now to the reactor 20, the reactor can comprise any suitable component and characteristic that allows the fuel to obtain and/or sustain a nuclear chain reaction by passing through the reactor. By way of non-limiting illustration, FIGS. 2A-2B show that, in some embodiments, the reactor 20 comprises one or more housings 65, reactor cores 70, reflectors 75, fuel inlets 80, fuel outlets 85, reactor control mechanisms 90, and/or drains 95. With regards to the housing 65, the housing can comprise any suitable component or characteristic that allows the housing to contain the reactor core 70 and to prevent undesired amounts of neutrons and/or gamma radiation from escaping housing. While the housing can further comprise any suitable component that allows it to substantially envelope the core reactor, FIGS. 2A-3B show that, in some embodiments, the housing 65 includes a container 100 having a cover 105 that is selectively removable and/or openable to provide access to the reactor core, reflectors 75, and/or any other suitable component. In some such embodiments, the housing 65 (as shown in FIGS. 2B-2C) further comprises one or more seals 110, which may include, but are not limited to, one or more carbon seals, carbon ropes, carbon-containing materials, rubber seals, gaskets, and/or any other suitable sealing material. Indeed, in some embodiments, FIG. 2C shows the seal 110 between the cover and the container comprises one or more carbon ropes 115. The housing 65 can comprise any suitable material that allows it to function as intended. Indeed, in some embodiments, the housing comprises one or more metals (e.g., lead, steel, tungsten, nuclear grade metals, and/or any other suitable metals), alloys (e.g., one or more nickel alloys, low-chromium nickel-molybdenum alloys (e.g., HASTELLOY-N™), nuclear grade alloys, and/or other suitable alloys), cements, types of nuclear gunnite, types of nuclear shotcretes, types of mortar, types of reinforced cement, ceramics, synthetic materials, polymers, plastics, hydrogen-based materials, fiberglass, and/or any other suitable materials. In some embodiments, however, the housing comprises a low-chromium nickel-molybdenum alloy, such as HASTELLOY-N™. Additionally, in some embodiments, the housing further comprises one or more liners (e.g., lead, steel, and/or plastic liners), a secondary containment housing, and/or one or more reinforcement elements (e.g., steel rods, steel meshes, fiber reinforcements, composites, and/or any other suitable reinforcements). Turning now to the reactor core 70, the core can comprise any suitable component or characteristic that allows it to act as a moderator as the fuel passes through it, such that the core is able to help the fuel reach (and/or maintain) a critical state. Some non-limiting examples of such elements include a reactor core tube and one or more end caps, internal moderators, and/or diffusers. With reference to the reactor core tube, the tube can comprise any suitable characteristic that allows it to function as described herein. In this regard, the tube can be any suitable shape, including, without limitation, being cylindrical, polygonal, cuboidal, symmetrical, asymmetrical, tubular, spherical, prism-shaped, and/or any other suitable shape. By way of non-limiting illustration, FIG. 4A shows an embodiment in which the reactor core tube 120 is substantially cylindrical and tubular in shape, having a first end 125 and a second end 130 with an internal space 135 defined between the two ends. The reactor core tube 120 can be any suitable size. Indeed, while the reactor core tube can be any suitable length, in some non-limiting embodiments, the tube has a length that is between about 0.05 meters (m) and about 30 m, or any length that falls in such range. In this regard, some embodiments comprise a reactor core tube having a length between about 0.3 m and about 2.5 m (e.g., between about 0.5 m and about 0.8 m). In other embodiments, the reactor core tube has length that is even greater than the lengths set forth herein. While the reactor core tube 120 can have any suitable width or diameter, in some embodiments, the tube has an inner diameter (or ID) that is between about 0.04 m and about 6 m, or any width/diameter that falls in such range. In this regard, some embodiments of the reactor core tube comprise an ID that is between about 0.2 m and about 1 m (e.g., between about 0.25 m and about 0.76 m). Indeed, in some embodiments, the ID (and/or other one or more other measurements of the reactor core tube) is adjusted or otherwise set to meet the needs of a particular fuel, application, and/or a desired energy output. The walls of the reactor core tube 120 can be any suitable thickness. Indeed, in some embodiments, the distance between the tube's outer diameter (OD) and ID (or wall thickness) is between about 0.1 cm and about 1 m, or any thickness that falls in such range. Indeed, in some embodiments, the tube has a wall thickness that falls between about 1 cm and about 13 cm (e.g., between about 1.5 cm and about 3.5 cm). In other embodiments, the tube's wall can be any other suitable thickness (e.g., based on energy output needs). With reference now to the end caps, although some embodiments of the reactor core 70 are formed with one or both ends (e.g., ends 140 and/or 145) being closed, in some embodiments, the first and/or second ends of the reactor core tube 120 are capped with an end cap. While the end caps can perform any suitable function, in some embodiments, the end caps are configured to help direct the fuel into and out of the reactor core tube. While the end caps can comprise any suitable component that allows them to perform their desired function, FIG. 4A shows a representative embodiment in which the first end cap 140 and second end cap 145 each comprises one or more (e.g., 1, 2, 3, 4, 5, 6, or more) fuel ports 150. Additionally, while the end caps can be any suitable shape, FIG. 4A shows an embodiment in which the first 140 and second 145 end caps are flared to respectively help channel fuel from the port 150 in the first cap 140 to the internal space 135 of the reactor core tube 120, and then from the internal space 135 of the reactor core tube 120 to the port 150 in the second cap 145. Where the reactor core 70 comprises a first 140 and/or second 145 end cap (or fuel heads), the end caps can be coupled to the core through any suitable method. Some example of such methods include, without limitation, being integrally formed with, being threaded together with, via a pressure and/or friction fitting, via one or more mating surfaces (e.g., grooves and corresponding ridges or otherwise), via a luer-taper connection, via one or more seals (e.g., carbon seals, carbon rope seals, rubber seals, and/or other suitable seals), via welding, via one or more adhesives, via one or more mechanical fasteners (e.g., rivets, clamps, clamping mechanisms, reflectors 75 and/or other objects that help press the caps into the reactor core tube 120, screws, bolts, clips, pegs, crimps, pins, brads, threads, brackets, catches, couplers, key-way splines, and/or any other suitable mechanical fasteners), and/or other suitable fastening mechanism. Indeed, in some embodiments, the end caps are coupled to the reactor core tube via a friction fitting, with one or more seals (e.g., carbon ropes and/or other suitable seals) being disposed between the end caps and the reactor core tube to help maintain an air-tight and/or fluid-tight seal between the caps and the reactor core tube. As mentioned, in some embodiments, the internal space 135 in the reactor core tube 120 comprises one or more internal moderators that are configured to help the fuel reach (and/or maintain) a critical state in the reactor core 70. In this regard, the internal moderators can comprise any suitable component or components that are capable of performing the described function. Some examples of suitable internal moderators include, but are not limited to, one or more rods, balls, pellets, beads, granules, particles, blocks, articles, pipes, graphite gels, gels, pieces, and/or other objects that can be surrounded by and/or filled with the fuel so as to allow the material of the moderator (e.g., carbon, graphite, and/or any other suitable material capable of bringing the cracked fuel to a critical state) to function as a moderator. Indeed, in some embodiments, the internal moderators comprise graphite balls, and more particularly a substantially pure graphite having a purity level of about 99% or greater (e.g., having a graphite purity of at least about 99.9%). In some other examples, the internal moderators comprise one or more cylinders, blocks, wedges, pins, rods, balls, solid block inserts defining a plurality of holes, the reactor core 70 itself (e.g., wherein the internal space 135 comprises one or more fuel channels or holes extending through a portion of the reactor core), and/or other suitable objects that define one or more holes therein, wherein such holes are configured to channel the fuel from a first portion (e.g., a first end 125 portion, a first diffuser (as discussed below), and/or a first end cap 140) to a second portion (e.g., a second end 130 portion, a second diffuser (as discussed below), and/or a second end cap 145) of the reactor core. Indeed, in some embodiments, the reactor core itself acts as the internal moderator. In some other embodiments, however, the internal moderators comprise one or more fuel pin rods, fuel wedges, and/or graphite spheres. Where the reactor core 70 itself acts as the internal moderator, the reactor core can comprise any suitable characteristic that allows it to bring and/or maintain the fuel at a critical state. In some embodiments, the core comprises (e.g., by itself and/or houses) a solid block of material (e.g., graphite, as discussed below) defining one or more fuel channels. In this regard, the core can comprise any suitable number of fuel channels, including, without limitation, between about 1 fuel channel and about 1,000 fuel channels, or any number of channels falling within such range. Indeed, in some embodiments, the reactor core defines between about 3 and about 80 (e.g., between about 3 and about 60) fuel channels. By way of non-limiting illustration, FIGS. 4B and 4C respectively show some embodiments in which the reactor core 70 itself defines 9 and 37 fuel channels 155. Additionally, FIG. 4D illustrates an embodiment in which the reactor core 70 comprises a cylindrical insert 156 that is disposed within the reactor core tube 120, and which defines 9 fuel channels 155. Where the reactor core 70 comprises one or more fuel pin rods, the fuel pin rods can comprise any suitable component or characteristic that allows them to bring a portion of the molten fuel to (or to be maintained at) a critical state. Indeed, while the pins can be any suitable length, in some embodiments, they are of a sufficient length that allows them to direct fuel from the first end 125 to the second end 130 of the reactor core tube 120. Additionally, in some embodiments, the pins define one or more holes, or fuel channels, that extend through a length of the pins to channel the fuel from the reactor core tube's first end 125 to its second end 130. The channels can be disposed in the pins in any suitable manner, including, without limitation, by running substantially parallel with a longitudinal axis running through a length of the pins, by cork-screwing through the pins, by extending through the pins at an angle, by rotating though the pins, by spiraling through the pins, and/or in any other suitable manner. In accordance with some embodiments, however, FIGS. 4A-4E show that the fuel channels 155 (which may also be referred to as internal fuel conduits and holes) run substantially straight through the pins 160 (e.g., parallel with the pins' longitudinal axes). Where the reactor core 70 comprises one or more pins 160, the pins can each define any suitable number of holes that allow the core to bring and/or maintain the fuel at a critical state. In this regard, each pin can comprise 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, or more holes. By way of non-limiting illustration FIG. 4F shows an embodiment in which several pins 160 comprise four fuel channels 155, while a center pin 165 comprises eight fuel channels 155. The pins 160 can be any suitable shape, including, without limitation, being substantially cylindrical; tubular; cuboidal; rectangular-prism-shaped; triangular-prism-shaped; polygonal-prism-shaped; pill-shaped (e.g., cylindrical with rounded ends); having an outer perimeter with a cross-sectional appearance resembling that of a peanut, cells in anaphase, cells in telophase, and/or a double-barreled shotgun; having a cross-sectional view resembling 2, 3, 4, 5, 6, or more intersecting circles; having or more corresponding shapes that fit together to substantially fill a portion of the reactor core 70; and/or any other suitable shape. By way of non-limiting illustration, FIGS. 4A and 4F show some embodiments in which the pins 160 have a cylindrical shape and/or (in the case of the center pin 165 shown in FIGS. 4A and 4F) a cross-sectional view resembling cells in telophase. Where the reactor core 70 comprises, one or more pins 160, the reactor core can comprise any suitable number of pins that allows the reactor core to function as described herein. In this regard, while some embodiments of the core comprise no pins, other embodiments comprise between about 1 and about 1,000 pins, or any subrange thereof. Indeed, in some embodiments, the reactor core comprises between about 1 and about 80 pins, or any subrange thereof (e.g., between about 12 and about 50 pins). By way of non-limiting illustration, FIG. 4F shows an embodiment in which the reactor core 70 comprises a total of 15 pins (as shown by pins 160 and 165). Although, in some embodiments, the internal space 135 is mostly (if not entirely) filled with fuel pin rods 160, in other embodiments, in addition to (or in place of) the pins, the internal space houses one or more wedges. In this regard, the term wedge may be used to describe any suitable internal moderator having a surface that is configured to substantially contour with an inner surface of the reactor core 70 (e.g., an inner surface of the reactor core tube 120) and/or to come into contact with such inner surface at more than one place. For instance, in some embodiments in which the reactor core tube 120 defines an interior surface having a polygonal, rounded, contoured, and/or irregular surface, an outer surface of one or more fuel wedges is configured to substantially contour such interior surface and/or to at least contact such surface in more than one location at a time. In this regard, FIG. 4F shows an embodiment in which the reactor core tube 120 defines a cylindrical interior surface 170, and in which an outer surface 175 of each of the fuel wedges 180 is curved and configured to substantially correspond in shape with the interior surface 170 of the reactor core tube 120. The fuel wedges 180 can have any suitable shape that allows the reactor 20 to function as intended. Some non-limiting examples of suitable shapes include that of geometrical sector-shaped prism, an arc-shaped prism, a polygonal prism, a rounded prism, and/or any other suitable shape. In accordance with some embodiments, however, FIG. 4F (and FIG. 4A) illustrates an embodiment in which the fuel wedges 180 comprise a substantially wedge-shaped prism 181, having a plurality of rounded surfaces 185 that are configured to hold one or more pins (e.g., pins 160 and/or 165). FIG. 4G illustrates an embodiment in which the reactor core 70 comprises multiple substantially-sector-shaped wedges 182, having a pin 160 disposed between the wedges. In particular, while the reactor core 70 can comprise any suitable number of wedges (i.e., 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, or more) FIG. 4G shows an embodiment in which the core 70 comprises four wedges 180. Additionally, FIG. 4H illustrates an embodiment in which the wedges 180 are substantially sector-shaped, and wherein there are no pins disposed within the reactor core 70. Furthermore, FIG. 4I illustrates an embodiment in which the reactor core 70 comprises a plurality of arc-shaped prism wedges 190 surrounding a plurality of arc-shaped prism internal moderators 195 and a fuel pin 160. Where the reactor core 70 itself, an insert in the core (e.g., the cylindrical insert 156), the pins 160, the wedges 180, and/or one or more other internal moderators each comprise one or more fuel channels 155 that are configured to direct fuel from a first portion (e.g., a first end 125 portion, a first diffuser (as discussed below), and/or a first end cap 140) to a second portion (e.g., a second end 130 portion, a second diffuser (as discussed below), and/or a second end cap 145) of the reactor core 70, the channels can be any suitable size that allows the fuel to flow through the channels. In some embodiments, the holes have an ID that is between about 0.05 cm and about 60 cm, or any ID that falls in such range (e.g., between about 0.5 cm and about 4 cm). Indeed, in some embodiments, the holes in the pins have an ID between about 0.9 cm and about 30.5 cm. In other embodiments, the fuel channels have an ID between about 0.95 cm and about 23 cm. By way of non-limiting illustration, FIG. 4B illustrates an embodiment in which the reactor core 70 defines fuel channels 155 that have an ID of about 0.95 cm (±0.9 cm). FIG. 4B illustrates an embodiment in which the reactor core 70 defines fuel channels 155 of two different sizes, which have an ID of between about 20 cm (±2 cm) and about 12 cm (±2 cm). FIG. 4C, on the other hand, illustrates an embodiment in which the reactor core 70 defines fuel channels 155 having an ID of about 7.6 cm (±2 cm). Although in some embodiments, the internal moderator or moderators (e.g., the fuel pins 160, fuel wedges 180, cylindrical insert 156, and/or other suitable moderators) are configured to substantially fill the reactor core 70 when the core is cool, in some embodiments, internal moderators are sized so as to be slightly smaller than the internal space 135 of the reactor core tube 120—thus allowing the internal moderators to expand (as they are heated) to substantially fill the internal space without expanding so much that they crack or break the reactor core tube. While the internal moderators can be any suitable size at standard temperature and pressure (or STP) that allows the reactor 20 to function as intended, in some embodiments, the volume (and/or length) of all of the internal moderators is configured to be between about 0.01% and about 15%, or any subrange thereof, smaller than the internal volume (and/or diameter or length) of the reactor core tube 120 at STP. Indeed, in some embodiments, the internal moderators (as a whole) have a total volume (and/or diameter or length) that is anywhere between about 1% and about 10% (e.g., between about 2.5% and about 5.5%) smaller than the internal volume (and/or diameter or length) of the reactor core tube at STP. The ends of the internal moderators (e.g., the reactor core 7 itself, the cylindrical insert 156, the fuel pins 160, and/or fuel wedges 180) can have any suitable shape that allows them to be used in the reactor core 70. Indeed, in some embodiments, the ends of the pins, wedges, inserts, etc. are substantially flat; are rounded; include one or more walls, spacers, protuberances, and/or other standoffs that are configured to space openings to the various fuel channels 155 away from an object (e.g., an end cap 140 or 145, or a diffuser, as discussed below); and/or are otherwise shaped to allow the fuel to enter into one end of, and to exit from an opposite end of, the various moderators. By way of non-limiting illustration, FIGS. 4J-4K illustrate some embodiments in which the pins 160 and 165 and the wedges 180 each comprise one or more standoffs 200 that are configured to space openings for the fuel channels 155 away from an object (e.g., a diffuser 205, the first end cap 140, the second end cap 145, and/or any other suitable object). Where one or more of the internal moderators (e.g., the fuel pins 160, fuel wedges 180, etc.) comprise one or more standoffs, the standoffs can be any suitable length. Indeed, in some embodiments, the standoffs at a first end or second end of the fuel pins, and/or fuel wedges are, individually, any suitable length between about 0.01 cm and about 20 cm, or any subrange thereof. Indeed, in some embodiments, the standoffs at one or both ends of the pins and/or wedges are, at each end, between about 1 cm and about 5 cm. In still other embodiments, the standoffs at one or both ends of the pins and/or wedges are, individually, between about 2 cm and about 4 cm (e.g., about 3.8 cm±0.5 cm). In still other embodiments, the standoffs are any other suitable length (e.g., based on energy output needs, fuel flow needs, and/or any other suitable factor). With reference now to the diffusers 205, the reactor core 70 can comprise any suitable baffle, channels, meshes, tubing, blocks, and/or other diffusers that are capable of distributing fuel from the first end cap 140 into the fuel channels 155 in the pins 160 and/or wedges 180, and/or from the fuel channels in the pins and/or wedges and into the second end cap 145. More particularly, the diffuser can comprise any suitable component (e.g., a manifold connected, fuel lines, holes, flutes, and/or any other suitable characteristic) that allows the diffuser to direct fuel to one or more portions of the reactor core (or internal moderator). In accordance with some embodiments, FIG. 4J shows the diffuser 205 comprises a plate 210 with one or more holes 215, with the plate being disposed between the fuel port 150 of the corresponding end cap (e.g., end caps 140 and/or 145) and the pins 160 and/or wedges 180. Additionally, FIG. 4A shows an embodiment in which the diffusers 205 are formed with the end caps (e.g., end caps 140 and/or 145). In accordance with some other embodiments, however, (and as shown in FIG. 4J) the diffusers 205 are formed separate from the end caps (e.g., end caps 140 and/or 145) so as to be inserted into one of the end caps, sandwiched between an end cap and a portion of the reactor core 70, and/or to be placed in any other suitable location. Where the reactor core 70 comprises one or more diffusers 205 defining a plurality of holes (see holes 215 in FIG. 4J), any suitable portion of the diffusers' surface area define holes that are configured to channel fuel. Indeed, in some embodiments, the area of the holes in a face of each diffuser is between about 50% and about 150% (or falls in any suitable subrange thereof) of the area of the fuel channels 155 in a face of the reactor core and/or the internal moderator. Indeed, in some embodiments, the area of the holes in a face of each diffuser is about equal (±10%) to the area of the fuel channels in a face of the reactor core and/or the internal moderator. Turning now to the fuel inlets 80 fuel outlets 85, the reactor 20 can comprise any suitable number of fuel inlets and outlets (e.g., 1, 2, 3, 4, 5, 6, or more) that allows fuel to selectively pass through one or more fuel ingress ports 151 (or inlets) at a first end of the reactor (e.g., the first end cap 140) and to then exit through one or more fuel egress ports 152 (or outlets) at a second end of the reactor (e.g., the first end cap 145). In one non-limiting illustration, however, FIG. 2B shows an embodiment in which the reactor 20 comprises one fuel inlet 80 and one fuel outlet 85. Additionally, while the fuel inlets can be any suitable shape (e.g., circular, polygonal, and/or any other suitable shape), in some embodiments, an egress from the fuel inlet and ingress to the fuel outlet substantially correspond with a shape of a corresponding fuel port 150. Indeed, in some embodiments, in which the fuel ports are substantially circular in shape, the egress from the fuel inlet and the ingress to the fuel outlet are also substantially circular in shape. While the fuel inlets 80 and fuel outlets 85 can be made of any suitable materials (e.g., graphite, one or more nickel alloys, low-chromium nickel-molybdenum alloys (such as HASTELLOY-N™), metals, cements, ceramics, synthetic materials, composites, and/or any other suitable materials), in some embodiments, the fuel inlet and outlet each comprise a low-chromium nickel-molybdenum alloy (e.g., HASTELLOY-N™ materials), with one or more seals (e.g., carbon seals, carbon rope seals, composites, and/or other suitable seals) being disposed between the inlet and outlet and the corresponding end cap (e.g., the first 140 or second 145 end cap) to which they extend. Indeed, in some embodiments, the fuel inlets 80 and outlets 85 comprise a HASTELLOY-N™ material that is lined with graphite. With reference now to the reactor control mechanism 90, some embodiments of the described system 10 are configured to selectively modify the rate at which fuel flows through the reactor core 120. In this regard, in some cases and within some limits, as fuel is forced through the reactor core at higher and higher rates, the fuel is able to interact with the internal moderators to allow the fuel to reach higher and higher temperatures. Conversely, in some cases and within some limits, as the rate at which fuel flows through the reactor core is slowed, the temperature of the fuel also drops. Indeed, in some embodiments, if the fuel is allowed to stay stagnant in the reactor core for an extended period of time, the fuel will lose its critical state and will (if left long enough) even harden. Thus, by varying the rate at which fuel moves through the reactor core, the described system can vary the amount of heat (and hence the amount of electricity) that the system produces. Moreover, by stopping the flow of fuel through the core, the system can be permanently and/or temporarily shut down. Where the described system 10 comprises one or more mechanisms for varying the rate at which fuel flows through the reactor core 20, the reactor control mechanisms 90 can comprise any suitable component or mechanism that is capable of performing such a function. In this regard, some non-limiting examples of suitable reactor control mechanisms include one or more variable frequency fuel pumps, valves, mechanisms in which the reactor core is rotatable so as to move the fuel ports 150 and the corresponding fuel inlet 80 and outlet 85 into and out of alignment with each other, and/or any other suitable mechanism. Indeed, in at least some embodiments, the reactor core is configured to be rotated to increase and/or decrease the rate at which fuel passes through the reactor 20. Where the reactor core 70 is configured to rotate to vary the rate at which fuel passes through the reactor, the reactor core can be rotated in any suitable manner that allows a passage between the fuel inlet 80 and/or outlet 85 and a corresponding fuel port 150 (e.g., in the first 140 and/or second 145 end cap) to become more and/or less occluded as the reactor core rotates. Indeed, in some embodiments, the reactor core is configured to be rotated manually, via one or more motors, servos, actuators, gear drives, worm drives, kelley drives, and/or other suitable mechanisms. In this regard, FIGS. 2A, 4L, and 4M show some embodiments in which the reactor core 70 is coupled with a partial gear 220 (or a sector gear) that is intermeshed with a second gear 225 that is sealed within the housing 65 and that comprises a pinion, gear, and/or other contact surface 230, which can be used to turn the second gear (e.g., via a wrench, pry bar, motor, servo, pneumatic driver, kelley shaft, and/or other suitable mechanism) to rotate the reactor core, to thereby vary the rate at which fuel is moved through the reactor and, hence, the amount of energy that is produced by the system 10. Additionally, FIGS. 4M and 6A-6D show that in some embodiments, at least one reflector 75 (e.g., the second reflector 240, as discussed below) is configured to allow the partial gear 220 and, hence, the reactor core 70 to rotate clockwise and counterclockwise. While this ability to rotate the reactor core in two directions may serve many purposes, in some embodiments, it allows the reactor core to move back and forth to break any fuel that has solidified and become crusted between the core and a reflector. Turning now to the reflectors 75, some embodiments of the described reactor 20 comprise one or more reflectors that are configured to reflect neutrons and/or gamma rays released from the fuel as the fuel moves through the reactor core 70. As a result, the reflectors may help the reactor bring and/or maintain the fuel at a critical state, while (in some embodiments) preventing radiation from escaping from the reactor 20 and harming individuals in proximity to the reactor. In this regard, the reflectors can comprise any suitable characteristic that allows them to function as intended. In one example of a suitable characteristic of the reflectors 75, the reflectors can be any suitable thickness that allows them to function as described herein. Indeed, in at least some embodiments, the reflectors ensure that an outer surface of the reactor core tube 120 and/or either of the end caps 140 or 145 is separated from an internal wall of the housing 65 by between about 2 cm and about 100 cm (or any subrange thereof) of a suitable material (e.g., graphite, as discussed below). Indeed, in some embodiments, the reflectors ensure that an outer surface of the reactor core tube 120 and/or either of the end caps 140 or 145 is separated from an internal wall of the housing 65 by between about 20 cm and about 600 cm (e.g., about 40 cm±10 cm) of reflector material. More specifically, in some embodiments, the reflectors ensure that an outer surface of the reactor core tube and/or either of the end caps is separated from an internal wall of the housing by at least about 30 cm. As another example of a suitable characteristic of the reflectors 75, although some embodiments of the reactor core 70 are permanently enveloped in a reflector, in other embodiments, the reactor core is surrounded in the reactor housing 65 by one or more reflectors that are configured to be selectively removed and replaced. As a result, in some embodiments, if the reactor core, an internal moderator, a reflector, and/or another portion of the reactor 20 breaks, cracks, ages, and/or otherwise becomes damaged, one or more reflectors can be removed such that the damaged portion of the reactor can be removed, accessed, repaired, and/or replaced. In this regard, while the reflectors can be assembled in any suitable manner that allows them to surround the reactor core, FIGS. 4M, 5A-7C, and FIGS. 2A-2B show that, in some embodiments, the reflectors 75 comprise a first 235 and second 240 reflector that are configured to fit together to encase the reactor core 70 (e.g., as a clam shell), with a third 245 and fourth 250 reflector that each flank the first end cap 140 and the second end cap 145. Accordingly, in such embodiments, one or more reflectors can be removed and/or replaced relatively easily. The various components of the reactor core 70 (including, without limitation, the reactor core itself, the reactor core tube 120, the first 140 and second 145 end caps, the cylindrical insert 156, the fuel pins 160, the fuel wedges 180, the diffusers 205, the reflectors 75, and/or the partial gear 220) can be made of any suitable materials. Some non-limiting examples of such materials include, but are not limited to, graphite (e.g., a substantially pure graphite having a purity level of about 99% or greater (such as a graphite purity of at least about 99.9%), a boron-free graphite, a pyrolytic graphite, a CGB grade graphite, and/or any other suitable graphite), and/or any other suitable material. Indeed, in some embodiments, the reactor core, the reactor core tube, the end caps, the cylindrical insert, the fuel pins, the fuel wedges, the diffusers, the reflectors, and/or the partial gear each comprise a 99.9% pure, boron-free graphite. In some other embodiments, one or more portions of the reactor core comprise one or more other metals, cements, ceramics, graphite spheres, and/or other suitable materials. For instance, some embodiments of the partial gear comprise a metal (e.g., HASTELLOY-N™ alloy) that is placed on and/or used to form teeth on the gear. Turning now to the drains 95, some embodiments of the reactor 20 comprise one or more drains that are configured to drain (e.g., into a suitable holding tank) fuel that seeps from the reactor core 70, and/or that is released when (or if) the reactor core cracks and/or breaks. While such drains can comprise any suitable component that allows them to function as intended, in some embodiments, the drains comprise one or more ball valves, butterfly valves, gate valves, diaphragm valves, and/or other suitable valves comprising one or more suitable ceramic materials, metals, alloys, composites, and/or other suitable materials. Indeed, in some embodiments, the drain 95 (as shown in FIGS. 1B-2B) comprises a ceramic ball valve. With reference now to the heat exchanger 25, in some embodiments of the described system 10, fuel that is brought to the critical state in the reactor core 70 is pumped (or otherwise moved) through the first fluid line 40 (which can be any suitable size and length), from the reactor 20, through the heat exchanger 25, and then back into the reactor for reheating. In some such embodiments, the heat exchanger is configured in such a manner that heat from fuel in the first fluid line is passed to a heat transfer medium running through the second fluid line (which can also be any suitable size and length). Accordingly, the described system can heat the heat transfer medium without ever contaminating it with radioactive materials from the fuel. While the transfer of heat from the first line 40 to the second line 45 can be done in any suitable manner, in some embodiments, the first fluid line is disposed in proximity to the second fluid line (e.g., as shown in FIGS. 8A-8E). Additionally, in some embodiments, in order to better pass heat from the first fluid line to the second fluid line, both lines are at least partially submerged in and/or are otherwise surrounded by the heat transfer medium. Moreover, while the first and second fluid lines can run through the heat exchanger 25 in any suitable manner (by having one run in a top portion of the heat exchanger while the other line runs in the bottom portion, by having portions of the lines disposed in close proximity to each other, etc.), in some embodiments, a portion of the first fluid line is configured to be disposed in a bottom portion of the heat exchanger while a portion of the second fluid line is configured to be disposed in an upper portion of the heat exchanger. With regards to the heat transfer medium, the heat transfer medium can comprise any suitable material or materials that allow it to safely absorb heat from the first fluid line 40 and, in some embodiments, to flow through the second fluid line 45. Some non-limiting examples of suitable heat transfer mediums include one or more salts that are free from fissionable materials, water, coolants, graphite gels, and/or other suitable materials. Indeed, in some embodiments, the heat transfer medium comprises one or more salts, which may include, but are not limited to, potassium nitrate; sodium nitrate; lithium fluoride; beryllium fluoride; a mixture of lithium fluoride and beryllium fluoride; a metal salt mixture of lithium fluoride, sodium fluoride, and potassium fluoride; a thermal graphite gel; and/or any other suitable salt or salts. Indeed, in some embodiments, the heat transfer medium comprises potassium nitrate and/or sodium nitrate. In some other embodiments, the carrier medium comprises potassium fluoride, sodium fluoride, and/or a graphite gel. Where the heat transfer medium comprises more than one ingredient, the various ingredients can be present at any suitable concentration in the fuel. Indeed, in some embodiments, the two components are respectively used at a molar ratio between about 100:1 and 1:100, or at any suitable subrange thereof. In this regard, in some embodiments, the carrier medium respectively comprises potassium nitrate and sodium nitrate at a molar ratio between about 6:1 and about 0.5:1 (e.g., at a ratio of about 1.5:1). In other embodiments, however, the carrier medium comprises potassium nitrate and sodium nitrate at any molar ratio that is suitable for a desired energy output, thermal fluid, system, and/or other suitable factor. The first 40 and second 45 fluid lines can be made of any suitable materials (e.g., one or more nickel alloys, low-chromium nickel-molybdenum alloys (such as HASTELLOY-N™), metals, cements, ceramics, synthetic materials, composites, and/or any other suitable materials) that allow the lines to function as intended. In some embodiments, however, the lines each comprise a low-chromium nickel-molybdenum alloy. In addition to the aforementioned characteristics, the heat exchanger 25 can comprise any other suitable component, including, without limitation, a housing (e.g., a housing comprising one or more of the materials and components similar to those discussed above with respect to the reactor 20), one or more drains (e.g., drains comprising one or more of the materials and characteristics similar to those discussed above with respect to the drain 95), one or more baffles and/or supports, mixers (e.g., as discussed above with respect to the heater 15), pumps, seals (e.g., as discussed above with respect to the reactor), and/or other suitable components. By way of non-limiting illustration, FIGS. 8A-8E show some embodiments in which the heat exchanger 25 comprises one or more supports 255 with openings 260, drain 256, housings 265, and seals 270. With reference now to the steam generator 30, in some optional embodiments, once the fuel (which has been brought to a critical state by passing through the reactor core 70) heats the heat transfer medium in the second fluid line 45 of the heat exchanger 25, the heated heat transfer medium is circulated (e.g., via one or more pumps or otherwise) in the second line from the heat exchanger to the steam generator, and then back to the heat exchanger. In some such embodiments, the second line (and/or an object heated thereby) is brought into contact and/or close proximity with water, such that heat from the heat transfer medium in the second line is able to convert the water to steam, which can then be used to turn a turbine connected to an electric generator 35 (which may include any suitable turbine and/or generator). In addition to the aforementioned components, the steam generator 30 can comprise any other suitable component that allows it to function as intended. Indeed, in some embodiments, the steam generator comprises a housing (e.g., a housing comprising one or more of the materials and components similar to those discussed above with respect to the reactor 20), one or more drains (e.g., drains comprising one or more of the materials and characteristics similar to those discussed above with respect to the emergency drain 95), one or more baffles and/or supports, mixers (e.g., as discussed above with respect to the heater 15), pumps, seals (e.g., as discussed above with respect to the reactor), water inlets, steam outlets, and/or other suitable components. By way of non-limiting illustration, FIGS. 9A-9E show some embodiments in which the steam generator 30 comprises one or more supports 280 with openings 285, drain 290, housings 295, seals 300, water inlets 305, and steam outlets 310. The various portions of the described system 10 can be made in any suitable manner. In this regard, some non-limiting examples of methods for making the described reactor core 70 include boring, machining, etching, cutting, drilling, grinding, shaping, plaining, molding, extruding, sanding, lathing, smoothing, buffing, polishing, and/or otherwise forming various pieces of graphite (and/or another suitable material) to form one or more pieces of the reactor core (e.g., the reactor core tube 120, end caps 140 and 145, fuel pins, 160, fuel wedges 180, diffusers, 205, reflectors 75, and/or other suitable parts). Furthermore, the other portions of the described system can be formed in any suitable manner, including, without limitation, via cutting; bending; tapping; dying; sanding; plaining; shaping; molding; extruding; drilling; grinding; buffing; polishing; connecting various pieces with one or more adhesives, mechanical fasteners (e.g., nails, clamps, rivets, staples, clips, pegs, crimps, pins, brads, threads, brackets, etc.), welds, and/or by melting pieces together; and/or any other suitable method that allows the described system to perform its intended functions. In addition to the aforementioned features, the described system 10 can be modified in any suitable manner that allows the system to generate heat and/or electricity. In one example, the various components of the described system can be coupled together in any suitable manner (e.g., via the first fluid line 40, the second fluid line 45, one or more connectors, ball valves, valves, and/or in any other suitable manner). By way of non-limiting illustration, FIG. 1B shows an embodiment in which the reactor 20 is coupled to the heat generator 25, which (in turn) is coupled to the steam generator, via one or more connection points 315 (e.g., lugs, recesses, mechanical fasteners, hammer pin rocks, catches, etc.) and connectors 320 (e.g., brackets, catches, braces, couplers, ball connections, joints, etc.). In another example, one or more components of the described system 10 are coupled to a common object. In this regard, some examples of such objects include, but are not limited to, a trailer (e.g., for a truck), a skid, a platform, a pallet, a train car, a vehicle (e.g., a train, car, truck, tractor, boat, ship, submarine, submergible, airplane, hovercraft, trolley, tank, motorcycle, bus, transports, heavy machinery, machinery, motor home, van, helicopter, military vehicle, space shuttle, drone, UAV, etc.); and/or any other suitable object. In another example, some embodiments of the reactor core 70 comprise one or more fuel pins 160 having rounded ends with a fuel channel 155 running between the two ends. In such embodiments, the pins can have any suitable characteristics that allows the reactor core to bring the fuel to (or to maintain the fuel at) a critical state. Indeed, in some embodiments, the rounded ends comprise one or more threads or other connection mechanisms configured to attach the rounded ends to the pin. The rounded ends of the pins 160 can further comprise any suitable number of holes, of any suitable size, that are configured to direct fuel into (and/or out of) the fuel channel(s) running in the pin. Indeed, in some embodiments, each of the rounded ends comprise 1, 2, 3, 4, 5, or more openings. Moreover, while the openings in the rounded ends of the pin can extend in any suitable manner, in some embodiments, the openings are disposed at an angle that directs fuel from the openings to (and/or from) the fuel channel in the pin. Furthermore, in some embodiments, a cross-sectional area of all of the openings in a rounded end of a pin are between about 80% and about 120% (or any subrange thereof) of a cross-sectional area of the fuel channel 155 in the pin. In one non-limiting illustration, FIG. 9F shows an embodiment in which a fuel pin 160 comprises two rounded ends 161 defining at least one opening 162, with a fuel channel 155 running through the pin. In another example, instead of being configured to generate steam, which is then used to generate heat, in some embodiments, the heat exchanger 25 and/or the second fluid line 45 are configured to heat and expand air. In turn, such expanded air can be used to turn a turbine (or otherwise actuate another suitable device) and generate electricity. In yet another example, instead of generating steam, the heat exchanger 25 and/or the second fluid line 45 are used to heat any other suitable object and/or medium. Indeed, in some embodiments, the heat exchanger and/or second fluid line are used to heat: a body of water (e.g., for distillation, desalination, evaporation, aquaculture, and/or any other suitable purpose), a building, a stadium, a neighborhood, an area, air, a complex, an underground reservoir containing fossil fuels, a heat transfer fluid, tar sands, oil shale, a biofuel waste water treatment plant, and/or any other suitable object and/or material. In still another example, instead of having the heat exchanger 25 and the steam generator 30 comprise two discrete components that are disposed next to each other, in some embodiments, one is contained (at least partially) within the other. Indeed, in some embodiments, at least a portion of the heat exchanger is disposed within the steam generator. In another example of a manner in which the described system 10 can be modified, in some embodiments, the rate at which fuel is passed through the reactor core 70 is controlled by a computer processor (e.g., as discussed below in the Representative Operating Environment system). Accordingly, in some embodiments, a computer (e.g., a special-purpose computer that is configured to regulate the reactor and/or a general purpose computer configured to perform the same function) is configured to increase the flow of fuel through the reactor core when more energy is needed (e.g., during peak hours of electrical consumption), to slow the flow of fuel through the reactor core when less energy is needed (e.g., during off-peak hours), and/or to shut down the reactor 20 when desired (e.g., in case of an emergency, maintenance, etc.). In yet another example, some embodiments of the described reactor 20 comprise one or more bearings and/or low friction surfaces that help allow for the reactor core 70 to rotate with respect to one or more reflectors 75. In even another example of a suitable modification, some embodiments of the described system 10 comprise one or more condensers that are configured to recycle some or all of the steam produced by the steam generator 30. In still another example of a suitable modification, some embodiments of the described system 10 are configured to extract one or more materials (e.g., chemicals, composition, mixtures, gases, and/or other desired materials) from the fuel as it cycles through the system. Indeed, in some embodiments (as illustrated by FIG. 10) the described system 10 comprises a processing center 325 that is configured to remove isotopes (e.g., medical grade isotopes) and/or other materials that are generated as the fuel is cycled. In another example, the described system 10 can comprise any other suitable component, including, without limitation, a secondary containment structure; a tertiary containment structure; a radiator configured to dissipate heat from the reactor core and/or fuel; one or more dump tanks configured to receive the fuel and/or heat transfer medium; one or more additional reactors 20 used in parallel, series, and/or any other suitable manner with the first reactor core 70; one or more emergency programs that are configured to automatically slow and/or stop the flow of fuel through the reactor core; one or more other components and/or programs that are configured to shut in and/or to dump the fuel from the reactor core 20; and/or any other suitable component. In addition to the aforementioned features, the described system 10 can comprise any other suitable feature. Indeed, some embodiments of the described reactor core 70 are configured to be used in any orientation, including, without limitation, in a horizontal, vertical, diagonal, and/or variable orientation. Indeed, unlike some reactors, some embodiments of the described reactor core are configured to be used in a horizontal orientation (e.g., as shown in FIG. 1C). Additionally, in some embodiments, the reactor core is configured to function as its orientation is changed (e.g., from vertical orientation, to diagonal orientation, and/or to a vertical orientation. Accordingly, some embodiments of the described reactor core are well suited for submarines, aircraft, and/other moving objects which may slightly or significantly vary the orientation of the reactor core. As another example of a feature of the described system 10, some embodiments of the system are configured to drain out some or all of the fuel in the reactor core 70 to shut down the reactor 20. Indeed, in some embodiments, the system is configured to allow a significant portion of the fuel to be drained from the reactor core (e.g., via the fuel outlet 85) such that the remaining fuel in the reactor cools down and solidifies. In some such embodiments, the reactor can be restarted by cracking the fuel (e.g., via the heater 15), introducing the cracked fuel into the reactor, and then recirculating the cracked fuel until the solidified fuel in the core is heated and brought to a critical state. As still another example, unlike some nuclear power plants that require a relatively large amount of real estate, some embodiments of the described system 10 have a relatively small footprint. Indeed, as discussed above, some embodiments of the described system can fit on a trailer, a train car, and/or in a variety of other locations that are relatively small. In still another example, some embodiments of the described system 10 are configured to actually use or “burn” nuclear waste from other nuclear reactors. As a result, in some embodiments, the described systems are quite beneficial for the environment and relatively inexpensive to operate. In still another example of a feature of the described system 10, in some embodiments, as the various components of the fuel are mixed, such components become polluted from their pure state—thus making them relatively undesirable to terrorists or others who may seek to create weapons from such materials. In still another example, some embodiments of the described system 10 are configured to produce relatively small amounts of plutonium in comparison to other nuclear power plants. In yet other examples of features associated with the described system, the reactor 20, in some embodiments of the described system, is configured to be air cooled, and to thus require rather small amounts of water when compared with some conventional nuclear power reactors. In even another example of a feature, some embodiments of the described system 10 comprise a reactor core 70 that has an internal space 135 that is relatively full with internal moderators. In this regard, some such embodiments leave relatively little room for gas (e.g., hydrogen, and/or other gases) to build-up in the reactor core 70. As a result, in some embodiments, some gases are prevented from forming and/or some gases are readily purged from reactor core, thus reducing the chances of unwanted chemical reactions and/or explosions. In yet another example, some embodiments of the described system are readily made mobile, thus making them ideal for power generation in locations with relatively little infrastructure (e.g., at oil drilling sites, offshore oil drilling platforms, off-planet locations, the theater of war, etc.). As mentioned, some embodiments of the described system 10 are configured to be operated (at least in part) by one or more special-purpose computers (e.g., computers configured to control the reactor core 70) and/or general purpose computers. Indeed, the described systems and methods can be used with or in any suitable operating environment and/or software. In this regard, FIG. 11 and the corresponding discussion are intended to provide a general description of a suitable operating environment in accordance with some embodiments of the described systems and methods. As will be further discussed below, some embodiments embrace the use of one or more processing (including, without limitation, micro-processing) units in a variety of customizable enterprise configurations, including in a networked configuration, which may also include any suitable cloud-based service, such as a platform as a service or software as a service. Some embodiments of the described systems and methods embrace one or more computer readable media, wherein each medium may be configured to include or includes thereon data or computer executable instructions for manipulating data. The computer executable instructions include data structures, objects, programs, routines, or other program modules that may be accessed by one or more processors, such as one associated with a general-purpose processing unit capable of performing various different functions or one associated with a special-purpose processing unit capable of performing a limited number of functions. Computer executable instructions cause the one or more processors of the enterprise to perform a particular function or group of functions and are examples of program code means for implementing steps for methods of processing. Furthermore, a particular sequence of the executable instructions provides an example of corresponding acts that may be used to implement such steps. Examples of computer readable media (including non-transitory computer readable media) include random-access memory (“RAM”), read-only memory (“ROM”), programmable read-only memory (“PROM”), erasable programmable read-only memory (“EPROM”), electrically erasable programmable read-only memory (“EEPROM”), compact disk read-only memory (“CD-ROM”), or any other device or component that is capable of providing data or executable instructions that may be accessed by a processing unit. With reference to FIG. 11, a representative system includes computer device 400 (e.g., a digital ratings device or other unit), which may be a general-purpose or (in accordance with some presently preferred embodiments) special-purpose computer. For example, computer device 400 may be a personal computer, a notebook computer, a PDA or other hand-held device, a workstation, a digital pen, a digital ratings device, a digital ratings device dock, a digital ratings device controller, a minicomputer, a mainframe, a supercomputer, a multi-processor system, a network computer, a processor-based consumer device, a cellular phone, a tablet computer, a smart phone, a feature phone, a smart appliance or device, a control system, or the like. Computer device 400 includes system bus 405, which may be configured to connect various components thereof and enables data to be exchanged between two or more components. System bus 405 may include one of a variety of bus structures including a memory bus or memory controller, a peripheral bus, or a local bus that uses any of a variety of bus architectures. Typical components connected by system bus 405 include processing system 410 and memory 420. Other components may include one or more mass storage device interfaces 430, input interfaces 440, output interfaces 450, and/or network interfaces 460, each of which will be discussed below. Processing system 410 includes one or more processors, such as a central processor and optionally one or more other processors designed to perform a particular function or task. It is typically processing system 410 that executes the instructions provided on computer readable media, such as on the memory 420, a magnetic hard disk, a removable magnetic disk, a magnetic cassette, an optical disk, or from a communication connection, which may also be viewed as a computer readable medium. Memory 420 includes one or more computer readable media (including, without limitation, non-transitory computer readable media) that may be configured to include or includes thereon data or instructions for manipulating data, and may be accessed by processing system 410 through system bus 405. Memory 420 may include, for example, ROM 422, used to permanently store information, and/or RAM 424, used to temporarily store information. ROM 422 may include a basic input/output system (“BIOS”) having one or more routines that are used to establish communication, such as during start-up of computer device 400. RAM 424 may include one or more program modules, such as one or more operating systems, application programs, and/or program data. One or more mass storage device interfaces 430 may be used to connect one or more mass storage devices 432 to the system bus 405. The mass storage devices 432 may be incorporated into or may be peripheral to the computer device 400 and allow the computer device 400 to retain large amounts of data. Optionally, one or more of the mass storage devices 432 may be removable from computer device 400. Examples of mass storage devices include hard disk drives, magnetic disk drives, tape drives, solid state mass storage, and optical disk drives. Examples of solid state mass storage include flash cards and memory sticks. A mass storage device 432 may read from and/or write to a magnetic hard disk, a removable magnetic disk, a magnetic cassette, an optical disk, or another computer readable medium. Mass storage devices 432 and their corresponding computer readable media provide nonvolatile storage of data and/or executable instructions that may include one or more program modules, such as an operating system, one or more application programs, other program modules, or program data. Such executable instructions are examples of program code means for implementing steps for methods disclosed herein. One or more input interfaces 440 may be employed to enable a user to enter data (e.g., initial information) and/or instructions to computer device 400 through one or more corresponding input devices 442. Examples of such input devices include a keyboard and/or alternate input devices, such as a digital camera, a sensor, bar code scanner, debit/credit card reader, signature and/or writing capture device, pin pad, touch screen, mouse, trackball, light pen, stylus, or other pointing device, a microphone, a joystick, a game pad, a scanner, a camcorder, and/or other input devices. Similarly, examples of input interfaces 440 that may be used to connect the input devices 442 to the system bus 405 include a serial port, a parallel port, a game port, a universal serial bus (“USB”), a firewire (IEEE 1394), a wireless receiver, a video adapter, an audio adapter, a parallel port, a wireless transmitter, or another interface. One or more output interfaces 450 may be employed to connect one or more corresponding output devices 452 to system bus 405. Examples of output devices include a monitor or display screen, a speaker, a wireless transmitter, a printer, and the like. A particular output device 452 may be integrated with or peripheral to computer device 400. Examples of output interfaces include a video adapter, an audio adapter, a parallel port, and the like. One or more network interfaces 460 enable computer device 400 to exchange information with one or more local or remote computer devices, illustrated as computer devices 462, via a network 464 that may include one or more hardwired and/or wireless links Examples of the network interfaces include a network adapter for connection to a local area network (“LAN”) or a modem, a wireless link, or another adapter for connection to a wide area network (“WAN”), such as the Internet. The network interface 460 may be incorporated with or be peripheral to computer device 400. In a networked system, accessible program modules or portions thereof may be stored in a remote memory storage device. Furthermore, in a networked system computer device 400 may participate in a distributed computing environment, where functions or tasks are performed by a plurality networked computer devices. While those skilled in the art will appreciate that the described systems and methods may be practiced in networked computing environments with many types of computer system configurations, FIG. 12 represents an embodiment of a portion of the described systems in a networked environment that includes clients (465, 470, 475, etc.) connected to a server 485 via a network 460. While FIG. 12 illustrates an embodiment that includes 3 clients connected to the network, alternative embodiments include at least one client connected to a network or many clients connected to a network. Moreover, embodiments in accordance with the described systems and methods also include a multitude of clients throughout the world connected to a network, where the network is a wide area network, such as the Internet. Accordingly, in some embodiments, the described systems and methods can allow for remote monitoring, observation, adjusting, and other controlling of one or more of the described systems 10 from many places throughout the world. Thus, as discussed herein, embodiments of the present invention embrace molten salt reactors. More particularly, some implementations of the described invention relate to systems and methods for providing a thorium molten salt reactor. In this regard, some implementations of the reactor are configured to rotate a reactor core to vary a flow rate of fissionable fuel through the reactor. Moreover, in some implementations, the reactor core houses two or more fuel wedges that each define at least one fuel channel that extends through the wedges. In some implementations, one or more of the wedges, components of the reactor core, and/or reflectors surrounding the core are configured to be replaced relatively easily. The present invention may be embodied in other specific forms without departing from its spirit or essential characteristics. The described embodiments, examples, and illustrations are to be considered in all respects only as illustrative and not restrictive. The scope of the invention is, therefore, indicated by the appended claims rather than by the foregoing description. All changes that come within the meaning and range of equivalency of the claims are to be embraced within their scope. In addition, as the terms on, disposed on, attached to, connected to, coupled to, etc. are used herein, one object (e.g., a material, element, structure, member, etc.) can be on, disposed on, attached to, connected to, or coupled to another object—regardless of whether the one object is directly on, attached, connected, or coupled to the other object, or whether there are one or more intervening objects between the one object and the other object. Also, directions (e.g., front back, on top of, below, above, top, bottom, side, up, down, under, over, upper, lower, lateral, etc.), if provided, are relative and provided solely by way of example and for ease of illustration and discussion and not by way of limitation. Where reference is made to a list of elements (e.g., elements a, b, c), such reference is intended to include any one of the listed elements by itself, any combination of less than all of the listed elements, and/or a combination of all of the listed elements. Furthermore, as used herein, the terms a, an, and one may each be interchangeable with the terms at least one and one or more.
	claims	1. A left radial access, right room operation peripheral intervention system for use with an imaging system, the peripheral intervention system comprising:a left radial base being configured to stabilize a left arm of a patient across a midsagittal plane of the patient during a lower extremity peripheral intervention on a procedure table;a right radial base disposed substantially parallel to an operating surface of the procedure table, the right radial base being transradiant and configured to position a right arm of the patient in a direction away from the midsagittal plane during the lower extremity peripheral intervention; anda radiation reduction barrier disposed substantially perpendicular to the right radial base and spaced apart from the left radial base, the radiation reduction barrier having a radiodense material disposed between the patient and an attending staff member to reduce scatter radiation from the patient in a direction of the staff member during a procedure. 2. The left radial access, right room operation peripheral intervention system as in claim 1, wherein the left radial base includes a support surface for contacting the left arm. 3. The left radial access, right room operation peripheral intervention system as in claim 2, further comprising a pad depending from the support surface. 4. The left radial access, right room operation peripheral intervention system as in claim 1, further comprising a base board being configured to receive the right radial base, the left radial base and the radiation reduction barrier. 5. The left radial access, right room operation peripheral intervention system as in claim 1, further comprising at least one radiodense apron in releasable connection proximate the procedure table. 6. The left radial access, right room operation peripheral intervention system as in claim 5, wherein the radiodense apron includes a plurality of pockets and respective radiodense inserts configured for insertion in the pockets. 7. The left radial access, right room operation peripheral intervention system as in claim 5, wherein the radiodense apron and the radiation reduction barrier are disposed in a plane substantially parallel to each other. 8. The left radial access, right room operation peripheral intervention system as in claim 7, wherein scatter radiation reduction exposure to an attending physician is reduced by at least sixty percent over a system without the radiodense apron and the radiation reduction barrier. 9. The left radial access, right room operation peripheral intervention system as in claim 7, wherein scatter radiation reduction exposure to attending staff is reduced by between about thirty percent to about seventy-seven percent over a system omitting the radiodense apron and the radiation reduction barrier. 10. A left radial access, right room operation peripheral intervention system for use with an imaging system, the peripheral intervention system comprising:a base board being configured for connection proximate a table having a left side and a right side corresponding to a left arm and a right arm of a patient;a left radial base attached to the base board and being configured to cushion and stabilize a left arm of a cardiac patient across a midsagittal plane of the patient during a lower extremity peripheral intervention on the table, wherein an attending cardiologist may perform the intervention from the right side of the table; anda radiation reduction barrier arranged vertically and spaced apart from the left radial base, the radiation reduction barrier having a radiodense material disposed between the patient and an attending staff member to reduce radiation scattering from the patient in a direction of the staff member. 11. The left radial access, right room operation peripheral intervention system as in claim 10, further comprising a right radial base disposed substantially parallel to an operating surface of the table, the right radial base being transradiant and configured to position a right arm of the patient in a direction away from the midsagittal plane during the lower extremity peripheral intervention. 12. The left radial access, right room operation peripheral intervention system as in claim 11, wherein the right radial base is attached to the base board. 13. The left radial access, right room operation peripheral intervention system as in claim 10, further comprising at least one radiodense apron releasably connected to the base board. 14. A left radial access, right room operation peripheral intervention system for use with an imaging system, the peripheral intervention system comprising:a base board being configured for attachment proximate a procedure table;a left radial base attachable to the base board and being configured to stabilize a left arm of a patient across a midsagittal plane of the patient during a lower extremity peripheral intervention on the procedure table;a right radial base attachable to the base board and disposed substantially parallel to an operating surface of the procedure table, the right radial base being transradiant and configured to position a right arm of the patient in a direction away from the midsagittal plane during the lower extremity peripheral intervention;a radiation reduction barrier attachable to the base board and spaced apart from the left radial base and from the right radial base, the radiation reduction barrier having a radiodense material disposed substantially vertically between the patient and an attending staff member to reduce radiation scattering from the patient in a direction of the staff member during an imaging procedure; anda radiodense apron releasably connected to the base board. 15. The left radial access, right room operation peripheral intervention system as in claim 14, wherein the radiation reduction barrier includes radiation attenuating material being configured for extension. 16. The left radial access, right room operation peripheral intervention system as in claim 14, wherein the radiodense apron includes a plurality of sleeves and respective radiodense material inserts. 17. The left radial access, right room operation peripheral intervention system as in claim 14, further comprising at least two radiodense aprons. 18. A left radial access, right room operation peripheral intervention system for use with an imaging system, the peripheral intervention system comprising:a right radial base having a base board attachable to a procedure table with a right side and a left side corresponding to a right arm and a left arm of a patient, the right radial base being disposed substantially parallel to an operating surface of the operating table, the right radial base board being disposed under the operating surface, the right radial base being transradiant and configured to position the right arm of the patient in a direction away from the midsagittal plane during the lower extremity peripheral intervention;a left radial base in connection with the right radial base board and being configured to stabilize a left arm of a patient across a midsagittal plane of the patient during a lower extremity peripheral intervention on the procedure table;a radiation reduction barrier attachable to a right radial base board under the table surface, the radiation reduction barrier spaced apart from the left radial base and from the right radial base, the radiation reduction barrier having a radiodense material disposed between the patient and an attending staff member to reduce radiation scattering from the patient in a direction of the staff member during an imaging procedure; anda radiodense apron releasably connected to the base board. 19. The left radial access, right room operation peripheral intervention system as in claim 18, wherein the radiodense apron includes at least two radiodense aprons, at least one apron movable relative to the other.
054426685	description	DETAILED DESCRIPTION OF THE INVENTION Referring now to FIGS. 1-4, one embodiment of a passive light water cooled and moderated pressure tube wet calandria type reactor 10 is shown. The reactor has an air baffle 11 surrounding a steel containment vessel 12 which defines an air path therebetween where air 18 enters from the environment and passes by the inlet and outlet plena of modular air heat exchangers 13 which have tilted horizontal water fin tubes 14 cooled by air. The reactor also has the primary inlet/outlet header 15 and the moderator inlet/outlet plenum 16 situated in moderator light water 17. The heat flow path for the reactor is shown in FIG. 1. wherein calandria tank 19 has the primary system inlet header 15a, a primary system outlet header 15b, moderator inlet plena 16a and a moderator outlet plena 16b. Between the primary system headers are pressure tubes 21 each having a fuel matrix 20 therein. High pressure and temperature primary cooling fluid is outlet at 23 from the primary system outlet headers 15b and low pressure and temperature moderator recirculating fluid passes through line 24 through heat exchanger 22 and back into primary system inlet header 16a. The heat exchanger 22 has a secondary fluid inlet 25 for ambient air and a secondary fluid outlet 26. In accordance with the invention, decay energy is stored in the fuel matrix through temperature redistribution. Heat is transported through the pressure/calandria tubes to the moderator system by conduction and radiation. The light water moderator absorbs the decay heat and circulates by natural circulation to provide an intermediate heat sink. The moderator is cooled within an air chimney which communicates with the ultimate heat sink, the environment. Thus this embodiment uses light water in both the coolant and moderator regions in place of heavy water. The use of light water as the moderator also thermally couples the fuel to the ultimate heat sink, allowing consideration of a variety of passive natural circulation configurations. A moderator recirculating system, together with a natural draft air cooling system, are used to ultimately remove the decay heat. The modular air heat exchangers 13 around the steel containment vessel are used with an outer air baffle to create a tall air chimney as shown in FIG. 2. Single phase water in the moderator recirculating loop is directed to the moderator outlet header and then distributed 360.degree. circumferentially to the inlet plenum of each modular air heat exchanger at elevated positions outside the steel containment vessel. The hot water is then directed from the inlet plenum to the outlet plenum through the tilted finned tubes and cooled by air and finally returned to the calandria by natural circulation. The system is always operating to remove the heat loss under normal conditions. FIGS. 3 and 4 show a fuel channel design according to the invention wherein a moderator water tube 31 is surrounded by void 30 and defines with calandria tube 33 a moderator water path 32. The calandria tube 33 and fuel runner 37 define a space between which axial interlocking fins 34 are disposed. Within the fuel runner 37 is a coolant gap 36 in which the fuel matrix 38 is disposed and which includes fuel rods 39 in coolant holes 40. The fuel rods 39 have contact lands 41 disposed in the coolant hole between the rods and the fuel matrix 38. In a preferred embodiment of the present invention, the fuel matrix is composed of 24 coolant holes with fuel rods in each hole within the matrix material. The cylindrical fuel bundle in a Zr-2.5% Nb pressure tube is supported by six runners 37 around the outer surface of the bundle. Conventional Zircaloy clad fuel with two weight per cent enrichment is supported in each coolant hole by contact lands 41, which in turn provide a good conduction path between the fuel and the matrix material. Graphite, with a silicon carbide coating to prevent water permeation, is employed as the matrix material. The interlocking fins 34 increase the thermal radiation heat transfer by increasing heat transfer area for both the pressure and calandria tubes. The axial fins are uniformly distributed peripherally and attached to the outer surface of the pressure tubes and the inner surface of the calandria tubes. An increase in temperature under accident conditions naturally enhances the radial heat transfer which in turn results in a thermal switch effect. The fin material must have good thermal conductivity such as that of an aluminum alloy. Where less conductive Zircaloy is used as the fin material, the insulating gap between the fins can be filled with an SiC coated graphite particle bed. The packed particle bed is then operated under vacuum for normal conditions and is filled with air or with steam under accident conditions to enhance the effective conductivity. The fin annular ring 32 of moderator water replaces the heavy water moderator of conventional designs. Decay heat can be transported radially across the fuel channel from the fuel matrix to the low temperature moderator water in the annular ring to protect the fuel and pressure tube from overheating under accident conditions. The moderator water annulus allows retention of on line refueling and reduces the heat loss from direct gamma and neutron heating occurred by a heavy water pressure tube reactor. FIG. 5 is a schematic cross-sectional view of a passive pressure tube light water cooled and moderated dry calandria type reactor 50 according to the present invention. As shown therein, a calandria 61 surrounded by a solid neutron reflector 60 is submerged in a tank 71 with light water 58 which serves for both cooling of the reflector and for the shielding of neutrons which escape from the reflector. The tank 71 with calandria 61 are submerged in a large amount of containment water 56 in a tank 55 embedded in concrete 54. During normal operation, containment water 56 is kept away from the calandria space by a gas lock 59. The gas pressure is maintained at such a level as to balance the water column outside the calandria. Any disturbance in the primary cooling system pressure, for example, loss of coolant or loss of heat sink, beyond a predefined envelope of safety limits, will result in the opening of fail safe passive valve 70 and the flooding of the entire calandria space by the containment water. The flooding water has the function of storing a large amount of decay energy in the form of latent heat, thus substantially reducing the instantaneous heat rate which must be transported through containment walls in the early stages of an accident, it removes the decay heat from the reactor's pressure tubes by boiling and condensing on containment walls, it shuts down the reactor and renders it deeply subcritical by excessive neutron absorption even in a boiling mode and it considerably reduces the decay heat load on the fuel matrix by absorbing a large portion of gamma heating which would have been otherwise deposited in the fuel matrix. The steel containment vessel 53 of reactor 50 is surrounded by a protective wall 51 made of prestressed concrete such that an annular air path is formed with an air intake 72 and an air outlet 52 at the top of the reactor. Within the reactor and disposed above the floor 62 is the calandria pressure control 65 which controls valves 67 and 68 which in turn control the input and output of gas to the calandria space in order to maintain the gas lock 59. The primary system inlet header 64 and outlet header 63 a part of the heat exchange path to steam generator 66 and heat exchanger 57 is disposed in the containment water as in conventional systems. In order to enable the fuel matrix in the pressure tubes in the calandria to store excess energy while the calandria is being flooded and enhance the heat removal rate from the fuel to the pressure tube boundary, improved designs are shown for the fuel matrix in FIGS. 6 and 7. As shown in FIG. 6, the calandria design 80 includes a calandria tube 81 with an inner pressure tube 82 forming a gap 83 therebetween filled with a high conductivity gas. Within the pressure tube 82 is fuel matrix 84 in the form of a solid cylindrical block and which forms a gap 85 between it and the pressure tube in which light water coolant flows. Within the fuel matrix 84 are a plurality of holes 86 filled with particle fuel in compacts. Also distributed between concentric layers of the particle fuel are holes 87 in which light water coolant flows. The design of FIG. 6 also includes a central hole 89 through which light water coolant flows. In an alternative embodiment of the design 80' shown in FIG. 7, pressure tube 81' forms a gap 83' for high conductivity gas with pressure tube 82'. The solid fuel matrix with dispersed coated particle fuel 84' forms a gap 85' with the pressure tube and in which the light water coolant flows. Within the solid cylindrical block 84' are a plurality of holes 87' coated with a protective coating 90 and in which light water coolant flows. The matrix materials are preferably materials with good thermal conductivity and high specific heat, relatively low neutron absorption cross-section, high resistance to radiation damage and good water and steam compatibility at high pressures. Graphite, silicon carbide and aluminum oxide are usable as matrix material. The fuel coated particle is similar to that employed in conventional designs and which can withstand high temperature. The kernels of the particles are uranium enriched to several per cent in the form of uranium oxide or preferably uranium carbide. The separated coolant and fuel holes are distributed homogeneously inside the matrix material and the designs shown in FIGS. 6 and 7 are representative examples of distribution patterns. It is to be appreciated by those of skill in the art that various arrangements of coolant and fuel holes may be placed within the envelope of a pressure tube and that the arrangement is not limited to the separated coolant and fuel holes. It is also noted that the embodiment of FIG. 7 wherein a homogenous dispersion of the coated particle fuel within the solid block equipped with homogeneously distributed coolant holes which are protected by a high temperature and wear-resistant coating and also have other arrangements within the scope of the present invention. The advantages of the fuel matrix design of the present invention is that it provides the feature of passive decay heat removal even in the absence of emergency coolant, it enables utilization of the high burn-up capabilities of particle fuel in a light water cooled reactor, it eliminates the problems of pellet cladding interaction existent with conventional pressure tube fuel, it enhances the prevention of fission gas released due to the multiple barriers of the fuel design such as the coating layers, matrix material and pressure tubes and the low density of the heavy metal per pressure tube combined with a dry calandria design increases prompt neutron lifetime and reduces greatly the potential concerns with reactor runaway. FIGS. 8 and 9 show in more detail the operation of the gas lock according to the present invention. As shown therein, the dry calandria design comprises the low pressure vessel 71 with the pressure tubes in a conventional arrangement. However, in the present invention, the heavy water moderator is replaced by gas under pressure which is slightly above atmospheric pressure to create a void space as shown in FIG. 8. The void space provides a gas lock for the containment water 56 which is in direct communication with the calandria space 61 and it increases the neutron migration length and lifetime and thus creates an environment for a core with a super flat profile of neutron thermal flux and hence power density. The gas used for the gas lock should be non-corrosive, have low neutron absorption cross-section and should be stable in neutron flux and in a high temperature environment. Helium and carbon dioxide are each useful for this function. As shown in FIGS. 8 and 9, the calandria bottom is equipped with large passages 93 with extended vertical walls submerged in the containment water 56. During normal operation, the containment water level is kept below the calandria bottom in the space within the extended vertical walls by maintaining the gas pressure in balance with the containment water column. The pressure level is maintained by the DC powered blower 69 and the control valve 68 shown in FIG. 5. The top of the calandria comprises several orifices 93 of appreciable size, hermetically closed by sealing members 91 during normal operation. The sealing member is connected to the piston 92 of a passive fail safe valve 70 which opens the sealing member due to disturbances in primary system pressure exceeding the envelope of safety limits, allowing rapid decrease of gas pressure and the flooding of the calandria. It is appreciated that other devices for effecting calandria flooding can also be used, for example, an electrically powered blower connected by a pipe to the calandria space and maintaining the gas pressure while operating and otherwise releasing the pressure would also suffice. Calandria flooding has the advantage of shutting down the reactor and rendering it deeply subcritical by excessive neutron absorption even in a boiling mode, providing a large amount of water which stores a large amount of decay energy in the form of latent heat and thus substantially reducing the instantaneous heat rate which must be transported through the containment walls in the early stages of an accident, insures the removal of decay heat from calandria tubes by boiling and condensing on containment walls and considerably reduces the decay heat load on the fuel matrix by absorbing a significant portion of the gamma heating which would have otherwise been deposited in the fuel matrix. The reflector 60 which surrounds the calandria space has the properties of having very good reflecting power for both fast and thermal neutrons, very low neutron absorption cross-section and good resistance towards irradiation damage. Preferable materials for the reflector are graphite and beryllium with graphite being more preferable due to its lower cost and lower radiation damage. In connection with this design, the calandria tube design shown in FIGS. 6 and 7 is particularly useful. The designs shown in FIGS. 6 and 7 have the primary function to protect the pressure tubes from excessive stresses which would have resulted had the cold water, during flooding, come into contact with a hot pressure tube still under pressure. The space between the calandria tube and the pressure tube is filled with high conductivity gas to maximize heat transport capabilities of the gap. The gas serves additionally for monitoring any leakage from the pressure tube. This can also be accomplished by use of duplex tubes. Reactor control is accomplished by control rods containing neutron absorbers, inserted preferably in the solid reflector on both faces of the dry calandria in the space between the columns of pressure tubes. This design eliminates the use of heavy water and substantially decreases heat losses during normal operations. The design also provides for an exceedingly flat profile of the neutron thermal flux and hence a flat power density profile in both axial and radial directions. This decreases the peaking factor by a factor of 2 compared with conventional thermal reactors. FIGS. 10 and 11 show one embodiment of the fail safe valve 70 of FIG. 5 which consists of a pilot valve shown in FIG. 10 and an air operated main valve shown in FIG. 11 which are preferably connected in one body (not shown). The pilot valve includes main piston 103, balancing piston 105, dead band positioning means including springs 108, 109, nuts 111, 112 and a nitrogen pressurized balancing tank 114. The air operated main valve comprises an air pressurized tank 139 of pressure P2 which is much lower than pressure P1 from the nitrogen tank 114, a loading piston 132, a release piston 133 on the same shaft 136 and sharing a common shaft 121 with that of the pilot valve, and an acting piston 137 connected to the shaft 92 which in turn is connected to the sealing member 91. The sealing member 91 is also biased upwardly by safety spring 94 shown in FIG. 11. The valve works on primary system pressure P1 which is input via inlet 102 to chamber 116 of main piston 103. The dead bands are set such that envelop all conceivable pressure disturbances and transients which do not require reactor shutdown. As shown in FIG. 10, the area 104a is connected to chamber 120 via fluid passages 118, 119 and nitrogen tank 114 is connected to area 120a on the other side of balancing piston 105. In the main valve, air tank 139 communicates with space 134a and space 134 via passage 151. Chamber 135 communicates with the atmosphere via passage 148 and space 135a communicates with the atmosphere via passage 149. Space 134 is also connectable to space 138a via passage 141 and 142 and space 134a communicates with space 138 via passage 144 and can communicate with space 138a via passage 141 and passage 143 when the load piston 132 is moved to the left. Space 138 communicates with space 135 through passages 145 and 146 when release piston 133 is moved to the right and communicates with space 135a via passages 145 and 147 when release piston 133 is moved to the left. In case of a low pressure accident, that is, loss of coolant, the primary system pressure decrease leads to the movement of pistons to the left due to the difference between the force exerted by pressure in the balancing tank 114 and the force from the primary system pressure at inlet 102. This movement is opposed by spring 108 and continues without any action until the distance traveled by the pistons 103 and 105 equals .DELTA.X.sub.2. At this point the lower pressure limit is reached, the loading piston 132 of the air operated valve switches the air passages such that the air pressure is released from the space 138 above the acting piston 137 and introduced below the acting piston and hence the acting piston moves upwardly and opens the closure element 91. In the case of a high pressure accident such as the loss of a heat sink, the primary system pressure acting on the main piston 103 moves the pistons 103 and 105 to the right until the upper limit proportional to force K.DELTA.X.sub.1, exerted by spring 109 is reached. At this point, the passage 117 around the main piston is opened allowing primary system pressure into area 104 and act on area S.sub.2. The resulting force to the right is increased since it becomes proportional to area S.sub.2 and moves the pistons to the right. This leads to reswitching of air passages and opening of the closure seal in the same manner as the case for low pressure discussed above. The safety spring 94 is intended for the liftup of the closure seal in case of a failure of the air tank 139, although the pressure within the calandria will also act to open the sealing member. As a result of the above-described design, all conceivable failures of the various valve parts lead to valve opening including a break in the pressure junction from the primary system, a break in the pressure junction from the nitrogen pressurized tank or break in the tank itself, a break in spring 108 or 109, a break of the air tank 139 or the low pressure junction to the air tank, failure of the safety spring 94 or all of the above failures combined. Table I attached is a listing of parameters for preferred embodiments of wet and dry calandria reactors according to the present invention as described hereinabove. The preceding specification describes, by way of illustration and not of limitation, preferred embodiments of the invention. Equivalent variations of the described embodiments will occur to those skilled in the art. Such variations, modifications, and equivalents are within the scope of the invention as recited with greater particularity in the following claims, when interpreted to obtain the benefits of all equivalents to which the invention is fairly entitled. TABLE I ______________________________________ PARAMETERS FOR BOTH WET AND DRY CALANDRIA REACTORS (EXCEPT WHERE NOTED) ______________________________________ Reactor System Core thermal Power MWth 4029 Moderator, Coolant H.sub.2 O, H.sub.2 O No. of fuel channels 740 Core/calandria diameter cm 877.32 Core/calandria length cm 594 Channel pitch cm 28.575 Void Coefficient Negative Calandria Tube Outside diameter mm 222.62 (148.62)* Inside diameter mm 219.58 (145.58)* Tube thickness mm 1.52 Tube material Zr 2 Gas gap thickness mm 40 (3)* Pressure Tube Outside diameter mm 139.58 Inside diameter mm 128 Tube thickness mm 5.79 Channel length mm 5940 Tube material Zr-2.5Nb Fuel Diameter of the matrix mm 125.6 No. of pins per bundle 24 Rod diameter mm 12.7 Clad thickness mm 0.419 (N.A.)* Pellet diameter mm 11.774 (N.A.)* No. of lands (per rod) 4 (N.A.)* Thickness of lands mm 1.0 (N.A.)* Enrichment % 2 (7)* Refueling on-load Burnup, MWd/MT - 18,800 (80,000)* Primary System Pressure (channel in.out) MPa 15.34, 14.0 (15.0, 14.3)* Channel .DELTA.P MPa 1.06 (0.7)* Temp. (channel in.out) C 298.9, 338.3 Channel .DELTA.T C 39.4 Outlet quality % 1.34 (0.5)* Mass flow rate/Core Mg/s 15.51 Moderator System Temp. (in.out) C 55.98 (N.A.)* ______________________________________ *Dry Calandria Reactor N.A. = Not Applicable
	claims	1. A method of measuring impedance change in a boiling water reactor (BWR), the method comprising:packing a high dielectric constant, non-linear mineral powder along a desired length of an electrical conducting electrode to provide an insulated coaxial type cable comprising a high dielectric constant, non-linear mineral powder insulator devoid of powder agglomeration and uniformly distributed along the desired length;bundling the desired length of high dielectric, non-linear mineral insulated coaxial cable together with a plurality of nuclear fuel rods to form a transmission line assembly having a characteristic impedance;inserting at least a portion of the transmission line assembly within the boiling water reactor core;backfilling an inert gas into the high dielectric constant, non-linear mineral powder insulator; andmeasuring a characteristic impedance change associated with the high dielectric constant, non-linear mineral powder insulated coaxial type cable in response to at least one of neutron or gamma irradiation generated via the reactor core. 2. The method according to claim 1, further comprising measuring a change in the ionization level of the inert gas in response to at least one of neutron or gamma irradiation generated via the reactor core. 3. The method according to claim 1, wherein providing a desired length of high dielectric constant, non-linear mineral powder insulated coaxial type cable comprises providing a high dielectric, non-linear mineral powder insulated coaxial type cable having a length equal to or greater than the length of the reactor core.
	claims	1. A method for automatic monitoring of at least one component of a physical system,where at least one sensor automatically monitors a characteristic of a component of a physical system and emits an electrical signal corresponding to a value of the characteristic via a measurement chain connected to a data reader of a computer system, which electrical signal is read by the data reader at a computer system, processed by the computer system and then output by the computer system to be saved as a data file in a data record of the computer system,where subsequently in a first checking step the computer system checks the data file of the data record for first errors caused by the preceding processing by the computer system,where subsequently in a second checking step the computer system checks the data file of the data record for second errors that result from problems with the at least one sensor and the measurement chain,where subsequently in a third checking step the computer system checks the data file of the data record for third errors that result from problems with the component, andwhere subsequently in a fourth checking step the first, second and third errors are checked against one another and known data for possible errors to isolate an error source and are either rejected if the isolated error source is not deemed sensible in light of the first, second and third errors or known data or output as an error message with reference to the isolated error source;wherein the computer system automatically takes corrective action to correct the isolated error source. 2. The method in accordance with claim 1, where after the second checking step in a further checking step, where the at least one sensor includes a plurality of equivalent sensors each measuring the characteristic of the component of the physical system, the data file of the data record is checked for comparative errors between the plurality of equivalent sensors, which comparative errors result from problems with at least one of the plurality of equivalent sensors. 3. The method in accordance with claim 1, wherein data corresponding to the first, second and third errors is automatically stored in the data record. 4. The method in accordance with claim 1, wherein the output of the error message includes at least one automatic generation of a proposal for error detection. 5. The method in accordance with claim 1, wherein the physical system is a pump system and the component is a pump of the pump system, and the at least one sensor includes at least one chosen from a rotational speed sensor sensing a rotational speed of the pump, an inlet pressure sensor sensing an inlet pressure of the pump, and an outlet pressure sensor sensing an outlet pressure of the pump. 6. The method in accordance with claim 5, wherein the at least one sensor includes each of a rotational speed sensor sensing a rotational speed of the pump, an inlet pressure sensor sensing an inlet pressure of the pump, and an outlet pressure sensor sensing an outlet pressure of the pump.
	abstract	A Thorium molten salt energy system is disclosed that includes a proton beam source for producing a proton beam, that can vary between a first energy level and a second energy level of, where the generated proton bean can be directed into a main assembly containing both Thorium-containing molten salt and Thorium fuel rods, each containing an inner Beryllium element and an outer solid Thorium element. The generated proton beam can be shaped and directed to impinge upon Lithium within the molten salt to promote the generation of thermal neutrons and the fission of Uranium within the molten salt. The generated proton beam can also be shaped and directed to impinge upon the Beryllium within the Thorium fuel rods to promote the generation of high energy neutrons.
048184721	abstract	A method and an apparatus for the wet dismantling of radioactively contaminated or activated components of nuclear reactor plants such as pressure vessels, includes optionally enclosing the component with a jacket or enclosure tube at the end of the service life of the component, the jacket or tube may be cast or molded, providing a sheathing between the component and the jacket, or between the component and a concrete shield, having a thickness sufficient to support at least part of the component after dismantling the component into individual pieces, or joining the jacket to the pressure vessel or to a bottom plate, flooding the component with water for radiation shielding, at least partly dismantling the component into individual pieces through a material-removing tooling or erosion method, and removing the individual pieces.
	abstract	An X-ray CT apparatus is capable of obtaining an image of arbitrary size for an arbitrary part of an object to be examined. The X-ray CT apparatus prepares two-dimensional data and three-dimensional data from a plurality of X-ray information obtained by driving an X-ray tube and an X-ray information detector through one rotation around a patient in a range between a lower jaw and the eyes of the patient, and it displays a tomographic image based thereon on a display unit. A rotational mechanism is fitted to a support, a U-shaped arm is mounted on the support with the X-ray tube and the X-ray information inputting means facing each other and fitted, and the image of an arbitrary size of an arbitrary part of the patient is collected and displayed as the patient is supported in a chair which is capable of being electrically-driven in the vertical direction.
051868881	claims	1. Device for recovering and cooling the core (5) of a nuclear reactor in meltdown, following an accident, said reactor comprising a generally cylindrical vessel (4) enclosing said core (5), said vessel being disposed with its axis vertical in a cylindrical reactor pit (3) having a lower bottom (3a) located vertically below said vessel (4), and said device being constituted by a metal structure (10) resting on said lower bottom (3a) and being submerged in a mass of water filling a lower portion of said reactor pit (3), said metal structure (10) comprising (a) a central chimney (11) comprising a cylindrical body (11a) disposed coaxially with said reactor pit (3) and a deflecting upper wall (15) inclined in relation to the horizontal plane, disposed above said cylindrical body (11a); (b) a wall (12, 42) for recovering and cooling said core, said wall being disposed around said body of said chimney (11) and being constituted by a first assembly of contiguous dihedral elements (22, 52) consisting of metal sheets and having a straight ridge (24), said elements being fixed radially around said body of said chimney (11) by inner end portions, in a region of triangular openings (16, 46) traversing said wall of said body of said chimney (11), said ridges (24) of said dihedra constituting upper portions of said dihedra and being inclined upwards in the direction of said chimney (11) and a space located between said wall (12, 42) and said bottom of said reactor pit (3) communicating with said chimney (11) via said triangular openings (16, 46); and (c) a peripheral wall (13, 43) fixed to external ends of said dihedra (22) of said first assembly, said external ends being opposite the ends fixed to said chimney (11), and said peripheral wall being disposed in the vicinity of an internal surface of said reactor pit (3), in such a manner as to provide at least one passage (30, 55) bringing the space located between said recovery wall (12) and said bottom (3a) of said reactor pit (3) into communication with the internal volume of said reactor pit (3), above said recovery wall (12). 2. Device according to claim 1, wherein said peripheral wall (13, 43) comprises a wall element (13, 44) of polygonal cross-section constituted by a second assembly of dihedral elements fixed to the external ends of said dihedra (22) of said first assembly and comprising straight ridges (26, 27) disposed vertically on the periphery of said reactor pit (3). 3. Device according to claim 2, wherein said peripheral wall further comprises a cylindrical hoop (45) fixed coaxially with said peripheral wall (44) along external vertical ridges of said peripheral wall, said hoop (45) having a diameter less than a diameter of said reactor pit, so as to provide on the periphery of said recovery and cooling device (40) an annular space (55) bringing the space located between said recovery wall (42) and said bottom (53a) of said reactor pit into communication with the internal volume of said reactor pit (53), above said recovery wall (42). 4. Device according to any one of claims 1 to 3, wherein said cylindrical body (11a) of said central chimney (11) has an upper portion comprising gaps (17) for passage of steam. 5. Device according to any one of claims 1 to 3, wherein said chimney (41) further comprises a cross-shaped central element (47), to which is fixed said deflecting upper wall (50), having a height in an axial direction greater than a height of said cylindrical body of said chimney (41). 6. Device according to any one of claims 1 to 3, wherein said dihedra (22) of said first assembly comprise rectangular inclined faces folded along a diagonal (23) in such a manner that the dihedra (22) have two different angles of opening in the radial direction.
041475887	abstract	Disclosure is made of a device for recharging a fast-neutron reactor, intended for the transfer of fuel assemblies and rods of the control and safety system, having profiled heads to be gripped on the outside. The device comprises storage drums whose compartments for rods of the control and safety system are identical to compartments for fuel assemblies. In order to store and transport rods of the control and safety system from the storage drums to the recharging mechanism provision is made for sleeve-type holders. When placed in such a holder, the dimensions of a rod of the control and safety system are equal to those of a fuel assembly. To join a holder to a rod of the control and safety system, on the open end of each holder there is mounted a collet, whereas on the surface of each rod of the control and safety system, close to its head, there is provided an encircling groove to interact with the collet. The grip of the recharging mechanism is provided with a stop interacting with the collet in order to open the latter and withdraw the safety and control system rod from its holder.
	claims	1. A power module assembly comprising:a containment vessel;a reactor vessel housed in the containment vessel, wherein the reactor vessel is configured to release coolant into a containment region located between the reactor vessel and the containment vessel, and wherein the containment vessel is configured to prohibit the released coolant from escaping out of the containment vessel; anda thermal insulation of the reactor vessel including the containment region located between the reactor vessel and the containment vessel, wherein the containment region comprises a partial vacuum that substantially surrounds the reactor vessel and is maintained at a below atmospheric pressure of less than 300 mmHG absolute. 2. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, wherein an entirety of the containment region is maintained at the below atmospheric pressure prior to the high-pressure event, and wherein substantially all gases are evacuated from the containment region during normal operation of the power module assembly. 3. The power module assembly of claim 2, wherein the evacuated gases comprise non-condensable gases, and wherein the containment region remains substantially evacuated of all non-condensable gases during an emergency operation comprising the high-pressure event, when the coolant is released into the containment region. 4. The power module assembly of claim 2, wherein the evacuated gases comprise oxygen, and wherein any hydrogen that is released together with the coolant is released at a level that maintains an oxygen-hydrogen mixture within the containment vessel at a non-combustible level without using a hydrogen recombiner. 5. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, and wherein the below atmospheric pressure prohibits substantially all convective heat transfer between the reactor vessel and the containment vessel during normal operation of the power module assembly prior to the high-pressure event. 6. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, and wherein the containment region remains substantially dry during normal operation of the power module assembly prior to the high-pressure event. 7. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, and wherein an outer surface of the reactor vessel is substantially surrounded by the below atmospheric pressure during normal operation of the power module assembly prior to the high-pressure event. 8. The power module assembly of claim 1, further comprising a flow limiter configured to controllably vent the released coolant as steam into the containment region during an emergency operation comprising a high-pressure event. 9. The power module assembly of claim 8, wherein a condensation of the steam on an interior wall of the containment vessel reduces pressure in the containment vessel at approximately the same rate that the vented steam adds pressure to the containment vessel. 10. The power module assembly of claim 8, wherein the steam is released into the containment vessel to remove a decay heat of a reactor core primarily through condensation of the steam on an inner surface of the containment vessel. 11. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, wherein an inner surface of the containment vessel substantially surrounds the reactor vessel, and wherein the inner surface of the containment vessel is substantially dry during normal operation of the power module assembly prior to the high-pressure event. 12. The power module assembly of claim 1, wherein the containment vessel comprises an inner wall, wherein the reactor vessel comprises an outer wall, and wherein the containment region is located between the inner wall and the outer wall. 13. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, and wherein all of the containment region is maintained at the below atmospheric pressure prior to the high-pressure event. 14. The power module assembly of claim 1, wherein the reactor vessel is configured to release coolant into the containment region in response to a high-pressure event, wherein the partial vacuum substantially surrounds the reactor vessel at a below atmospheric pressure of less than 300 mmHG absolute prior to the high-pressure event, and wherein the containment vessel prohibits the released coolant from escaping out of the containment vessel in response to the high-pressure event. 15. A power module assembly comprising:a containment vessel;a reactor vessel housed in the containment vessel, wherein the reactor vessel is configured to release coolant into a containment region, and wherein the containment vessel is configured to prohibit a release of the coolant out of the containment vessel; anda thermal insulation of the reactor vessel including the containment region located between the reactor vessel and the containment vessel, wherein the containment region comprises a partial vacuum that substantially surrounds the reactor vessel and is maintained at a below atmospheric pressure, and wherein the below atmospheric pressure prohibits substantially all convective heat transfer between the reactor vessel and the containment vessel. 16. The power module assembly of claim 15, wherein the below atmospheric pressure is less than 300 mmHG absolute. 17. The power module assembly of claim 15, wherein the containment vessel is surrounded by a heat sink. 18. The power module assembly of claim 17, wherein the heat sink comprises water or gas. 19. The power module assembly of claim 17, wherein the heat sink comprises rock, soil, or other solid material. 20. The power module assembly of claim 15, wherein the partial vacuum is maintained at the below atmospheric pressure, which is less than 300 mmHG absolute, prior to a high-pressure event and during normal operation of the power module assembly, and wherein the reactor vessel is configured to release coolant into the containment region during an emergency operation of the power module assembly comprising the high-pressure event. 21. The power module assembly of claim 20, wherein the containment vessel is configured to prohibit a release of the coolant out of the containment vessel during the emergency operation. 22. The power module assembly of claim 20, wherein the containment region comprises a first containment region and a second containment region, wherein the first containment region is maintained at atmospheric pressure, and wherein the second containment region is maintained at the below atmospheric pressure. 23. The power module assembly of claim 22, wherein the coolant is released as steam into the first containment region during the emergency operation, and wherein the steam condenses as liquid coolant in the containment vessel. 24. The power module assembly of claim 23, further comprising one or more valves connecting the first containment region to the second containment region, and wherein the one or more valves are operatively configured to transfer the liquid coolant from the first containment region to the second containment region. 25. The power module assembly of claim 23, wherein the reactor vessel is insulated by conventional thermal insulation in the first containment region, and wherein the reactor vessel is insulated by a reflective insulation in the second containment region. 26. A power module assembly, comprising:a thermal insulation including a containment region, wherein the containment region comprises a partial vacuum;means for releasing coolant into the containment region, wherein a reactor core is housed in the means for releasing; andcontainment means for maintaining the partial vacuum at less than 300 mmHG absolute, wherein the containment region is located between the means for releasing and the containment means, wherein the partial vacuum substantially surrounds the means for releasing, and wherein the containment means is configured to completely retain the released coolant within the containment means. 27. The power module assembly of claim 26, wherein the means for releasing coolant comprises a reactor vessel, and wherein the containment region is bordered by an inner wall of the containment means and an outer wall of the reactor vessel. 28. The power module assembly of claim 27, wherein the partial vacuum prohibits substantially all convective heat transfer between the outer wall of the reactor vessel and the inner wall of the containment vessel. 29. The power module assembly of claim 26, wherein the means for releasing comprises means for releasing the coolant into the containment region during an emergency operation of the power module assembly comprising a high-pressure event, and wherein the containment means comprises containment means for maintaining the partial vacuum at less than 300 mmHG absolute during a normal operation of the power module assembly prior to the high-pressure event. 30. The power module assembly of claim 26, wherein the means for releasing comprises means for releasing the coolant into the containment region during a high-pressure event, wherein the partial vacuum substantially surrounds the means for releasing prior to the high-pressure event, and wherein the containment means is configured to completely retain the released coolant within the containment means during the high-pressure event.
	claims	1. A control module arrangement for a compressed air maintenance unit, the control module arrangement being adapted to be assembled in stacked arrangement with several maintenance modules of the maintenance unit in a row one after the other, wherein the control module arrangement has a bus communications means for communicating with the maintenance modules via an internal bus of the maintenance unit, wherein the control module arrangement further comprises a diagnostic means for communicating with an input/output device of the maintenance unit, wherein the input/output device has a connection means for connecting at least one diagnostic sensor separate from the maintenance unit and/or for connecting at least one pneumatic compressed air influencing actuator separate from the maintenance unit, wherein the control module arrangement can control the one or more compressed air influencing actuators via the input/output device and/or the control module arrangement can receive diagnostic messages from the one or more separate diagnostic sensors, and wherein the diagnostic means are constructed for generating diagnostic data for a higher-order diagnostic device for diagnosing at least one function of the compressed air maintenance unit with reference to diagnostic messages of the one or more separate diagnostic sensors and/or with reference to diagnostic messages of the maintenance modules, so that the compressed air maintenance unit forms a diagnostic node, and wherein the control module arrangement further comprises first communications means for communicating with a higher-order control device for controlling and/or monitoring the compressed air maintenance unit and second communications means for communicating with the higher-order diagnostic device, to which the diagnostic data can be output, wherein the arrangement operates the first communications means before the second communications means in order of priority. 2. A control module arrangement according to claim 1, wherein the one or more separate diagnostic sensors and/or the one or more compressed air influencing actuators are allocated to a separate component, supplied with compressed air from the maintenance unit. 3. A control module arrangement according to claim 1, wherein the control module is constructed for controlling the maintenance modules and/or the one or more compressed air influencing actuators separated from the maintenance unit with reference to the diagnostic messages of the one or more separate diagnostic sensors and/or the maintenance modules. 4. A control module arrangement according to claim 1, wherein the one or more compressed air influencing actuators comprises a pneumatic valve and/or a pneumatic regulator. 5. A control module arrangement according to claim 1, wherein the one or more sensors comprises a magnetic sensor and/or an optical sensor and/or a pressure sensor and/or a temperature sensor and/or a flow rate measurement device and/or an electric measurement contact probe. 6. A control module arrangement according to claim 1, wherein the diagnostic means communicate with the input/output device via a proprietary connection and/or via the internal bus. 7. A control module arrangement according to claim 1, wherein the input/output device is constructed as a module of the compressed air maintenance unit. 8. A control module arrangement according to claim 1, wherein the diagnostic means link diagnostic messages of the one or more separate diagnostic sensors and/or the maintenance modules with reference to at least one logic condition to the diagnostic data. 9. A control module arrangement according to claim 1, wherein the diagnostic means perform data reduction for the formation of the diagnostic data. 10. A control module arrangement according to claim 1, wherein the diagnostic data have a format that can be viewed through an Internet browser. 11. A control module arrangement according to claim 1, wherein the connection means of the input/output device are constructed for connecting a fluid hose to at least one electrical or optical conductor to the one or more diagnostic sensors and/or to the one or more compressed air influencing actuators. 12. A control module arrangement according to claim 1, further comprising control means, especially a memory programmable control for controlling the maintenance modules. 13. A control module arrangement according to claim 1, wherein the first communications means are real-time capable. 14. A control module arrangement according to claim 1, wherein the second communications means have a reception block for predetermined control commands for controlling the compressed air maintenance unit. 15. A control module arrangement according to claim 1, wherein the first and/or the second communications means have a bus interface. 16. A control module arrangement according to claim 1, wherein the arrangement can be set by its transmission parameters for transmitting diagnostic data. 17. A control module arrangement according to claim 1, further comprising a control module, which contains the control means for controlling the maintenance modules, and a diagnostic module, which contains the diagnostic means. 18. A control module arrangement according to claim 1, further comprising an operating device interface for connecting a local operating device and/or visualization device. 19. A control module arrangement for a compressed air maintenance unit, the control module arrangement being adapted to be assembled in stacked arrangement with several maintenance modules of the maintenance unit in a row one after the other, wherein the control module arrangement has a bus communications means for communicating with the maintenance modules via an internal bus of the maintenance unit, wherein the control module arrangement further comprises a diagnostic means for communicating with an input/output device of the maintenance unit, wherein the input/output device has a connection means for connecting at least one diagnostic sensor separate from the maintenance unit and/or for connecting at least one pneumatic compressed air influencing actuator separate from the maintenance unit, wherein the control module arrangement can control the one or more compressed air influencing actuators via the input/output device and/or the control module arrangement can receive diagnostic messages from the one or more separate diagnostic sensors, and wherein the diagnostic means are constructed for generating diagnostic data for a higher-order diagnostic device for diagnosing at least one function of the compressed air maintenance unit with reference to diagnostic messages of the one or more separate diagnostic sensors and/or with reference to diagnostic messages of the maintenance modules, so that the compressed air maintenance unit forms a diagnostic node, and wherein the control module arrangement further comprises first communications means for communicating with a higher-order control device for controlling and/or monitoring the compressed air maintenance unit and second communications means for communicating with the higher-order diagnostic device, to which the diagnostic data can be output, wherein the arrangement transmits fewer diagnostic messages via the first communications means than via the second communications means. 20. A control module arrangement for a compressed air maintenance unit, the control module arrangement being adapted to be assembled in stacked arrangement with several maintenance modules of the maintenance unit in a row one after the other, wherein the control module arrangement has a bus communications means for communicating with the maintenance modules via an internal bus of the maintenance unit, wherein the control module arrangement further comprises a diagnostic means for communicating with an input/output device of the maintenance unit, wherein the input/output device has a connection means for connecting at least one diagnostic sensor separate from the maintenance unit and/or for connecting at least one pneumatic compressed air influencing actuator separate from the maintenance unit, wherein the control module arrangement can control the one or more compressed air influencing actuators via the input/output device and/or the control module arrangement can receive diagnostic messages from the one or more separate diagnostic sensors, and wherein the diagnostic means are constructed for generating diagnostic data for a higher-order diagnostic device for diagnosing at least one function of the compressed air maintenance unit with reference to diagnostic messages of the one or more separate diagnostic sensors and/or with reference to diagnostic messages of the maintenance modules, so that the compressed air maintenance unit forms a diagnostic node, and wherein the control module arrangement further comprises first communications means for communicating with a higher-order control device for controlling and/or monitoring the compressed air maintenance unit and second communications means for communicating with the higher-order diagnostic device, to which the diagnostic data can be output, wherein the second communications means have a greater transmission bandwidth than the first communications means.
	claims	1. A closed drift ion source configured to produce an ion beam, said closed drift ion source comprising:an ionizable gas source;a cathode emitting electrons;an anode;a magnetic field source comprising a first magnetic pole and a second magnetic pole configured to produce a magnetic field that passes through said anode, a closed confinement region and into a magnetic outer pole shunt surface, said electrons being confined along said magnetic field by a first electrically floating surface intermediate between said electrons and said anode. 2. A closed drift ion source in accordance with claim 1, wherein:an insulating layer extends substantially perpendicular to said magnetic field. 3. A closed drift ion source in accordance with claim 2, wherein:said insulating layer comprises a ceramic material. 4. A closed drift ion source in accordance with claim 1, wherein:said anode and at least said first magnet pole are formed as one integrated component. 5. A closed drift ion source in accordance with claim 1, comprising:a magnet disposed within said anode to provide said first magnetic pole. 6. A closed drift ion source in accordance with claim 1, wherein:said anode comprises magnetically permeable material. 7. A closed drift ion source in accordance with claim 1, wherein:said anode is of substantially circular cross section such that said closed drift ion source is configured as a round ion source. 8. A closed drift ion source in accordance with in claim 1, wherein:said anode is elongate such that said closed drift ion source is configured as a linear ion source. 9. A closed drift ion source in accordance with claim 8, wherein:said closed drift ion source has a length at least 300 mm long. 10. A closed drift ion source in accordance with claim 9, wherein:said closed drift ion source is configured such that a uniform ion beam is produced over a predetermined width of a substrate disposed in a predetermined relationship to said closed drift ion source. 11. A closed drift ion source in accordance with claim 8, wherein:said closed drift ion source has a length at least 1000 mm long. 12. A closed drift ion source in accordance with claim 1, comprising:coolant fluid passages in fluid communication with said anode to provide cooling of said anode. 13. A closed drift ion source configured to produce an ion beam, said closed drift ion source comprising:an electron source;a magnetic field source comprising a center magnetic pole with a first and second magnetic poles such that said magnetic field source is integrated and in direct contact with a body of an anode, said anode serving as both a center magnetic pole and an electrical anode, where primary magnetic field lines extend between said first and said second magnetic poles;said anode being disposed and configured such that said primary magnetic field lines pass through said anode; anda closed drift confinement region that forces electrons to cross said primary magnetic field lines to reach said anode. 14. A closed drift ion source in accordance with claim 13, comprising:an insulating layer or an insulating cover is disposed on said anode such that electrons are prevented from reaching said anode until after they have crossed said closed drift confinement region. 15. A closed drift ion source in accordance with claim 14, wherein:said insulating layer is present and extends substantially perpendicular to said primary magnetic field lines. 16. A closed drift ion source in accordance with claim 14, wherein:said insulating layer comprises a ceramic material. 17. A closed drift ion source in accordance with claim 14, comprising:a magnet disposed within said anode to provide said first magnetic pole. 18. A closed drift ion source in accordance with claim 14, wherein:said anode comprises magnetically permeable material. 19. A closed drift ion source in accordance with claim 13, wherein:said anode and at least said first magnet pole are formed as one integrated component. 20. A closed drift ion source in accordance with claim 13, wherein:said anode is of substantially circular cross section such that said closed drift ion source is configured as a round ion source. 21. A closed drift ion source in accordance with in claim 13, wherein:said anode is elongate such that said closed drive ion source is configured as a linear ion source. 22. A closed drift ion source in accordance with claim 21, wherein:said closed drift ion source has a length at least 300 mm long. 23. A closed drift ion source in accordance with claim 22, wherein:said closed drift ion source is configured such that a uniform ion beam is produced over a predetermined width of a substrate disposed in a predetermined relationship to said closed drift ion source. 24. A closed drift ion source in accordance with claim 21, wherein:said closed drift ion source has a length at least 1000 mm long. 25. A closed drift ion source in accordance with claim 13, comprising:coolant fluid passages in fluid communication with said anode to provide cooling of said anode. 26. A closed drift ion source in accordance with claim 13, wherein:said electron source comprises a hollow cathode. 27. A closed drift ion source in accordance with claim 13, wherein:said electron source comprises a sputter magnetron. 28. A closed drift ion source in accordance with claim 13, wherein:said electron source comprises a thermionic filament. 29. A method for providing a closed drift ion source configured to produce an ion beam, said method comprising:providing an ionizable gas source;providing a cathode emitting electrons;providing an anode;providing a magnetic field from a magnetic field source comprising a first magnetic pole and a second magnetic pole, said magnetic field configured to pass through said anode, a closed confinement region and into a magnetic outer pole shunt surface, said electrons being confined along said magnetic field by a first electrically floating surface intermediate between said electrons and said anode. 30. A method in accordance with claim 29, wherein:an insulating layer is present and disposed on said anode. 31. A method in accordance with claim 30, comprising:providing said insulating layer on said anode, such that said insulating layer extends substantially perpendicular to a set of primary magnetic field lines. 32. A method in accordance with claim 30, wherein:a ceramic material is utilized for said insulating layer. 33. A method in accordance with claim 30, comprising:disposing a magnet within said anode to provide said magnetic field source. 34. A method in accordance with claim 30, comprising:forming said anode from magnetically permeable material. 35. A method in accordance with claim 29, wherein:forming said anode and said magnet field source as one integrated component. 36. A method in accordance with claim 29, comprising:configuring said anode to be of substantially circular cross section such that said closed drift ion source is configured as a round ion source. 37. A method in accordance with in claim 29, comprising:configuring said anode to be elongate such that said closed drift ion source is configured as a linear ion source. 38. A method in accordance with claim 29, comprising:providing coolant fluid passages in fluid communication with said anode. 39. A method in accordance with claim 29, comprising:configuring said closed drift ion source such that a uniform ion beam is produced over a predetermined width of a substrate disposed in a predetermined relationship to said closed drift ion source.
	abstract	Fissionable materials are distinguished from other high-effective atomic number materials by producing dual-energy x-ray radiation sufficient to cause fission in fissionable materials and directing the dual-energy x-ray radiation sufficient to cause fission in fissionable materials towards a physical region. X-ray radiation and a product of fission from the physical region are sensed. An absorption of the dual-energy x-ray radiation by the physical region is determined based on the sensed x-ray radiation, and whether the physical region includes fissionable material is determined based on the presence of a product of fission.
	claims	1. An electron beam irradiating apparatus with monitoring device, comprising:an electron beam irradiation means irradiating materials in an irradiation chamber with an electron beam, the electron beam being generated by accelerating thermal electrons, the thermal electrons being emitted from a plurality of filaments;a photographing means capturing lights emitted by the irradiated materials;a storage means storing states of the electron beam irradiation in advance; anda calculating means processing an image captured by the photographing means to decide the state of electron beam irradiation stored in the storage means,wherein the storage means stores luminance of the images that correspond to the states of electron beam irradiation, and stores at least three states of electron beam irradiation selected from a group consisting of normal, axis deviation, broken filament, and vacuum window deterioration, andthe calculating means loads the image captured by the photographing means to compare the loaded image with the luminance of the image stored in the storage means, reads the states of electron beam irradiation related to the luminance of images stored in the storage means, and thereby decides the state of electron beam irradiation. 2. The electron beam irradiating apparatus with monitoring device according to claim 1,wherein the calculating means divides the image captured by the photographing means into a plurality of segments and compares the emitted light luminance of the each segment with the threshold value stored in the storage means. 3. An electron beam irradiating apparatus with monitoring device, comprising:an electron beam irradiation means irradiating materials in an irradiation chamber with an electron beam, the electron beam being generated by accelerating thermal electrons, the thermal electrons being emitted from a plurality of filaments;a photographing means capturing lights emitted by the irradiated materials;a storage means storing states of electron beam irradiation in advance; and a calculating means processing an image captured by the photographing means to decide the state of electron beam irradiation stored in the storage means,wherein the storage means storesa first threshold value that is set at the maximum value of the emitted light luminance when the electron beam is irradiated normally;a second threshold value that is set at the minimum value of the emitted light luminance when the electron beam is irradiated normally, and is set at higher value than the emitted light luminance when the electron beam is irradiated with axis deviation;a third threshold value that is set at lower than the second threshold value, is set at higher value than the emitted light luminance when the electron beam is irradiated with broken filament, and is set to the minimum value of the emitted light luminance when the electron beam is irradiated with axis deviation; andat least three states of electron beam irradiation selected from a group consisting of normal, axis deviation, broken filament, and vacuum window deterioration are stored, and the each state corresponds to state areas of the storage means that are divided by the three threshold values;the calculating means loads the value of the emitted light luminance of the image captured by the photographing means to compare the loaded luminance value with each of the threshold values stored in the storage means,reads the state of electron beam irradiation stored in the storage means when the loaded luminance value is equal to or higher than the second threshold value and equal to or lower than the first threshold value, and decides that the state of electron beam irradiation is normal,reads the state of electron beam irradiation stored in the storage means when the loaded luminance value is lower than the second threshold value and equal to or higher than the third threshold value, and decides that the state of electron beam irradiation is axis deviation, anddecides that the state of electron beam irradiation is broken filament among the states of electron beam irradiation stored in the storage means when the loaded luminance value is lower than the third threshold value. 4. The electron beam irradiating apparatus with monitoring device according to claim 3,wherein the storage means stores a first threshold value that is set at the maximum value of the emitted light luminance when the electron beam is irradiated normally, and is set at lower value than the emitted light luminance when the electron beam is irradiated with the state of vacuum window deterioration, and the storage means also stores the states of electron beam irradiation each of which represents normal, axis deviation, broken filament, and vacuum window deterioration; andthe calculating means reads the state of electron beam irradiation stored in the storage means when the value of the emitted light luminance of the image captured by the photographing means is higher than the first threshold value and decides that the state of electron beam irradiation is vacuum window deterioration. 5. The electron beam irradiating apparatus with monitoring device according to claim 4,wherein the electron beam irradiation means has a constant current controlled filament power supply and a voltmeter, the constant current controlled filament power supply being connected to a plurality of the filaments, the voltmeter measuring the filament voltage;the storage means stores a voltage setting that is higher than the filament voltage of the vacuum window deterioration and is equal to or lower than the filament voltage of the filament deterioration, and the storage means also stores the states of electron beam irradiation each of which represents normal, axis deviation, broken filament, vacuum window deterioration, and filament deterioration; andthe calculating means loads the filament voltage from the voltmeter when the value of the emitted light luminance of the image captured by the photographing means is higher than the first threshold value to compare with the voltage setting stored in the storage means and decides that the state of electron beam irradiation is the filament deterioration when the loaded filament voltage is equal to or higher than the voltage setting. 6. The electron beam irradiating apparatus with monitoring device according to claim 5,wherein the voltage setting stored in the storage means is set 1.1 times the initial filament voltage. 7. The electron beam irradiating apparatus with monitoring device according to claim 4,wherein the electron beam irradiation means has a constant voltage controlled filament power supply and an ammeter, the constant voltage controlled filament power supply being connected to a plurality of the filaments, the ammeter measuring the filament current;the storage means stores a current setting that is equal to or larger than the filament current of the filament deterioration and is smaller than the filament current of the axis deviation, and the storage means also stores the states of electron beam irradiation each of which represents normal, axis deviation, broken filament, vacuum window deterioration, and filament deterioration; andthe calculating means loads the filament current from the ammeter when the value of the emitted light luminance of the image captured by the photographing means is lower than the second threshold value and equal to or higher than the third threshold value to compare with the current setting stored in the storage means and decides that the state of electron beam irradiation is the filament deterioration when the loaded current is equal to or smaller than the current setting. 8. The electron beam irradiating apparatus with monitoring device according to claim 7,wherein the current setting stored in the storage means is set 0.9 times the initial filament current. 9. The electron beam irradiating apparatus with monitoring device according to claim 4,wherein the electron beam irradiation means has a constant current controlled filament power supply, a voltmeter, a grid, and a control means,the constant current controlled filament power supply being connected to a plurality of the filaments,the voltmeter measuring the filament voltage,the grid being connected to a grid power supply oppositely facing the filament, andthe control means controlling the amount of thermal electrons emitted from the filament by regulating the voltage of the grid power supply;the storage means stores a voltage setting that is higher than the filament voltage of the normal and is equal to or lower than the filament voltage of the filament deterioration, and the storage means also stores the states of electron beam irradiation each of which represents normal, axis deviation, broken filament, vacuum window deterioration, and filament deterioration; andthe calculating means loads the filament voltage from the voltmeter when the value of the emitted light luminance of the image captured by the photographing means is equal to or higher than the second threshold value and equal to or lower than the first threshold value to compare with the voltage setting stored in the storage means and decides that the state of electron beam irradiation is being filament deterioration when the loaded voltage is equal to or higher than the voltage setting. 10. The electron beam irradiating apparatus with monitoring device according to claim 4,wherein the electron beam irradiation means has a constant voltage controlled filament power supply, an ammeter, a grid, and a control means,the constant voltage controlled filament power supply being connected to a plurality of the filaments,the ammeter measuring the filament current,the grid being connected to a grid power supply oppositely facing the filament,the control means controlling the amount of thermal electrons emitted from the filament by regulating the voltage of the grid power supply;the storage means stores a current setting that is equal to or larger than the filament current of the filament deterioration and smaller than the filament current of the normal, and the storage means also stores the states of electron beam irradiation each of which represents normal, axis deviation, broken filament, vacuum window deterioration, and filament deterioration; andthe calculating means loads the filament current from the ammeter when the value of the emitted light luminance of the image captured by the photographing means is higher than the first threshold value to compare with the current setting stored in the storage means and decides that the state of electron beam irradiation is filament deterioration when the loaded current is equal to or smaller than the current setting.
	description	The present application claims priority to Korean Patent Application No. 10-2019-0144909, filed Nov. 13, 2019, the entire contents of which is incorporated herein for all purposes by this reference. The present invention relates to an overpower penalty calculation system for a generic thermal margin analysis model. More specifically, the present invention relates to a system calculating an overpower penalty for a thermal margin analysis model with a statistical method from a distribution formed of random numbers arbitrarily extracted within a range of each variable constituting the overpower penalty. When an initial core and a reload core of a Korean standard nuclear power plant are to be designed, three thermal margin analysis models, namely, reload core analysis model, generic analysis model, and on-line analysis model, are respectively used for evaluation of a thermal margin of a corresponding cycle's core. Among the three thermal margin analysis models, the generic thermal margin analysis model and the reload core thermal margin analysis model are compared with each other, whereby conservatism is verified. A conservatism comparison of the thermal margin analysis model is quantified in a form of an overpower penalty, and the overpower penalty is defined as a ratio of a power operating limit (POL) of the generic thermal margin analysis model and a POL of the reload core thermal margin analysis model, with respect to a same nucleate boiling ratio (DNBR) level. Overpower ⁢ ⁢ penalty = POL ⁢ ⁢ of ⁢ ⁢ reload ⁢ ⁢ core ⁢ ⁢ thermal ⁢ ⁢ margin ⁢ ⁢ analysis ⁢ ⁢ model POL ⁢ ⁢ of ⁢ ⁢ generic ⁢ ⁢ ⁢ thermal ⁢ ⁢ margin ⁢ ⁢ analysis ⁢ ⁢ model The overpower penalty is calculated by combining 4 types of axial power distributions with respect to COLSS Narrow 1˜8 and CPC Narrow 1, 3, 5, and 7 operating conditions and is calculated for each of approximately 140 conditions by combining 10 types of axial power distributions with respect to nominal and CPC wide 1˜8 operating conditions. Among the calculated overpower penalties, a minimum value is selected as a representative value of the overpower penalties of the generic thermal margin analysis model for a corresponding reload core. However, as a limited number of operating conditions and axial power distributions are used, the overpower penalty of the generic thermal margin analysis model selected by this method is not able to reflect various combinations of operating conditions and axial power distributions. Therefore, the overpower penalties may be derived on the basis of limited information. In addition, since the minimum value among the overpower penalties of the limited range is selected as the representative value of the overpower penalty, the thermal margin may be excessively conservative. The foregoing is intended merely to aid in the understanding of the background of the present invention and is not intended to mean that the present invention falls within the purview of the related art that is already known to those skilled in the art. Accordingly, the present invention is an overpower penalty calculation system for a thermal margin analysis model and is intended to provide a reasonable and improved thermal margin by adjusting the thermal margin using an overpower penalty calculated by a statistical method. In order to achieve the above objective, the present invention may be provided with a statistical overpower penalty calculation system for a generic thermal margin analysis model adjusting a thermal margin of a nuclear power plant using an overpower penalty calculated by a statistical method, the system including: a random number generating unit arbitrarily generating a plurality of random numbers; an power distribution generating unit generating axial and radial power information of a core; an operating condition generating unit extracting an arbitrary value for a plurality of operating conditions using the random numbers generated in the random number generating unit; a POL calculating unit calculating a POL of a reload core thermal margin analysis model and a POL of a generic thermal margin analysis model and calculating a plurality of the overpower penalties through the POL of the reload core thermal margin analysis model and the POL of the generic thermal margin analysis model; and a statistics processing unit calculating tolerance limit values according to the core burnup by statistically analyzing a distribution formed of the plurality of the overpower penalties and selecting a smallest tolerance limit value among the tolerance limit values as a representative value of the overpower penalty by comparing the tolerance limit values to each other, wherein the system may adjust the thermal margin of the nuclear power plant by reflecting the representative value in a calculation of the thermal margin. The random number generating unit may generate a plurality of random numbers corresponding to a DNBR probability distribution, pressure, inlet temperature, and flow rate for each axial power distribution, and the operating condition generating unit may extract an arbitrary value according to the operating condition with the DNBR probability distribution, pressure, inlet temperature, and flow rate as the operating conditions. The statistics processing unit may execute a normality test and perform a statistical analysis in which a parametric or nonparametric method is applied according to a result of the normality test. As described above, according to the present invention, values corresponding to an operating condition range are arbitrarily extracted using a plurality of random numbers, and the tolerance limit values of the calculated overpower penalties are statistically analyzed and reflected in the thermal margin, so that a more rational and non-conservative thermal margin than the existing thermal margin can be calculated. In other words, by applying an optimized overpower penalty to a calculation of the uncertainty of DNBR, which is the main operating variable of a nuclear power plant, the thermal margin of the nuclear reactor is increased and the operational margin of the nuclear reactor is increased, whereby an economic effect may be obtained. Herein below, the present invention will be described in detail with reference to the contents described in the accompanying drawings. However, the present invention is not limited or restricted by exemplary embodiments. The same reference numerals shown in each drawing indicate members that perform substantially the same function. Objectives and effects of the present invention may be naturally understood or be made more obvious by the following description, and the objectives and effects of the present invention are not limited only by the following description. In addition, in describing the present invention, when it is determined that a detailed description of a known technology related to the present invention may unnecessarily obfuscate the subject matter of the present invention, a detailed description thereof will be omitted. FIG. 1 shows a block diagram of a system 1 calculating an overpower penalty of a thermal margin analysis model according to an embodiment of the present invention. The system 1 calculating the overpower penalty of the thermal margin analysis model may adjust a thermal margin of a nuclear power plant by using the overpower penalty calculated through a statistical method. As the overpower penalty is calculated by the statistical method, an improved thermal margin may be obtained. With reference to FIG. 1, the system 1 calculating the overpower penalty of the thermal margin analysis model may include a random number generating unit 11, an power distribution generating unit 13, an operating condition generating unit 15, a power operating limit (POL) calculating unit 17, and a statistics processing unit 19. The random number generating 11 may generate a plurality of arbitrary random numbers. In addition, the random number generating unit 11 may generate a plurality of random numbers corresponding to a DNBR probability distribution, pressure, inlet temperature, and flow rate for each axial power distribution. In addition, the random number generating unit 11 assumes the DNBR probability distribution as a normal distribution, and generates a normal random number through Box-Muller transformation using two random numbers. In addition, each of the pressure, inlet temperature, and flow rate is assumed as a uniform distribution and is required to have one random number, whereby three random numbers are required. Therefore, five random numbers may be generated for each one axial power distribution. Representative core burnups are SBOC, LBOC, LMOC, and LEOC, and 1,200 axial power distributions may be applied to each burnup. Therefore, the random number generating unit 11 may calculate a random number for each of 1,200 axial power distributions of each of the four burnups according to conditions of the DNBR probability distribution, pressure, inlet temperature, and flow rate, thereby generating a total of 24,000 (4×1,200×5) random numbers. The power distribution generating unit 13 may generate power information of the axial direction and the radial direction of the core burnup. In addition, while the SBOC, LBOC, LMOC, and LEOC are the representative core burnups, the power distribution generating unit 13 may generate 1,200 axial power distributions at each burnup. In addition, the power distribution generating unit 13 may generate whole power information by combining axial power information and radial power information. At this time, the power distribution generating unit 13 may use geometry information of the core analysis model of the radial output information as it is and combine a maximum value of a ¼ fuel assembly of a reload core limiting fuel assembly candidate model with the axial power information, thereby generating the whole power information. The operating condition generating unit 15 may extract an arbitrary value for a plurality of operating conditions from the random number generated in the random number generator. In addition, the operating condition generating unit 15 may extract an arbitrary value in accordance with the operating condition with the DNBR probability distribution, pressure, inlet temperature, and flow rate as the operating condition. In addition, the operating condition generating unit 15 may assume the DNBR probability distribution as a normal distribution and calculate a value by applying two random numbers to the Box-Muller transformation. In addition, the operating condition generating unit 15 may assume the operating condition of each of the pressure, inlet temperature, and flow rate as a uniform distribution and apply one random number to a range for each condition, thereby calculating a value. The POL calculating unit 17 may calculate a POL of the reload core thermal margin analysis model and a POL of the generic thermal margin analysis model and calculate a plurality of the overpower penalties through the POL of the reload core thermal margin analysis model and the POL of the generic thermal margin analysis model. The POL of the reload core thermal margin analysis model is expressed as POLcycle, and the POL of the generic thermal margin analysis model may be expressed as POLgeneric. An N-th overpower penalty calculated using POLcycle and POLgeneric is as shown in equation 1. O ⁢ P N = POL ⁢ cycle POL ⁢ generic [ Equation ⁢ ⁢ 1 ] The statistics processing unit 19 may form a plurality of distribution for each of the burnups with a plurality of overpower penalties calculated in the POL calculating unit 17 and calculate a tolerance limit value according to the burnup by statistically analyzing the distributions formed above. Statistical analysis executes a normality test, and when the distribution of the overpower penalties follows the normal distribution, a parametric method may be used, and when the distribution of overpower penalties does not follow a normal distribution, a nonparametric method may be used. The statistics processing unit 19 may proceed statistical analysis with 95% confidence level for the distribution of overpower penalties and set the tolerance limit value calculated on the basis of 95% probability according to each burnup as the overpower penalty of the corresponding burnup. In addition, the statistics processing unit 19 may compare the plurality of the tolerance limit values according to each burnup and set the smallest tolerance limit value among the plurality of the tolerance limit values as a representative value of the overpower penalty of the corresponding core. The thermal margin of the corresponding core may be calculated by applying the representative value of the overpower penalties calculated statistically to the thermal margin analysis model. FIG. 2 shows a flowchart of calculating the overpower penalty by the system calculating the overpower penalty for the thermal margin analysis model according to the embodiment of the present invention. The overpower penalty calculation system 1 for the thermal margin analysis model generates five random numbers for each of 1,200 axial output distributions in the random number generating unit 11, and such a process is applied to each of a plurality of burnups. The power distribution generating unit 13 combines the axial power information and the radial power information to generate whole power information. The operating condition generating unit 15 calculates each condition value by applying random numbers to the Box-Muller transformation in the case of the DNBR probability distribution and by applying the random number to the range of the operating condition in the case of pressure, inlet temperature, and flow rate. The POL calculating unit 17 calculates POLcycle by performing POL calculation of the reload core thermal margin analysis model, and calculates POLgeneric by performing POL calculation of the generic thermal margin analysis model. In addition, the POL calculating unit 17 calculates and stores the overpower penalty OPN through a ratio of the calculated POLcycle and POLgeneric. The generation of the overpower penalties is performed 1,200 times as the same as the number of the axial power distributions, and 1,200 overpower penalties are generated and stored. The statistics processing unit 19 proceeds statistical analysis by generating a distribution of the overpower penalties generated in the POL calculating unit 17. In addition, a normality test is executed on the distribution of overpower penalties, and a parametric or nonparametric method is applied according to a result of the normality test. At this time, an tolerance limit value is derived on the basis of 95 percent confidence level and a 95 percent probability. The derived tolerance limit value is set as an overpower penalty for a specific burnup. A plurality of overpower penalties calculated for each burnup is compared to each other, and a minimum value is set as the representative value of the overpower penalty. It is possible to calculate an optimized thermal margin by applying the representative value of the overpower penalties to DNBR uncertainty analysis. FIG. 3 shows a radial power distribution of a reload core according to the embodiment of the present invention, and the limiting fuel assembly candidate is indicated by a thick solid line. FIG. 4 shows a result of statistical analysis of a distribution formed of a plurality of overpower penalties according to the embodiment of the present invention and shows an overpower penalty at a specific burnup according to the statistical analysis result. According to the exemplary embodiment of the present invention, FIG. 4 is a result in which a nonparametric method is applied after statistical analysis of the distribution of 1,172 overpower penalties has been performed showing the distribution of the overpower penalties not to follow the normality. At this time, the tolerance limit value calculated by applying the 95 percent confidence level and 95 percent probability is 1.012507, and this value is set as the overpower penalty of the corresponding burnup. The statistics processing unit 19 proceeds statistical analysis by generating a distribution of the overpower penalties generated in the POL calculating unit 17. In addition, a normality test is executed on the distribution of the overpower penalties, and a parametric or nonparametric method is applied according to a result of the normality test. At this time, an tolerance limit value is derived on the basis of 95 percent confidence level and a 95 percent probability. The derived tolerance limit value is set as an overpower penalty for the specific burnup. FIG. 5 shows an tolerance limit value calculated by a conventional overpower penalty calculation method and an tolerance limit values calculated by a statistical method in a distribution of overpower penalties formed of a plurality of overpower penalties according to the embodiment of the present invention. The conventional method was calculated in a conservative direction by selecting a minimum value among the overpower penalties calculated through the limited number of operating conditions and the axial power distributions when the overpower penalties of the generic thermal margin analysis model are calculated. According to FIG. 5, the overpower penalty calculated by the conventional method is 0.99. On the other hand, when the statistical method proposed according to the embodiment of the present invention is used, the calculated overpower penalty is 1.013. The method according to the embodiment of the present invention uses 1,200 axial power distributions of each core burnup and uses the statistical method using random numbers arbitrarily selected within a range of the operating conditions. As shown in FIG. 5, when the statistical method according to the embodiment of the present invention is used, the overpower penalty may be calculated in a direction in which the thermal margin is increased compared to the conventional method. It may be seen that the thermal margin increased by the statistical method according to the exemplary embodiment of the present invention is numerically increased by about 0.023 and proportionally by about 2.3 percent. Although the present invention has been described in detail through the representative embodiment above, those of ordinary skill in the art to which the present invention pertains will understand that various modifications may be made to the above embodiment without departing from the scope of the present invention. Therefore, the scope of the present invention should not be determined by being limited to the described embodiment and should be determined by all changes or modified forms derived from the equivalent concept of claims as well as the claims to be described later.
042736145	abstract	Handling device for a fast neutron reactor cooled by a liquid metal such as sodium and of the type comprising a container with a vertical axis containing the core and a given volume of sodium contained in said container, defining above the core a free level surmounted by an inert gas cushion, a thick horizontal slab placed above the gas cushion and constituting the upper closure of a protective caisson surrounding the container, an inclined ramp which passes into the container up to a loading and unloading station located in the vicinity of the core and extending at its opposite end into a handling chamber supported by the slab, a second inclined ramp connecting the inside of the handling chamber to a second loading and unloading station positioned externally of the caisson in a storage container, a carriage carrying a handling pot able to move along the said ramps and moving the pot from a position where it is submerged below the sodium level to a position where it is introduced into the inert gas atmosphere above the sodium level, said pot being open at its upper end so as to permit the introduction or removal of the assembly to be handled when it is positioned vertically on its carriage with respect to said loading and unloading stations provided for this purpose in the reactor vessel and in the storage container, wherein during its displacement between said stations the carriage is associated, to the right of the upper open end of the pot, with a generally cylindrical enclosure which is open towards the bottom and provided with an upper part and in which is mounted a hollow member, whereof at least the lower part has a spherical shape, whereby the hollow member bears via the spherical lower part against the upper open end of the pot when the latter emerges above said level.
	claims	1. A system for evaluation of objects, the system comprising:evaluation units;an object distribution system that is configured to receive the objects and distribute the objects between the evaluation units;at least one controller that is configured to control the evaluation units and the object distribution system;wherein each evaluation unit of the evaluation units comprises:a chamber housing that has an inner space;a chuck that is configured to support an object while the object is positioned within the inner space;a movement system that is configured to move the chuck;an intermediate element positioned between the chamber housing and the movement system;at least one sealing element configured to form a dynamic seal between the intermediate element and the chamber housing, wherein the dynamic seal is configured to seal the inner space from the movement system;a charged particle module that is configured to irradiate the object with a charged particle beam, and to detect particles emitted from the object; andwherein for each evaluation unit of the evaluation units, a length of the inner space is smaller than twice a length of the object, and a width of the inner space is smaller than twice a width of the object; andwherein the movement system is configured to perform, for each region of the object out of a plurality of regions of the object, (a) a rotation of the chuck to position a given portion of the region of the object within a field of view of the charged particle module; and (b) a movement of the chuck in relation to the charged particle module to position additional portions of the region of the object within the field of view of the charged particle module. 2. The system according to claim 1 wherein for each evaluation unit of the evaluation units:the length of the inner space is smaller than twice the length of the object and exceeds 1.5 times of the length of the object; andthe width of the inner space is smaller than twice the width of the object and exceeds 1.5 times of the width of the object. 3. The system according to claim 1 wherein the evaluation units are configured to evaluate the objects in parallel. 4. A system for evaluation of objects, the system comprising:evaluation units;an object distribution system that is configured to receive the objects and distribute the objects between the evaluation units;vibration dumping elements that are coupled between the evaluation units and the object distribution system;at least one controller that is configured to control the evaluation units and the object distribution system;wherein each evaluation unit of the evaluation units comprises:a chamber housing that has an inner space;a chuck that is configured to support an object while the object is positioned within the inner space;a movement system that is configured to move the chuck;a charged particle module that is configured to irradiate the object with a charged particle beam, and to detect particles emitted from the object; andwherein for each evaluation unit of the evaluation units, a length of the inner space is smaller than twice a length of the object, and a width of the inner space is smaller than twice a width of the object. 5. The system according to claim 4 wherein the vibration dumping elements are bellows. 6. A method for evaluating objects, the method comprising:receiving objects by an object distribution system;distributing the objects, by the object distribution system, between evaluation units; andevaluating the objects by the evaluation units;wherein the evaluating of the objects by the evaluation units comprises evaluating at least two objects by at least two evaluation units in parallel;wherein an evaluating of an object of the objects by an evaluation unit of the evaluation units comprises:positioning the object on a chuck of the evaluation unit and within an inner space that is defined by a chamber housing of the evaluation unit;sealing the inner space from a movement system of the evaluation unit;irradiating, by a charged particle module of the evaluation unit, the object with a charged particle beam;detecting, by the charged particle module, particles emitted from the object; andrepeating, for each region of the object out of a plurality of regions of the object, the steps of:rotating the chuck, by the movement system, to position a given portion of the region of the object within a field of view of the charged particle module; andmoving the chuck, by the movement system, in relation to the charged particle module to position additional portions of the region of the object within the field of view of the charged particle module;wherein for each evaluation unit of the evaluation units, a length of the inner space is smaller than twice a length of the object and a width of the inner space is smaller than twice a width of the object. 7. The method according to claim 6 wherein for each evaluation unit of the evaluation units the length of the inner space is smaller than twice the length of the object and exceeds 1.5 times of the length of the object, and the width of the inner space is smaller than twice the width of the object and exceeds 1.5 times of the width of the object. 8. The method according to claim 6 wherein the object has a radial symmetry and wherein the plurality of regions comprise four regions. 9. The method according to claim 6 comprising performing a calibration process of the evaluation units by comparing between evaluation results generated by different evaluation units in regard to a same object. 10. A method for evaluating objects, the method comprising:receiving objects by an object distribution system;distributing the objects, by the object distribution system, between evaluation units; anddumping vibrations by vibration dumping elements that are coupled between the evaluation units and the object distribution system;evaluating the objects by the evaluation units;wherein the evaluating of the objects by the evaluation units comprises evaluating at least two objects by at least two evaluation units in parallel;wherein an evaluating of an object of the objects by an evaluation unit of the evaluation units comprises:positioning the object on a chuck of the evaluation unit and within an inner space that is defined by a chamber housing of the evaluation unit;sealing the inner space from a movement system of the evaluation unit;irradiating, by a charged particle module of the evaluation unit, the object with a charged particle beam; anddetecting, by the charged particle module, particles emitted from the object;wherein for each evaluation unit of the evaluation units, a length of the inner space is smaller than twice a length of the object and a width of the inner space is smaller than twice a width of the object. 11. The method according to claim 10 wherein the vibration dumping elements are bellows.
	abstract	A method of decommissioning a nuclear facility, including: exposing the plurality of upper penetration holes by removing the plurality of sandboxes; enlarging an upper space of the cavity by cutting an upper portion of the biological shield concrete that is disposed between the plurality of upper penetration holes and between the plurality of upper penetration holes and the cavity; and separating the nuclear reactor pressure vessel from the biological shield concrete.
	claims	1. A system for filling a container with hazardous waste comprising:a primary confinement chamber;a filling head positioned in the primary confinement chamber, the filling head being configured to add the hazardous waste to the container, drive movement of a mixing mechanism in the container, and vent air from the container; anda lid handling mechanism positioned in the primary confinement chamber, the lid handling mechanism being configured to couple a lid to the container. 2. The system of claim 1 wherein the lid handling mechanism is configured to remove the lid from the container. 3. The system of claim 1 wherein the system is configured so that the interior of the container is only open to the primary confinement chamber as the container passes through the system. 4. The system of claim 1 wherein the system is configured to fill the container with radioactive waste. 5. The system of claim 1 wherein the system is configured to fill a steel drum with the hazardous waste. 6. The system of claim 1 comprising a secondary confinement system that includes a secondary confinement chamber configured to house the exterior of the container as the container is being filled by the filling head. 7. The system of claim 6 wherein the system is configured to, from the primary confinement chamber, add dry or wet cementitious material to the container, add cementitious material to seal off the lid of the container, measure the level of the container, and/or test whether the contents of the container meet quality assurance requirements. 8. A system for filling a container with hazardous waste comprising:a filling head configured to add the hazardous waste to the container, drive movement of a mixing mechanism in the container, and vent air from the container; anda lid handling mechanism configured to couple a lid to the container;wherein the system is configured to hold the container in a stationary position between when the filling head accesses the container and the lid handling mechanism accesses the container. 9. A system for filling a container with hazardous waste comprising:a primary confinement chamber;one or more mechanisms positioned in the primary confinement chamber, the one or more mechanisms being configured to add the hazardous waste to the container, drive movement of a mixing mechanism in the container, vent air from the container, add dry or wet cementitious material to the container, add cementitious material to seal off the lid of the container, measure the level of the container, and/or test whether the contents of the container meet quality assurance requirements; anda lid handling mechanism positioned in the primary confinement chamber, the lid handling mechanism being configured to couple a lid to the container. 10. A system for filling a container with hazardous waste comprising:one or more mechanisms that is configured to add the hazardous waste to the container, drive movement of a mixing mechanism in the container, vent air from the container, add dry or wet cementitious material to the container, add cementitious material to seal off the lid of the container, measure the level of the container, and/or test whether the contents of the container meet quality assurance requirements; anda lid handling mechanism configured to couple a lid to the container;wherein the system is configured to hold the container in a stationary position between when the filling head accesses the container and the lid handling mechanism accesses the container.
	claims	1. A method to evaluate an effect of applying a zinc compound to a reactor coolant system of a pressurized water reactor, comprising:quantitatively assessing a pressurized water reactor stress corrosion cracking initiation rate of a candidate reactor coolant system through analysis of operational eddy current testing data and pressurized water stress corrosion cracking failure history using empirical relationships;determining an extent of damage to the candidate system;quantitatively assessing pressurized water stress corrosion cracking initiation benefit for high-concentration (>10 ppb) and low-concentration (<10 ppb) zinc addition programs;calculating a pressurized water stress corrosion cracking initiation benefit from zinc addition at low concentrations (<10 ppb) in the candidate system;approximating a time T when zinc addition at low concentrations (<10 ppb) in the candidate system is needed for pressurized water stress corrosion cracking mitigation; andapplying zinc acetate to the reactor coolant system at the time T at concentrations of approximately between 1 to 10 parts per billion for pressurized water stress corrosion cracking mitigation,wherein zinc has not previously been applied to the candidate system and wherein the applying zinc acetate to the reactor coolant system at the time T at concentrations of approximately between 1 and 10 ppb for pressurized water stress corrosion cracking mitigation is not performed until after there are pressurized water stress corrosion cracking indications in the reactor coolant system. 2. The method according to claim 1, further comprising:applying a temperature scaling factor adjustment to the data to normalize differences between different locations within the reactor coolant system and between the reactor coolant system and reactor coolant systems of other pressurized water reactors. 3. The method according to claim 1, further comprising:developing a database of degradation rates, the database being adjusted to a common reference temperature. 4. The method according to claim 3, wherein the database is adjusted to effective full power years of operation of the pressurized water reactor. 5. The method according to claim 4, further comprising:developing a probabilistic predictive tool to trend and predict degradation in the reactor coolant system, the probabilistic predictive tool correlating the effective full power years of reactor operation and the normalized degradation rate. 6. The method according to claim 5, wherein the probabilistic predictive tool is developed from a change in a Weibull slope of pressurized water stress corrosion cracking initiation plotted data before and after zinc addition. 7. The method according to claim 1, further comprising:calculating a magnitude of pressurized water stress corrosion cracking mitigation due to zinc injection at nuclear power plants as a function of a mass of zinc incorporated into surface oxides of the reactor coolant system. 8. The method according to claim 1 wherein the applying zinc acetate to the reactor coolant system at the time T is performed at concentrations of approximately between 1 to 10 parts per billion that minimize a removal of zinc from the reactor coolant system. 9. The method according to claim 1 wherein the applying zinc acetate to the reactor coolant system at the time T at concentrations of approximately between 1 to 10 parts per billion for pressurized water stress corrosion cracking mitigation includes maintaining the zinc concentration in the reactor coolant system above the solubility of zinc chromite in the reactor coolant system. 10. A method to evaluate an effect of applying a zinc compound to a reactor coolant system of a pressurized water reactor, comprising:quantitatively assessing a pressurized water reactor stress corrosion cracking initiation rate of a candidate reactor coolant system through analysis of operational eddy current testing data and pressurized water stress corrosion cracking failure history using empirical relationships;determining an extent of damage to the candidate system;quantitatively assessing pressurized water stress corrosion cracking initiation benefit for high-concentration (>10 ppb) and low-concentration (<10 ppb) zinc addition programs;calculating a pressurized water stress corrosion cracking initiation benefit from zinc addition at low concentrations (<10 ppb) in the candidate system;approximating a time T when zinc addition at low concentrations (<10ppb) in the candidate system is needed for pressurized water stress corrosion cracking mitigation; andapplying zinc acetate to the reactor coolant system at the time T at concentrations of approximately between 1 to 10 parts per billion for pressurized water stress corrosion cracking mitigation,wherein zinc has not previously been applied to the candidate system and wherein the step of quantitatively assessing a pressurized water stress corrosion cracking initiation rate of a candidate system through analysis of operational eddy current testing data and pressurized water stress corrosion cracking failure history using empirical relationships comprises calculating a normalized degradation rate define as a number of tubes with new pressurized water stress corrosion cracking indications divided by a number of rotating coil examinations in a region of the number of tubes.
	description	1. Field of the Invention Embodiments of the present invention generally relate to the field of medical imaging and the field of radiography. More particularly, embodiments of the present invention relate to dual-energy scanners. 2. Description of the Prior Art Dual-energy scanners may be single-source scanners periodically emitting radiations with two different energies typically of the order of 80 kV and 140 kV respectively, and switching very rapidly from one to the other (of the order of about a thousand switches per revolution of the scanner, and from about 0.5 to about 5 revolutions per second). As these radiations with different energies are not transmitted or reflected in the same way by organic tissues, notable enrichment of the information obtained in the final image, and an increase in its resolution may be achieved. As shown in the graph of FIG. 1, which represents the energies of the emitted radiations versus time, these scanners nevertheless require transition times (called Trise and Tfall in the figure) for switching from one energy to the other and vice versa, during which the energy is variable and distinct from the two “useful” energies E1 and E2 used for the imaging of a patient. The emitted radiation doses during these transition times therefore only have a low interest for the image; furthermore, they provide an additional dose of radiations to the patient, whereas doses should be minimized in order not to be dangerous for the health of the patient. Efforts have already been made for limiting the transition time between the useful energies, but there remains a phase during which the patient receives an unnecessary dose. Furthermore, reducing the transition time complicates the structure of the electronic circuit used in the scanner and makes the latter heavier. Therefore, there exists a need for a novel technique with which images may be produced from radiations with two different energies without these images being altered by additional radiations of non-useful energies, and in which the doses of non-useful radiations absorbed by the patient are limited. Methods for applying sources of radiations have already been developed with which all or part of the radiations may be prevented from attaining a patient, in order to modulate the dose received by this patient. To achieve this, radiation sources of the type comprising a source of electrons and a target, adapted so as to emit a flux of X-rays towards a patient or an area of the patient to be imaged when it receives a flux of electrons, are used. The source further comprises a system for deflecting the flux of electrons, which modifies the path of the electron flux so that it attains another point of the target and the dose sent towards the patient is modified. However, with such a device, it is only possible to achieve dose modulation, but not to mask certain energies at given time intervals; the obtained result is therefore not satisfactory. According to an embodiment of the present invention, there is provided a medical imaging method. The method comprises generating a radiation at a first energy level by a radiation source, generating a radiation at a second energy level different from the first energy level by the radiation source, emitting the generated radiations at an output of the radiation source towards a detector, and blocking or diverting the emitted radiations during at least one intermediate phase during which the radiation source switches in a transient way from one of the first energy level and the second energy level to the other of the first energy level and the second energy level. According to another embodiment of the present invention, there is provided a medical imaging device. The device comprises a source of radiations, a detector of radiations, and a control module configured to control the source of radiations to generate a radiation at a first energy level, to generate a radiation at a second energy level different from the first energy level, and to emit the generated radiations at an output of the source of radiations towards the detector of radiations, wherein the emitted radiations are blocked or diverted during at least one intermediate phase during which the source of radiations switches in a transient way from one of the first and second energy levels to the other of the first and second energy levels. With reference to FIG. 2, a tomography device 10 comprising a source of radiations 11 and a detector of radiations 13, positioned on a rotating support 20 is illustrated. The source of radiations 11 emits a beam of radiations 12, for example X-rays, towards the detector 13 and through a patient P, or an area of a patient P to be imaged, lying on a support 21. When the detector 13 receives the radiations 12, with a processing unit 15 connected to the detector 13, it is possible to store the images obtained by the detector 13 and optionally perform additional processing on these images, in order to, for example, reconstruct a 3D image of the area of the patient P to be imaged. Further, the tomography device 10 comprises a module 14 for controlling the source 11, which is connected to the source 11 and in particular controls the dose and the energy of the radiations 12 emitted by the source 11 towards the patient P. In the case of a dual-energy tomography device, the source 11 should only emit towards the patient P, radiations with two distinct energy levels E1 and E2 useful for forming the image. The moments during which the transitions between both of these energy levels have to be made depend on the angular position of the rotating support 20. The rotating support 20 at these moments sends to the control module 14 the order to modify the energy of the radiations, and the control module 14 modifies the energy of the source 11 according to these orders. This control module 14 may also control the source 11 so that the latter only delivers towards the detector 13 and through the patient P, or an area of the patient P to be imaged, radiations with two distinct energy levels E1 and E2 useful for forming the image, and does not deliver transition energies between both of these energy levels. Optionally, the control module 14 may, in order to control the dose of the radiations emitted towards the patient P, also be connected to the detector 13 and use information on the radiation dose received by the detector 13 in order to adapt the dose of radiations 12 emitted by the source 11. The source of radiations 11 is illustrated in more detail in FIGS. 3a-3b. It comprises a source of electrons 111 and an anode or target 114 which emits radiation, for example of the X-ray type, when it receives a flux of electrons 112. The target 114, illustrated in FIG. 3a, comprises a first focal area F1 adapted so as to emit the radiation 12 towards the detector 13 through the patient P when it receives the flux of electrons 112 emitted by the source of electrons 111. The target 114 may also comprise a second focal area F2, illustrated in FIG. 3b, which, when it receives all or part of the flux of electrons 112 emitted by the electron source 111, may emit a beam of radiations 22 which is diverted or blocked so as not to be received by the patient P. Alternatively, the second focal area F2 may emit no radiation when it receives a flux of electrons. Finally, the source of radiations 11 comprises a deflection system 113, positioned between the source of electrons 111 and the target 114, and which may modify the trajectory of the electron beam 112. This deflection may be achieved in any known way, for example by magnetic or electrostatic deflection. During an examination of the patient P, the electron source 111 successively generates a flux of electrons 112 at a first energy level, at a second energy level distinct from the first energy level, and during intermediate phases at a variable energy level during the time for switching in a transient way from the first to the second energy level or vice versa. The energy levels are adapted so that the radiation emitted by the target 114 and resulting from this flux of electrons 112, has the profile illustrated in FIG. 1. In particular, the resulting radiation 12 of this flux of electrons 112 successively has, over time, first and second energy levels E1 and E2 respectively, and intermediate phases Trise and Tfall, during which the energy level switches in a transient way from the first energy level E1 to the second energy level E2 or vice versa. According to a first embodiment illustrated in FIG. 4a, the radiation source 11 is controlled so that during stationary phases, during which the energy level is constant or equal to E1 or E2, the flux of electrons 112 generated by the source of electrons 111 attains the focal area F1, so that the latter generates radiation 12 towards the detector 13 through the patient P. On the other hand, during the intermediate phases Trise and Tfall, the radiation source 11 is controlled so that the flux of electrons 112 does not reach the focal area F1, and thus no radiation is generated in the direction of the patient P. The intermediate phases Trise and Tfall, during which the energy of the electron flux 112 varies from a state E1 or E2 to the other state, and where no transient radiation is emitted towards the patient P, are shown as hatched lines in FIG. 4a. With reference to FIG. 4b, the duration of the phase during which no radiation is emitted through the patient P may also be adapted in order to modulate the dose received by the latter. Indeed, although it is possible to limit the duration of this phase to the intermediate phases Trise and Tfall, it is also possible to extend the blocking of the radiation before and after each of these phases, in particular in order to reduce the dose of radiations 12 received by the patient P. These areas are also illustrated with hatchings in FIG. 4b. In order to achieve this blocking of the radiations, or more generally to prevent the electron flux from reaching the focal area F1 of the target 114, several embodiments are possible. Generally, the electron flux 112 may reach the first focal area F1 during the stationary phases, and be only diverted during the intermediate phases Trise and Tfall. Alternatively, the flux of electrons 112 may only be diverted during the stationary phases so as to reach the first focal area F1, or further it may be diverted according to several different trajectories depending on whether one is in a stationary phase or an intermediate phase. Furthermore, the way to prevent the electron beam 112 from reaching the focal area F1 during intermediate phases may also vary. According to an embodiment, with reference to FIG. 5a, the radiation source 111 may comprise an electron collector 115 positioned between the deflection system 113 and the target 114. The deflection system 113 diverts the trajectory of the electron beam 112 onto the collector 115 so that this beam follows a circular trajectory onto the collector 115 for example. To do this, this electron collector 115 may be an axisymmetrical solid comprising a through-aperture through which passes the electron beams 112 during the stationary phases, in order to reach the focal area F1 of the target 114. For example, the collector 115 may be an axisymmetrical solid centered on the electron beam 112 during the stationary phases. It may also have an inner surface delimiting the through-aperture through which passes the electron flow, and towards which is diverted the trajectory of the electron flux 112 during the intermediate phases. The electron collector 115 may consist of a material such as copper, beryllium or ceramics of the alumina type (Al2O3), with which all or part of the electron flux may be absorbed without emitting any radiation, and further having good thermal properties such as good heat conduction, capacity and durability at high temperatures. Alternatively, the electron collector 115 may emit radiations and a collimator 116, positioned between the collector 115 and the patient P, may be adapted for blocking the radiations stemming from the collector 115 while transmitting the radiations stemming from the first focal area F1. The geometry of the collector 115 in the form of an axisymmetrical solid is preferable since it allows the collector to be driven into rotation around its axis of revolution, which allows an increase in the surface area against which the electron flux is diverted, and thus any overheating of this surface may be avoided. Alternatively, as illustrated in FIGS. 5b, 5c, and 5d, the electron beam 112 emitted during the intermediate phases may be diverted towards a second focal area F2 absorbing the electrons without emitting any radiation or emitting radiations towards a direction distinct from that of the patient P and of the detector 13. The target 114 may have the shape of axisymmetrical solid with axis X-X, having a surface tilted relatively to the incident electron flow 112, and the focal areas F1 and F2 may be portions of the target 114 in the form of concentric and distinct rings distinct from the target 114. This allows the target 114 to be driven into rotation around its axis X-X and the surface area of the focal areas may thereby be increased in order to avoid their overheating when they are exposed to the electron flux. Furthermore, it is also possible to defocus the electron beam 112 during the intermediate phases for limiting heating-up of the target 114. According to an embodiment, as illustrated in FIG. 5b, the second focal area F2 may be a focal area consisting of a material such as copper, beryllium or ceramics of the alumina type (Al2O3). The second focal area F2 may be adapted for absorbing all or part of the electron flux 112 without emitting any radiation, and further may have good thermal properties such as good heat conduction, heat capacity and durability at high temperature. A collimator 116 (not shown) positioned downstream from the source 111, if required, gives the possibility of only transmitting the radiations stemming from the first focal area F1. This collimator 116 may, for example, consist of at least two non-aligned windows, allowing limitation of the aperture transversely to the incident radiations, but also blocking of the radiations stemming from directions other than that of the first focal area F1. According to an alternative embodiment illustrated in FIG. 5c, the focal area F2 may be adapted in order to emit radiation 22 towards a direction different from that of the detector 13 and of the patient P. To do this, the second focal area F2 may, for example, have a different tilt relatively to the incident flux 112 of electrons from that of the first focal area F1. In this case, the radiation 22 is stopped by a collimator 116 positioned downstream from the target 114, and adapted for only letting through the radiation 12 stemming from the first focal area F1, and therefore for blocking radiations 22 stemming from the second focal area F2. This collimator 116 may then have a structure identical with the one shown earlier. Alternatively, the second focal area F2 may be a groove made in the target 114, as illustrated in FIG. 5d, emitting radiation 22 which is in majority confined in the groove. To do this, the groove may be positioned so that the radiation 22 is emitted towards a wall of the groove (not shown in the figures) and not towards the outside of the groove. Nevertheless, a small proportion of the radiations may be emitted outwards, so as to be blocked by the collimator 116 in the same way as earlier, since it is not emitted in the same direction as the first focal area F1. Finally, during the whole duration of exposure of the patient P, the control module 14 of the source 11 measures the radiation doses received by the detector 13 and therefore by the patient P. From these measurements, it may order the source 11 to block or to divert the radiations according to the embodiments described above, in order to limit the dose received by the patient P depending on a dose level per image to which the patient P may be subject. Thus in every case, during the intermediate phases Trise and Tfall, during which the energy of the radiations is variable over time, no radiation reaches the patient P or the detector 13, so that the patient P is not subject to too large of a dose and the detector 13 does not receive any parasitic radiation which may deteriorate the quality of the image obtained.
	description	This application is a continuation of and claims the benefit of priority under 35 U.S.C. §120 from U.S. Ser. No. 11/434,195 filed May 16, 2006, and claims the benefit of priority under 35 U.S.C. §119 from Japanese Patent Application No. 2005-143884 filed May 15, 2005, the entire contents of each of which are incorporated herein by reference. The present invention relates to a measurement method of an axial void fraction distribution in a core of a boiling water reactor (BWR), and also to an evaluating method of a neutron multiplication factor of a fuel assembly to be contained in a container apparatus. In a BWR, cooling water is heated and boiled in a reactor core as it flows from the bottom of the core to the top of the core. So, a void fraction that is a ratio of the bubbles to space where the cooling water flows in a channel box, increases as the cooling water flows from the upstream (the bottom of the core) to the downstream (the top of the core). The void fraction significantly affects core characteristics such as reactivity, power distribution and cooling characteristics. Therefore, it is important to quantitatively evaluate the void fraction distribution. However, no report has been made on an actual measurement of the void fraction distribution in the core of a commercial BWR because of lack of feasible measurement technique. Conventionally, a mock experiment is conducted out of the commercial reactor to build a theoretical model, and based upon the model, the void fraction distribution of the actual BWR is computationally evaluated. It has been recognized that the measurement of the void fraction distribution is necessary for a long time. It was reported that a change of nuclear characteristics due to a change of void fraction was indirectly measured by simulating voids in a critical facility that rarely emit thermal power. In this measurement, a wire containing manganese (Mn), called Manganin (Trade Mark) wire was installed in an experimental reactor core, and a thermal neutron absorber, cadmium (Cd), was wound around the Manganin wire. A cadmium ratio (Cd-ratio) in Mn55 (n, γ) reactions was measured. In this measurement, a wire containing manganese (Mn) (Manganin wire) was partly wrapped with cadmium (Cd), which is a thermal neutron absorber, and introduced into the reactor core of the experiment system. Then, the cadmium ratio in the reaction ratio of the reaction of Mn55 (n, γ) was measured. Various diameters of aluminum tubes were introduced in a test core and Cd-ratios were measured. These measurements show that Cd-ratio has a good correlation with a void fraction. So the report says that a void fraction can be measured by measuring Cd-ratio. The basic principle of the proposed technique is excellent, because it is based on the fact that a void fraction has a good correlation with the ratio of the non-thermal neutron flux and the thermal neutron flux. One of the inventors of the present invention proposed more practical techniques that use the basic principle described above but does not require the use of Cd in Japanese Patent Application Disclosure No. Sho 55-121195 and Japanese Patent Application Disclosure No. Sho 55-125489. According to these documents, a strong thermal neutron absorber such as Gd2O3 that has a high melting point or a weak neutron absorber such as stainless steel is locally arranged in fuel assemblies, in-core instrumentation tube, a fixed position neutron detector or a movable neutron detector to cause a local distortion in a thermal neutron flux. Then, a thermal neutron flux and a non-thermal neutron flux are separated, and a void fraction is determined from a ratio of the two neutron fluxes. The above-described technique is hard to apply to an operating nuclear reactor, because it employs Cd having a low melting point that might harm structure materials of the operating reactor. The inventors of the present invention have been studying new techniques of utilizing an in-core instrumentation tube. One of the techniques is a void fraction measurement method using a ratio of a non-thermal flux and a thermal flux, as well as a ratio of a thermal neutron flux and a fast neutron flux that has a larger dependency on a void fraction than non-thermal neutron flux. To use this technique, it is necessary to discharge the in-core instrumentation tube from the reactor and measure the tube during an outage of reactor operation. At the time of measurement, a dedicated holder has to be prepared in order to accurately place the instrumentation tube and the detector because the shape of the tube has to be maintained finely and accurately for measurement. In addition, since the instrumentation tube is arranged in the water gap between the fuel assemblies, a measured void fraction is an average of at least four surrounding fuel assemblies, and the instrumentation tube and the detection sensitivity is significantly lower than inside of a fuel assembly. As described above, while the axial void fraction distribution is very important for a BWR, no technique has been available for actual measurement of an axial void fraction distribution in an operating commercial nuclear reactor. When discharging an irradiated BWR fuel assembly from the core and containing it in a fuel assembly containing apparatus such as a transport vessel (cask) or a fuel storage rack in water, a neutron multiplication factor (or reactivity) at the axial position of ⅔ to ¾ from the lower end of the axial fuel active part tends to be higher. Such a trend appears due to design requirements such as that the flat axial thermal power distribution is preferable during the operation and due to a delay of burning of uranium and generated plutonium in an upper part (downstream of the coolant flow) because of an effect of the axial void fraction distribution and of a high generation rate of plutonium as a result of a high conversion ratio. Under these circumstances, for a purpose of an assured sub-criticality analysis, i.e. making sure of criticality safety, it is necessary to take an axial void fraction distribution into account to evaluate a multiplication factor. One way to achieve this purpose is to include a large design margin. In view of the above-identified problems, it is therefore the first object of the present invention to provide a novel and highly feasible method to evaluate an axial void distribution. The second object of the present invention is to provide a method to experimentally evaluate the neutron multiplication factor of a fuel assembly so that a large design margin may not be required to avoid a critical accident while containing an irradiated (discharged) BWR fuel assembly in a fuel assembly containing apparatus. The present invention has been made to solve the above problems, and has an object of providing axial void fraction distribution measurement method and neutron multiplication factor evaluating method. According to an aspect of the present invention, there is provided a method for evaluating an axial void fraction distribution of a fuel irradiated in a nuclear reactor, the method comprising: measuring a first intensity Az of a type of radiation (radioactive ray) emitted from a nuclide of a first group at an axial position of the fuel, the first group consisting of radioactive nuclides generated by a neutron capture reaction of a heavy nuclide or a fission product nuclide; measuring a second intensity Bz of a type of radioactive ray emitted from a nuclide of a second group at an axial position of the fuel, the second group consisting of radioactive fission product nuclides except nuclides generated by a neutron capture reaction; measuring a first reference intensity A0 of the same type of radioactive rays of the first intensity at an axial reference position of the fuel at which a void fraction of the fuel can be evaluated; measuring a second reference intensity B0 of the same type of radioactive rays of the second intensity at the axial reference position; calculating a exponent constant α used in an expression of Az=az×Eα and A0=a0×Eα where E is an exposure of the fuel, az and a0 are proportionality constants; evaluating a value of (az/a0) by an equation of (az/a0)=(Az/A0)(B0/Bz)α(bz/b0), where bz is a value used in an expression of Bz=bz×E as a proportionality constant, b0 is a value used in an expression of B0=b0×E as a proportionality constant; evaluating a correlation curve of (az/a0) and a void fraction; and evaluating the axial void fraction distribution based on the value of (az/a0) and the correlation curve. According to another aspect of the present invention, there is provided a method for evaluating an axial void fraction distribution of a fuel irradiated in a nuclear reactor, the method comprising: measuring a first intensity Az of gamma ray emitted from cesium 134 or europium 154 at an axial position of the fuel; measuring a first reference intensity A0 of a same type of radioactive rays of the first intensity at an axial reference position of the fuel at which a void fraction of the fuel can be evaluated; measuring a second intensity Bz of gamma ray emitted from cesium 137 or cerium 144 at an axial position of the fuel; measuring a second reference intensity B0 of a same type of radioactive rays of the first intensity at the axial reference position; calculating Gz/G0=(Az/Bz)/(A0/B0); evaluating a correlation curve of (Gz/G0) and a void fraction; and evaluating the axial void fraction distribution based on the value of (Gz/G0) and the correlation curve. According to yet another aspect of the present invention, there is provided a method for evaluating an axial void fraction distribution of a fuel irradiated in a nuclear reactor, the method comprising: measuring a radiation ray intensity of radioactive rays emitted from nuclides at an axial position of the fuel for at least twice with a definite interval, the nuclides generated by a neutron capture reaction of a heavy nuclide or a fission product nuclide; measuring a radiation ray reference intensity of a same type of radioactive rays of the radioactive ray intensity at a reference position for at least twice with a definite interval, the axial reference position of the fuel at which a void fraction of the fuel can be evaluated; dividing the radiation ray intensity into that of a first neutron emission rate Az from curium 242 and that of a second neutron emission rate Bz from nuclides except curium 242; dividing the radiation ray reference intensity into that of a first reference neutron emission rate A0 from curium 242 and that of a second reference neutron emission rate B0 from nuclides except curium 242; calculating a of an exponent constant expressed in equations of Az=az×Eα, A0=a0×Eα by using the exposure E, and proportionality constants az and a0; calculating p of an exponent constant expressed in equations of Bz=bz×Eβ, B0=b0×Eβ by using the exposure E, and proportionality constants bz and b0; calculating (az/a0)α(b0/bz)β by an equation of (az/a0)α(b0/bz)β=(Az/A0)α(B0/Bz)β; evaluating a correlation curve of (az/a0)α(b0/bz)β and a void fraction; and evaluating the axial void fraction distribution based on the value of (az/a0)α(b0/bz)β and the correlation curve. According to yet another aspect of the present invention, there is provided a method for evaluating a neutron multiplication factor of a fuel assembly irradiated in a nuclear reactor, the method comprising: measuring a neutron counting rate φ0 at a reference position of the fuel assembly where a void fraction is known, and evaluating a neutron multiplication factor k0; measuring a neutron counting rate φz at a multiplication factor evaluation point of the fuel assembly; calculating φ0/φz; measuring a gross gamma intensity ratio (γg/γg0) at the multiplication factor evaluation point and the reference position; evaluating a relationship between the gross gamma intensity ratio and an exposure of the fuel assembly; evaluating an exposure ratio (Ez/E0) at the multiplication factor evaluation point and the reference position based on the relationship between the gross gamma intensity ratio and an exposure of the fuel assembly; calculating (Ez/E0)α based on a value of α calculated otherwise; evaluating a relationship between (az/a0) and a void fraction; evaluating an axial void fraction distribution; evaluating the value of (az/a0) based on the axial void fraction distribution and the relationship between (az/a0) and a void fraction; and calculating the neutron multiplication factor k as k=1−(1−k0)(φ0/φz)(Ez/E0)α (az/a0). Hereinafter, embodiments of the present invention will be described with reference to the drawings. The present invention provides a highly practical method of evaluating an axial void fraction distribution of a fuel assembly irradiated in a nuclear reactor (discharged fuels and reload fuels). Attention is paid to the neutron emission rate from the neutron emitting nuclides (radioactivity) generated from uranium and plutonium by capturing neutrons, gamma ray intensity of fission products (radioactivity), and gamma ray intensity of transmuted fission products by capturing neutrons (radioactivity) among various radiations from an irradiated BWR fuel assembly. The present invention also provides a method of evaluating an axial distribution of a neutron multiplication factor of the fuel assembly based on an axial void fraction distribution. At first, to help understanding the above methods, phenomenon relating to the embodiments are described in detail below. <<Characteristics of Neutron Emission Rate of Target Nuclides>> As a spontaneously emitted neutron, a neutron is emitted by photonuclear reactions of high energy gamma rays emitted from fission product La-140 with hydrogen atoms shortly, for example 2 months, after a discharge of a fuel assembly from a reactor. The embodiments are for the fuel after such a cooling period. It is convenient to categorize nuclides into Cm-242 having a short half-life of 162.8 days, Cm-244 having a relatively long half-life of 18.1 years and other nuclides (residuals). The embodiments use this way to categorize. The neutron emission rate of Cm-242, that of Cm244 and that of other nuclides are expressed respectively as S2, S4 and SR hereinafter. While isotope generation and depletion analysis code “ORIGEN” takes U-234, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, Cm-243, Cm-246 and Cf-252 into account as nuclides that belong to SR, there is no problem to deal only with Pu-238 through Am-241 in the above list in practice. For example, a generation process of such neutron emitting nuclides in a uranium fuel is described below by referring to FIG. 9. In FIG. 9, σc represents neutron capture cross section (nuclear fission cross section is excluded) for thermal neutrons of which energy is 2200 m/s as its velocity. Ic represents a neutron capture cross section of a resonance region. Ic is also referred to as infinitely-diluted resonance integral of capture. When a void fraction of a moderator is large, neutrons are less moderated and a ratio of thermal neutrons decreases. Then a neutron capture ratio of a resonance region is higher than a neutron capture ratio in a thermal region. Hereinafter, Capture Ratio CR is defined as Ic/σc. A larger Capture Ratio may result in a higher sensitivity of the reaction on a void fraction. CR of B-10 (CR=0.449) is also cited in FIG. 9 for a reference. There is no resonance phenomenon in resonance region for B-10. A large amount of U-238 is contained in a fuel, and CR of U-238 contained in a fuel is very large. So, a neutron capture reaction of U-238 generating U-239 is largely affected by a void fraction. Although a self-shielding effect decreases an effective CR, CR remains large. U-239 decays to Np-239 with a half-life of 23.5 minutes, and Np-239 decays to Pu-239 with a half-life of 2.35 days. CR of Pu-239 is much smaller than that of U-239 or other nuclides. For example, CR of Pu-239 is smaller than twice of CR of B-10. So, a dependency of a neutron capture ratio of Pu-239 to a void fraction is small. Pu-239 fissions and also becomes Pu-240 by a neutron capture reaction. A void fraction dependency of Pu-240 is relatively large because it has a relatively large CR. Pu-240 spontaneously fissions and emits neutrons (spontaneous fission neutrons). Its neutron emission rate (NER) is relatively large. On the other hand, Pu-240 becomes Pu-241 by a neutron capture reaction. Pu-241 undergoes fission and also becomes Pu-242 by a neutron capture reaction. However, a void fraction dependency is small because of a small CR. Also, Pu-241 decays to Am-241 with a half-life of 14.4 years. Am-241 decays, and releases alpha particle. This alpha particle cause (α, n) reaction with oxygen in the fuel and emit neutrons ((α, n) neutrons). Am-241 becomes Am-242 by a neutron capture reaction. This reaction is dependent on a void fraction to a relatively small extent. Am-242 decays to Cm-242 with a half-life of 16 hours. A spontaneous fission neutron emission rate and an (α, n) neutron emission rate (S2) of Cm-242 is very large. Cm-242 decays to Pu-238 by alpha decay. An (α, n) neutron emission rate of Pu-238 is relatively large. Pu-241 becomes Pu-242 by a neutron capture reaction that is less dependent on a void fraction. Pu-242 becomes Pu-243 by a neutron capture reaction that is much dependent on a void fraction and it becomes Am-243 with a half-life of 5 hours. Am-243 becomes Am-244 by a neutron capture reaction that is much dependent on the void fraction. Am-244 decays to Cm-244 with a half-life of 10 hours. Cm-244 decays to Pu-240 again by a decay with a half-life of 18.1 years. But a neutron emission rate of Cm-244 (S4) is very large mainly due to spontaneous fissions. As described above, neutron capture reactions of any nuclides listed in the thick solid line boxes in FIG. 9 (U-238, Pu-240, Pu-242 and Am-243) depend on the void fraction. And neutron emission rates of nuclides generated by the reaction are influenced by the void fraction experienced during an irradiation of the fuel. So, a neutron emission rate of Cm-242 (S2) is defined mainly by a void fraction dependency of neutron capture reactions of U-238 and Pu-240. Also, a neutron emission rate of Cm-244 (S4) is defined mainly by a void fraction dependency of neutron capture reactions of Pu-242 and Am-243. A neutron emission rate of any nuclides belonging to SR is defined mainly by a void fraction dependency of neutron capture reactions of U-238 and Pu-240. <<Characteristics of Gamma Ray Measurement Target Nuclides>> Cesium 137 (Cs-137) is directly generated during a nuclear fission without any neutron capture reaction and it decays with a half-life of 30.2 years. A gamma ray of 662 keV emitted from Cs-137 as it decays is measured. A Cs-137 generation rate is roughly proportional to an exposure of the fuel. The measurement might be less accurate if the cooling time is less than a half year or the exposure of the fuel is small, because of a very high background of gamma rays emitted from Cs-134, Zr-95, Nb-95 and other nuclides. In such a case, it is preferable to utilize cerium 144 (Ce-144) or other nuclides in place of Cs-137. A generation rate of Cs-137 is barely dependent on a void fraction. In general, differences between Cs-137 yields of different fissile nuclides are small. Cs-133 is a stable nuclide generated in a fission, and it becomes Cs-134 by a neutron capture reaction. Cs-134 decays with a half-life of 2.06 years. The Cs-134 generation rate is roughly proportional to a square of the exposure. Since Cs-133 cause a large resonance capture like U-238, a generation rate of Cs-134 depends on a void fraction. Gamma rays of 796 keV and 1.365 MeV are measured easily among gamma rays emitted after beta decay of Cs-134. Differences between Cs-133 yields of different fissile nuclides are small. Europium 153 (Eu-153) is a stable fission product nuclide and it becomes europium 154 (Eu-154) by a neutron capture reaction. Eu-154 decays with a half-life of 8.5 years. Since Eu-153 causes a large resonance capture like U-238, a generation rate of Eu-154 depends on the void fraction. A generation rate of Eu-154 is approximately proportional to a square of the exposure. It is especially easy to measure a gamma ray of 1.274 MeV among gamma rays emitted by Eu-154 by beta decay. Differences between Eu-153 yields of different fissile nuclides are large. Therefore, the generation rate of Eu-154 is influenced by the composition of the fuel. Thus it is necessary to pay attention that amount of generated Eu-154 is slightly saturated for a high exposure regime because of the large neutron capture cross section. Cerium 144 (Ce-144) is directly generated during a nuclear fission without any neutron capture reaction and it decays to Pr-144 with a half-life of 284 days by a beta decay. It is particularly easy to measure a gamma ray of 2,186 MeV emitted immediately after beta decay with a half-life of 17.3 minutes. A generation rate of Ce-144 is proportional to the exposure for a low exposure regime, but the rate is saturated for a relatively low exposure because of the short half-life. Differences between Ce-144 yields of different fissile nuclides are small. Therefore, the generation of Ce-144 depends less on the void fraction. Because it has a short half-life of 284 days, it is preferable to use Ce-144 for the fuel irradiated for a short period and having a small exposure. It may be worth considering a use of Zr-95 or Nb-95 that has a short half-life if the exposure is low and the cooling time is short. The reasons why the above-described characteristics can be used for void fraction measurements are described below in detail. FIG. 10 is a plot illustrating the relationship between a neutron emission rate and an exposure of a BWR fuel assembly calculated by ORIGEN. The calculation was performed for an ordinary commercial BWR fuel of 4% enrichment irradiated at 40% of void fraction (void ratio 0.40). Both of the axes of the plot are logarithmic. The period elapsed since the end of irradiation, i.e. the cooling period, is 1 year. The vertical axis represents a quantity of neutrons emitted from one ton of fuel metal per second. A natural logarithmic values of the neutron emission rates, normalized as the neutron emission rate of 105 equal to zero, are also shown in the plot. The exposure is a thermal power derived in Giga-watt from one ton of fuel metal. Each curve in the plot includes a part that can be regarded as a straight line within a certain exposure range. The straight part of the neutron emission rate means that the rate is proportional to a power of the exposure in the range. A gradient of the straight part of the curve differs from each other. The difference indicates that the exponent values are different from each other. FIG. 11 shows curves shown in FIG. 10 normalized to the values at the exposure of 10 GWd/t. In FIG. 11, exposure dependencies of concentrations of gamma ray emitting nuclides relative to the values at 10 GWd/t are also shown. Although not shown in FIG. 11, the curve of Eu-154 is almost identical to that of Cs-134. Because of its long half-life of 30 years (the irradiation period is about 5 years for 60 GWd/t in FIG. 11), the Cs-137 concentration is almost proportional to the exposure. Because the concentration of Cs-137 is almost proportional to the exposure (the exponent value=1.0), it can be seen that the exponents to the exposures of curves except Cs-137 are greater than 1.0. FIG. 12 illustrates some of the results of a study whether it is possible to evaluate a void fraction by using calculated values by using two curves (straight lines in terms of their logarithms) having different gradients. FIG. 12 shows the relationships between natural logarithmic values of the neutron emission rate (S4R) of nuclides except Cm-242 in an irradiated fuel after one-year cooling and natural logarithmic values of the exposure. The curves in FIG. 12 are calculated by ORIGEN for different void fractions (0, 40, 70%) using void fraction dependent cross sections. It can be seen in FIG. 12 that these lines are on straight lines from about 18 GWd/t to about 45 GWd/t and can be approximated as S4R=az×Eα (E: exposure, az: proportionality constant). A neutron emission rate is higher for a higher void fraction. The value of α at 0% void is 4.9 in the range between 20 GWd/t and 40 GWd/t of exposure. Since a concentration or an intensity of gamma ray of Cs-137 is proportional to the exposure, it can be expressed that Cs137=bz×E (bz: proportionality constant). The exposure can be eliminated from this equation by raising both sides to the power of α and substituting Eα in the equation of S4R, i.e.,az=S4R×(bz/Cs137)α. The above equation can be rewritten by using a ratio relative to a reference void fraction (which is assumed to be 0% here for convenience) if the value of bz can be regarded as a constant for the entire axial position of the fuel assembly (bz/b0=1), i.e.,(az/a0)=[S4R/(S4R)0]×[(Cs137)0/(Cs137)]α,where the subscript “0” means that it is a value at a reference position. Thus, the relative value (az/a0) can be determined by the two measured ratios in the right term. On the other hand, this relationship between the relative value and the void fraction can be calculated as a ratio relative to the reference void fraction in each of the curves illustrated in FIG. 12. Thus, it is possible to evaluate a void fraction. Although it is desirable that the ratio would not be dependent on the exposure, it is slightly dependent on the exposure in reality. Therefore, it is also necessary to estimate the exposure. The exposure needs to be determined as accurately as possible depending on the combination of groups of radio activities as described below. In general, to measure a void fraction distribution, radio activities are categorized into two groups and they are evaluated. The first group consists of nuclides of which intensity Az can be expressed proportional to an exponent of the exposure of the fuel, i.e., Az=az×Eα. The intensity Az is referred as a first intensity. The second group consists of nuclides of which intensity Bz can be expressed proportional to the exposure of the fuel, i.e., Bz=bz×E. The intensity Bz is referred as a second intensity. As the first group of radio activities, nuclides that are generated by neutron capture reactions of heavy nuclides or that are transmuted from fission products by neutron capture reactions can be used. As the second group of radio activities, nuclides generated by fissions and capturing no neutrons can be used. Az and Bz are measured or evaluated at various axial position of the fuel assembly, as well as A0 and B0 corresponding to Sz and Bz at a reference position where the void fraction is known or the void fraction can be evaluated easily, such as a bottom of the fuel assembly where the void fraction is almost 0%. A0 and B0 are expressed as A0=a0×Eα and B0=b0×E. The exposure can be eliminated from the equation of Bz and B0 by raising both sides to the power of α and substituting Eα in the equation of Az and A0. Finally, the following equation is derived,(az/a0)=(Az/A0)(B0/Bz)α(bz/b0)α. On the other hand, a relationship between (az/a0) and a void fraction is evaluated. Based on the relationship and the value of (az/a0), a void fraction of the fuel is evaluated. FIG. 13 shows curves of the ratio of S4R and Cs137 selected from FIG. 11 for a void fraction of 40% corresponding to FIG. 12. From FIG. 13, it is understood that this method is applicable to a range of exposure between 20 GWd/t and 50 GWd/t. FIG. 14 illustrates an example of the use of a combination of Cs137 and S4, i.e., the neutron emission rate of Cm-244. It is convenient that the half-life of Cm-244 is as long as 18.1 years, but this combination would not be preferable for high exposure like S4R, nor for a low exposure, for example lower than 15 GWd/t, because it totally disappears in SR. The contribution of SR has to be removed by calculated within and near the range of 10 to 20 GWd/t. FIG. 15 illustrates an example of the use of a combination of Cs137 and S2 i.e., the neutron emission rate of Cm-242. This combination would not be preferable for high exposure, for example higher than 20 GWd/t, because of the short half-life of Cm-242. FIG. 16 illustrates an example of the use of a combination of Cs137 and SR, i.e., the neutron emission rate of nuclides except Cm-242 and Cm-244. It can be understood from FIG. 11 that it can be used for practical applications only for low exposure, for example lower than about 7 GWd/t, because it is not possible to separate S4 and SR. FIG. 17 illustrates examples of the use of a combination of Cs134 and Cs137 and a combination of Eu154 and Cs137. Since the half-life of Cs134 is about 2 years, it can suitably be used at or near the exposure of 20 GWd/t that corresponds to an irradiation period of about 2 years. Eu-154 has a long half-life and a part of Eu-154 becomes Eu-155 because of a large neutron capture cross section as the exposure increases. So, the increase of Eu154 is saturated and some corrections may be required to apply Eu154 to a high exposure. FIG. 18 illustrates an example of the use of Ce144 in place of Cs137 in FIG. 15. A preferable cooling period is not so long if the exposure is low. However, the accuracy of measurement of Cs137 may be poor in such a situation. Therefore, it may be advantageous to use Ce144 having a half-life that is not so long because it can be handled relatively easily for gauging. FIG. 19 illustrates an example of the use of Ce144 in place of Cs137 in FIG. 16. As described above, a preferable cooling period is not so long if the exposure is low. However, the accuracy of measurement of Cs137 may be poor in such a situation. Therefore, it may be advantageous to use Ce144 having a half-life that is not so long because it can be handled relatively easily for measurement as in the case of FIG. 18. FIG. 20 illustrates an example of the use of Ce144 in place of Cs137 in FIG. 17. Eu154 is excluded from this example because accuracy of measurement may be significantly poor if the exposure is low and the cooling time is short. FIGS. 21 through 25 show dependencies of the values corresponding to the void fraction relative to zero void fraction (az/a0) on the exposure. It can be seen that the sensitivity to void fraction is high if the value is high. From a viewpoint of dependency on the exposure, horizontally flat curves mean that no information is required on the exposure for the application. On the other hand, if the curves wind mildly, it means that only rough estimation of the exposure is necessary. If the curve wind sharply, it means that the exposure needs to be estimated with accuracy. FIG. 21 illustrates an example of the use of S4R. It shows that the sensitivity to void fraction is appropriate if the exposure is higher than about 18 GWd/t and very rough information on the exposure is required in and near the range of 18 through 30 GWd/t. FIG. 22 illustrates an example of the use of S4. Considering the fact that S4 may be hidden by SR and hence cannot be utilized if the exposure is low, it makes little difference with the use of S4R. FIG. 23 illustrates an example of the use of S2. It shows that the exposure is required to be accurate to a certain extent because the exposure is required to be smaller than 20 GWd/t. FIG. 24 illustrates an example of the use of SR. If the exposure is low, S2 and SR can be separated from each other by two or more neutron measurement sessions for different cooling periods. The value of S4 is small and hence can be corrected by calculations. So SR is convenient for measurement. However, the use of S4 is accompanied by a problem of low sensitivity relative to voids. No accurate information on the exposure is required. FIG. 25 illustrates an example of the use of Cs134. The sensitivity of the exposure dependency is generally low, and only rough information is required for the exposure. The use of Cs134 is accompanied by a problem of low sensitivity. Some embodiments of the present invention are described below. Since background of the procedures employed is described in detail above, the following description of the embodiments are focused on the procedures. FIG. 1 is a flow chart of an axial void fraction distribution measurement method according to the first embodiment of the present invention. In this embodiment, as the first intensity Az, a neutron emission rate Sz and S0 expressed as Sz=az×Eα or S0=a0×Eα is used. And as the second intensity Bz and B0, gamma ray intensity γz and γ0 expressed as γz=gz×Eα or γ0=g0×Eα is used. Then, the ratio (az/a0) is expressed as;(az/a0)=(Sz/S0)(γ0/γz)α(gz/g0)α. Based on the value of (az/a0) and the relationship between (az/a0) and a void fraction evaluated otherwise, a void fraction of the fuel is evaluated. Also a correlation function representing a relationship between (az/a0) and a void fraction is derived from computations. The void fraction experienced at each axial position of the fuel assembly, i.e., the void fraction distribution is obtained from this correlation function and (az/a0) derived from the first and second intensities of the radioactive rays. Usually the proportionality constant gz does not practically vary at different axial positions of a fuel assembly (gz/g0=1). However, it may slightly vary depending on the fuel design. For such a case, it is necessary to correct the gamma ray shielding effect of the fuel rod by calculations. The value of (Sz/S0) can be determined from the value of (φz/φ0). To be rigorous, the value of φz/φ0) has to be corrected in terms of the contribution to the axial variation of the neutron multiplication factor. However, the value is rather small relative to (φz/φ0−1) in a BWR fuel assembly if the contribution is disregarded, and the variation in the multiplication factor is about 10 to 20% thereof. Therefore, no problem actually arises if (Sz/S0) is approximated as (φz/φ0). Generally, the active part of a BWR fuel assembly is equally divided into 24 nodes and each node is referred to as node 1, 2, 3, . . . , 24 from the lower end (the upstream of cooling water). It is known that the void fraction is practically equal to 0% in nodes 2 and 3, and 70 to 75% in nodes 23 and 24 near the upper end. The shape of the void fraction distribution in the vertical direction of the core in the reactor varies depending on the operation of the reactor or as the exposure increases. The average void fraction distribution of the entire irradiation period is evaluated if the half lives of the radio activities to be measured are long. On the other hand, the void fraction distribution shortly before the end of the operation of the reactor is evaluated if the half lives of the radio activities to be measured are short. In this embodiment, if S4R and Cs137 are used for a measurement of an irradiated fuel assembly, since both of them have a long half-life, the average void fraction distribution of the entire burning period is measured. FIG. 2 is a flow chart of an axial void fraction distribution evaluation method according to the second embodiment of the present invention. This embodiment is different from the first embodiment in a point that the values of (φz/φ0) are corrected for the contribution of the axial variation of the neutron multiplication factor by a standard neutron multiplication factor distribution obtained from the design calculations. As described above, the effect is rather small relative to (φz/φ0−1) in a BWR fuel assembly so that no problem arises even if the value obtained by the design calculation is used, because the effect of the neutron multiplication factor is about 10 to 20%. It is well known that the neutron flux or neutron counting rate (φ) has a relationship of φ=c S/(1−k), where k is a neutron multiplication factor, S is a neutron emission rate, and c is a proportionality constant. This relation is used in this embodiment. The proportionality constant (c) can be determined for example by using a fixed source calculation method that solves a neutron transport diffusion equations with a given neutron source to determine a neutron flux at a target position. A neutron flux for an imaginary condition of no neutron multiplication (φNM: No Multiplication) and a neutron flux for an actual condition of a certain neutron multiplication (φM: Multiplication) are evaluated and the neutron multiplication factor k by using the ratio of the both neutron fluxes. From the neutron flux φ, the neutron multiplication factor k and the relationship φ=c S/(1−k), the proportionality constant c is derived. S is a know value because it is an input of a calculation of the fixed source calculation method. FIG. 3 is a flow chart of an axial void fraction distribution evaluation method according to the third embodiment of the present invention. A difference of this embodiment from the second embodiment is that (φz/φ0) is corrected for the contribution of an axial variation of the neutron multiplication factor based on a fuel assembly averaged exposure that seems to be available most easily from the operator of the reactor. This method of determining a neutron multiplication factor (k) comprising the steps of correcting an estimated void fraction distribution by measured values repeatedly, determining the infinite multiplication factor (k∞) that depends on the exposure and corresponds to the void fraction, and then determining k. The procedure of this embodiment is described in detail below. An axial neutron flux (or the axial neutron counting rate) distribution and a gamma ray intensity distribution, as well as (γ0/γz)α can be determined as described for the first and the second embodiments. It is not so simple to determine (Sz/S0) from (φz/φ0). But, it has been found that the factor for correcting the neutron multiplication factor variation can be determined by repeated calculations (can be converged by repeated calculations) because the factor for correcting the neutron multiplication factor variation is rather small relative to (φz/φ0−1). An axial exposure distribution (E) can be evaluated by normalizing the average value of axial gamma ray intensity distribution (Cs137 distribution in particular) to the average exposure of the fuel assembly reported by the operator of the reactor (“operator-declared average exposure”). The neutron multiplication factor (k) can be obtained by multiplying the infinite multiplication factor (k∞) by a computationally evaluated constant (Fk). However, the value of (k∞) is influenced by the cooling water void fraction during irradiation. Before describing the procedures illustrated in FIG. 3, (k∞) and (Fk) are described. Because it might be confusing that no void is found at room temperature in this embodiment but certain voids exist in and around the fuel assembly at high temperature under high pressure during irradiation. At a condition of high temperature and high pressure, exposure calculations of a fuel assembly are performed, and changes of the composition of fuel are evaluated. At a condition of room temperature, neutron spectrum calculations are performed by using the changed composition of the fuel so as to evaluate a neutron multiplication factor, and to evaluate group constants to be used for neutron transport diffusion calculation. For accurate design calculations, neutron spectrum calculation may be conducted as the exposure increases. For a simplified calculation, no spectrum calculation may be conducted A void fraction is given for high temperature under high pressure to calculate as the exposure increases. And, the updated composition of the fuel as a result of irradiation is evaluated. Then, the neutron spectrum is calculated by using the updated fuel composition as a result of irradiation at room temperature (also referred as low temperature or cold mode) to evaluate the infinite multiplication factor and also the group constants to be used for neutron transport diffusion calculations. Neutron transport diffusion calculations are conducted for the system to be measured by a fixed source calculation method to evaluate the neutron multiplication factor (k) at a certain position. While this value may be quantitatively different slightly from the effective multiplication factor used normally for the case of a BWR fuel assembly placed in water, but the difference is negligible in this embodiment. The inventors have found that the ratio (Fk) of the neutron multiplication factor and the infinite multiplication factor is not affected by the fuel composition if the position of measurement is more than 2 to 3 cm farther from the fuel assembly. “The infinite multiplication factor of the system to be measured is affected by the void fraction” means that the infinite multiplication factor changes because the composition is changed by the void fraction in cooling water during irradiation. The procedure of this embodiment is described below. The infinite multiplication factor (k∞) decreases substantially linearly as the exposure (E) increases. But the neutron multiplication effect appears due to build-up of plutonium at higher exposure. Therefore, the infinite multiplication factor (k∞) is approximated by the quadratic expression of exposure, i.e.;k∞=J0−J1×E+J2×E2 However, in many cases, it can be well approximated by a linear expression. The constant (J0) and the coefficients (J1, J2) used for the quadratic approximation depend on the void fraction and the characteristics of the dependency can be evaluated by calculations for the different void fractions (ratios). For a fuel assembly which has the different initial enrichments for upper and lower part, the constant and the coefficients should include effects of enrichment. Thus, the correlation of the infinite multiplication factor and the exposure is expressed as a function of the void fraction by calculations. Then, the axial distribution of infinite multiplication factor (k∞i) is calculated based on the exposure and the assumed void fraction (where i=1, the starting point of repeated calculations). And the axial neutron multiplication factor (ki) is evaluated by using (Fk). Then, the value of (Sz/S0)i is calculated from the multiplication factor and the measured value of (φ0/φ1). Also the axial distribution of (az/a0)i is evaluated. Finally, the distribution of the corrected axial void fraction (vi, i=2) evaluated by revising the assumed void fraction (i=1) is evaluated from the relationship between the void fraction and the calculated ratio of (az/a0)i. The corrected distribution is compared with the distribution that is not revised. And it is judged whether the axial distributions agree with each other within an allowable margin. If they do not agree, the calculation is performed again based on the revised distributions. The value obtained as a result of convergence is defined as the axial void distribution. FIG. 4 is a flow chart of an axial void fraction distribution method according to the fourth embodiment of the present invention. A difference of this embodiment from the third embodiment is that (φ0/φ1) is corrected for the contribution of the axial variation of the neutron multiplication factor based on the exposure E0 at the reference position that is evaluated by using the neutron emission rate of the fuel. The exposure is evaluated with a method substantially identical to the neutron emission rate technique described in Japanese Patent Application Publication No. Sho 61-262689. FIG. 5 illustrates a procedure of the neutron emission rate technique according to this embodiment. In the neutron emission rate technique, the neutron flux (φ) at a lateral surface of the fuel assembly is measured. Because an absolute value of the neutron flux is required for this method, a neutron detector is calibrated by a gold foil activation method or some other method. The neutron flux or neutron counting rate (φ) has a relationship with the neutron multiplication factor (k) and the neutron emission rate (S) that is expressed as φ=c×S/(1−k), where (c) is a proportionality constant. The proportionality constant (c) can be determined by calculations as described above. At the beginning (i=1) of the repeated calculation, an initial assumption of the neutron multiplication factor kz is given, and overall neutron emission rate Si is evaluated as Si=φi(1−ki)/c. If it is necessary to correct for the contribution of the neutron emission rate of Cm-242, the S2 component is eliminated by calculations or by two or more measurements for different cooling periods, utilizing the difference of half-lives. In short, the value of (S4R)i is calculated as (S4R)i=Si−S2. Then, (Ei) is determined by using the calibration curve that correlates S4R for the void fraction of 0%. The calibration curve can be calculated from this value and the exposure. Further, k∞i is evaluated by using the calibration curves that correlates the calculated k∞ and E, or correlates S4R and E. Then, the revised neutron multiplication factor ki is evaluated from this value by using the conversion factor (Fk) for the system to be measured. If the difference between the revised neutron multiplication factor and the corresponding unrevised value is found within an allowable range, the revised neutron multiplication factor is a final one. If the difference is greater than the range, the calculation is repeated by using the initial or last revised values, and the values obtained as a result of convergence is defined as the neutron multiplication factor. Repeated calculations are required because a transcendental function is practically included in this method and convergence can be achieved by about 3 to 5 times repeated calculation. This neutron emission rate technique can evaluate a plutonium concentration, an infinite multiplication factor and the neutron multiplication factor, as well as exposure in the same calculations. FIG. 6 is a flow chart of an axial void fraction distribution evaluation method according to the fifth embodiment of the present invention. In this embodiment, only gamma ray spectrum measurement is performed to achieve the objective, so that it is called a gamma ray spectrum analysis method. This embodiment utilizes the conventional measurement techniques such as a Ge semiconductor detector. A gamma ray collimator is arranged close to a lateral surface of the irradiated fuel assembly and gamma rays are led through it. Then, the gamma ray spectrum is measured by a semiconductor detector. By this measurement, the intensity of target gamma rays emitted from Cs-137, Cs-134, Ce-144, Eu-154 and so on are measured. While it is not always true that the intensity of these gamma rays can be measured with accuracy, it is well known to those skilled in the art what condition is required to measure with accuracy. In the gamma ray spectrum analysis method, the combinations of Cs134 with Cs137, Eu154 with Cs137, and Cs134 with Ce144 are expressed as their ratios. The axial distributions (Gz/G0) are evaluated. Then, the axial void fraction distribution is evaluated based on the calibration curve evaluated computationally that correlates (Gz/G0) and the void fraction. FIG. 7 is a flow chart of an axial void fraction distribution method according to the sixth embodiment of the present invention. This embodiment utilizes the difference of gradient between S2 and S4R in FIGS. 10 and 11. It can be applied if the exposure is very low, for example lower than about 7 to 10 GWd/t. This embodiment is a method to evaluate an axial void fraction distribution only by measurement of a neutron emission rate, and is preferable for low exposure. The ratio of the neutron emission rate (S2) of Cm242 and the neutron emission rate (S4R) of all the other nuclides is evaluated. Then the axial void fraction distribution is evaluated by using the calibration curve for the calculated axial distribution of the ratio. This procedure is similar to the procedures described as the fifth embodiment. More specifically, the emission rates of neutrons emitted from the nuclides in the fuel assembly irradiated in the reactor caused by neutron capture reaction are measured at least twice for cooling periods with a certain interval. Then, the first neutron emission rate (S2) of neutrons emitted from Cm-242 and the second neutron emission rate (S4R) Of neutrons emitted from the other nuclides are discriminated by using the difference of half-lives. The first neutron emission rate (Az=S2) is expressed as it is proportional to an exponent of the exposure (E), expressed as Az=az×Eα, where az is a proportionality constant. Additionally, the measured second neutron emission rate (Bz=S4R) is regarded proportional to an exponent of the exposure (E), expressed as Bz=bz×Eβ, where b is a proportionality constant. Then, the both sides of the equation of the first neutron emission rate are raised to the power of β and the both sides of the equation of the second neutron emission rate are raised to the power of α. Subsequently, the ratio of the two raised formulae is derived to eliminate the exposure term. The ratio is divided by the equation for the reference position. The following equation is derived.(az/a0)β(b0/bz)α=(Az/A0)β(B0/Bz)α. Numeral “0” denotes that the value is for the reference position. Then, the axial void fraction distribution is evaluated by using the calibration curve, and measured value of (Az/A0) and (B0/Bz), utilizing the characteristic that (az/a0)β(b0/bz)α depends on the axial void fraction distribution in the nuclear reactor. It is hard to apply this embodiment to a case that a contribution of Cm-242 to the neutron emission rate (S4R) is low. This is because that the dependency of (az/a0) and that of (b0/bz) on the void fraction distribution are inverse relative to each other resulting in elimination of the dependency of (az/a0)β(b0/bz)α on the void fraction distribution. So, this embodiment is applicable only to the low exposure fuel. Additionally, the use of Cm-242 makes it hard to apply this embodiment to a case that a cooling period of the fuel is longer than about two years. And the large sensitivity on the exposure makes the exposure data necessary. However, this method has a large merit because it requires measurements of only neutrons. FIG. 8 is a flow chart of a neutron multiplication factor evaluation method according to the seventh embodiment of the present invention. This embodiment is a method to evaluate a neutron multiplication factor of a fuel assembly before it is contained in a fuel assembly containing apparatus. A neutron multiplication factor at about ⅔ to ¾ from the lower end of the axial fuel effective part of an irradiated (or discharged) BWR fuel assembly tends to be higher. When the fuel assembly is contained in a container apparatus, this part decreases a sub-criticality (or makes criticality close to unity). Therefore, without a measurement of the multiplication factor of this part of the fuel, it is necessary to take a large design margin so as to secure the criticality safety. By this embodiment, measuring the multiplication factor of each fuel assembly actually makes it possible to reduce the design margin. In this embodiment, neutron fluxes are measured at a reference point and at a sub-criticality monitoring point of an irradiated fuel assembly. The ratio of the neutron multiplication factor (k0) and the neutron flux or neutron counting rate (φ0/φz) are evaluated for the reference point by using the neutron emission rate technique described above. Also, the gross gamma ray intensity ratio (γgz/γg0) is measured. In addition, the correlation of the gross gamma ray intensity relative distribution and the exposure relative distribution is calculated separately (the difference caused by the difference of distribution pattern is normally less than 10%). The gross gamma ray intensity ratio (γgz/γg0) is reduced to the ratio relative to exposure by using the correlation, and (Ez/E0)α is evaluated by using the calculated value of α that is a value of exponent to be raised to the exposure (E). If a typical void fraction distribution is evaluated by measuring some fuel assemblies of the same type of assembly by the method described in other embodiments or other means, it is possible to define the void fraction at the sub-criticality monitoring point. With this void fraction, the value of (az/a0) is evaluated reversely based on the correlation of (az/a0) and the void fraction, and the neutron multiplication factor at the sub-criticality monitoring point is evaluated by the following equation before the fuel assembly is contained in a fuel assembly containing apparatus.k=1−(1−k0)(φ0/φz)(Ez/E0)α(az/a0) The gross gamma ray measurement method is already established by the inventors of the present invention and adopted in the commercial reprocessing plants. Besides, it is known that similar phenomena occur in an irradiated (discharged) pressurized water reactor (PWR) fuel assembly due to the water temperature, the water density, the control rod effect and other factors but voids so that the method of this embodiment can be applied by taking their influences into the factor of (az/a0). Of course, a gamma ray spectrum analysis can be applied to this embodiment in place of the gross gamma ray measurement. But the gross gamma ray measurement may be more practical than the gamma ray spectrum analysis even if the accuracy of the gross gamma ray measurement might be lower, because the gamma ray spectrum analysis requires a larger devices and more complicated data processing in general. Numerous modifications and variation of the present invention are possible in light of the above teachings. It is, therefore, to be understood that, within the scope of the appended claims, the present invention can be practiced in a manner other than as specifically described herein.
040653510	claims	1. Apparatus for injecting charged particles into a confining magnetic field, comprising: a. vacuum tight housing means having an endless z first circular axis, and entrance ports and exit ports forming solid collecting walls out of the line of sight with the axis but communicating therewith through poloidal apertures in the sides of the housing means; b. vacuum pump means communicating with the collecting walls; c. means for forming a confining, toroidal, magnetic field having helical field lines forming inner and outer concentric magnetic surfaces centered on an endless magnetic second axis concentric with the first axis in the housing means; d. means for forming an equilibrium, neutral, thermal, target plasma column of disassociated electrons, tritons, and deuterium ions that are confined by the magnetic field in the column in the container means along the magnetic second axis at an elevated temperature that causes the ions to diffuse by collisions outwardly away from the magnetic axis toward the vacuum container means; e. means for continuously injecting neutral atomic beams of deuterium and tritium through the entrance ports along parallel trajectories in the plane of the circular axis at energies above the average energy of the plasma sufficient to penetrate the sides of the plasma column to a depth for producing thermal electrons and deuterons and tritons at energies of at least 10 keV above the average energy of the confined target plasma column, the injection forming fast, ordered, high energy, high velocity counterstreaming deuterons and tritons that drift along the helical field lines, and circulate around the length of the axis, these counterstreaming deuterons and tritons forming beams having distinct ion velocity distributions that are oppositely displaced in velocity along the magnetic axis, while the helical field lines provide strong restoring forces that maintain the directedness of the fast deuterons and tritons along the magnetic axis, some of said fast deuterons and tritons slowing down to the average energy of the confined target plasma column thereby to maintain the plasma density and temperature by balancing the diffusion rate; and f. divertor means communicating with the field lines of the outer magnetic surfaces through the poloidal apertures for collecting diffusing plasma particles through the exit ports from the outside of the plasma column and burying them in the vacuum pump means so as to maintain a high counterstreaming deuteron and triton number density. a. Magnetically confining a neutral, toroidal plasma column of ions and electrons having a thermal temperature, density, volume, average thermal energy and average confinement time along an endless magnetic axis in a tokamak vacuum container means containing a toroidal magnetic field having concentric magnetic field lines of force forming a magnetic container that is concentric with the magnetic axis and capable of confining the plasma for a sufficient period of time to produce a thermal ion diffusion loss rate and a thermal electron diffusion loss rate; b. injecting ordered, neutral, atomic beams having fast ordered ions and orbital electrons at densities that are directed into the confined plasma with trajectories that are generally tangent to the magnetic field lines and azimithally along the magnetic axis in the same and the opposite direction to the direction of the magnetic axis at an energy that is greater than the average thermal energy of the confined plasma, the neutral beams interacting with the confined thermal plasma ions to inject fast ions and thermal electrons into the plasma, the fast ions forming oppositely circulating, counterstreaming ion beams having directed, distinctly ordered, ion velocity distributions and associated beam currents that are oppositely displaced in velocity along the magnetic axis, the latter ion beams injecting thermal ions into the confined plasma due to the slowing down of the ion beams in the confined plasma; c. the magnetic confinement maintaining the aforesaid thermal electron injection in balance with the thermal electron diffusion to maintain a thermal electron density in the confined plasma; d. the magnetic confinement of the plasma also maintaining the directedness of the counterstreaming ion beams until the ions therein slow down to the average thermal energy of the confined plasma; and e. removing the diffusing thermal electrons and the ions that slow down to the average thermal energy of the confined plasma, said removal being at least as fast as the average time it takes for the thermal ions to slow down to the average thermal energy of the confined plasma so as to maintain a high counterstreaming ion number density in the counterstreaming ion beams that is at least as great as the confined thermal ion density, said removal also maintaining the sum of the aforesaid counterstreaming ion number density and the confined thermal ion density substantially in balance with the confined electron density in the confined plasma so that the counterstreaming ion beams continuously produce a large number of head-on collisions between the counterstreaming ions all along the magnetic axis. a. producing a thermal plasma column in a vacuum tight housing; b. confining the plasma along a circular axis in a toroidal magnetic field in the vacuum tight housing; c. injecting equal momentum, neutral atomic beams through the sides of the housing along parallel trajectories in the plane of the circular axis at energies above the average plasma energy sufficient to penetrate the outside of the plasma column to a depth for producing counterstreaming ion beams that slow down to the average energy of the thermal plasma and drift across the toroidal magnetic field in a direction away from the axis by diffusion; d. continuing the injection for a period sufficient to stack the counterstreaming beams for many orbits around the length of the axis at a rate in balance with the diffusion rate; and e. removing the thermal plasma particles that drift across the magnetic field by diffusion by magnetically diverting them through poloidal apertures in the sides of the housing selectively to neutralize the particles outside of the plasma column and out of a line of sight therewith for maintaining high counterstreaming ion beam densities. 2. Method for injecting charges particles into a tokamak confining magnetic field, comprising: 3. The method of claim 2 in which opposite deuterium and tritium neutral atomic beams are injected into the confining magnetic field to produce counterstreaming deuteron and triton beams that are tied to the magnetic field lines of force along trajectories that spiral helically along axes that are centered on the respective field lines. 4. The method of claim 3 in which the confined plasma has a weight of at least one gram at an electron temperature T.sub.e .gtoreq. 1 keV and a density n.sub.e .gtoreq. 10.sup.12 cm.sup.-3. 5. The method of claim 4 in which the removal of the thermal ions is provided by a divertor for selectively decreasing the number of relatively cold ions confined in the magnetic field all along the magnetic axis. 6. The method of claim 5 in which a plasma current is provided along the endless magnetic axis for twisting the field lines into helixes that helically twist around the magnetic axis. 7. The method of claim 6 in which the input and output are balanced to maintain a counterstreaming ion number density that is 50% of the electron number density of the confined plasma. 8. The method of claim 7 in which counterstreaming deuteron and triton beams collide to produce neutrons in the confined plasma along the length of the magnetic axis. 9. The method of injecting charged particles into a confining magnetic field, comprising the steps of evacuating a vacuum container means having an endless equilibrium axis to a vacuum of at least 3 .times. 10.sup.-5 torr; forming a magnetic field having concentric magnetic surfaces along a magnetic axis and field lines concentric with the magnetic axis that are co-axial with the equilibrium axis in the vacuum container means; admitting at least one gram of a gas containing tritium and deuterium into the vacuum container means to a density of at least 10.sup.12 particles/cm.sup.3, ohmically heating the gas to a temperature of at least 1.0 keV to produce an equilibrium target plasma column containing thermal ions and electrons that are confined at a number density of at least 10.sup.12 particles/cm.sup.3 in the magnetic field in the vacuum container means along the magnetic axis for at least 1 msec, injecting high energy, high velocity, ordered neutral atomic deuterium and tritium beams into the target plasma in the same and the opposite direction to the direction of the magnetic axis at energies of at least 10 keV greater than the energy of the confined plasma and at currents of at least 1.mu.A to produce ordered, high energy, high velocity counterstreaming deuterons and tritons with like high energies and velocities, the counterstreaming deuterons and tritons having number densities higher than the density of the injected beams and distinct ion velocity distributions that are oppositely displaced in velocity along the magnetic axis while the magnetic field maintains the directedness of the counterstreaming deuterons and tritons along the magnetic axis until they slow down to the average energy of the confined equilibrium target plasma column, and diverting thermal electrons and ions away from the outside of the plasma column into a poloidal divertor for burial therein. 10. Method for injecting charged particles into a confining magnetic field, comprising the steps of:
	description	The invention is related to nuclear fuel assembly channel fasteners. More specifically, the invention relates to a boiling water reactor fuel channel fastener which restricts rotation during torquing and detorquing. Fuel channel fasteners have been used for many years in boiling water reactor fuel assemblies. The purpose of the fastener is to mechanically attach the external fuel channel to the fuel assembly, so that under operating conditions, the reactor coolant is restrained around each fuel assembly. The fastener utilizes a spring to separate the fuel channel from other fuel assemblies in the proximity of the fuel channel. The current designs of fuel channel fasteners, therefore, provide a solid stop between adjacent channels. Previous fuel channel fastener designs have several significant shortcomings. The bolt in these designs was moved toward a more inward position, as compared to the exterior edge of the fuel channel. The placement of the bolt in this arrangement provided a tendency for the entire fastener to rotate during tightening of the bolt. As manufacturers have modified the bolt position of fuel channel fasteners from a position close to the edge of the fuel channel further toward an inside part of the fuel assembly, the additional distance from the bolt to the external parts of the fastener decreases the rotational resistance due to the additional moment arm. The rotation of the bolt caused the body of the fastener to rotate on the fuel assembly channel, thus allowing the bottom edge of the fastener body to protrude from the exterior of the channel wall. This rotation thereby allows an additional contact edge which may lead to premature failure of the fuel channel fastener as there is an additional contact surface for impact. The rotation also allows the end of the fuel channel spring to extend beyond the protective configuration of the body of the fastener. When the fastener body or spring extends outward, they are more likely to be damaged by interacting with in-reactor blade guides, fuel storage racks, other fasteners, or reactor components during, for example, fuel assembly movement. Industry experience has shown that fuel channel fasteners can prematurely fail using these designs. There is a need to provide a fuel channel fastener that will allow for adequate seating of the fuel channel fastener to the fuel assembly. There is also a need to provide a fuel channel fastener that will be rugged for anticipated operating and accident conditions for a nuclear reactor. There is a still further need to provide a fuel channel fastener that will be less susceptible to damage compared to current fuel channel fastener designs. It is an objective of the current invention to provide a fuel channel fastener that will allow for adequate seating against the fuel channel fastener to the fuel channel. It is also an objective of the current invention to provide a fuel channel fastener that will be rugged for anticipated operating and accident conditions for a nuclear reactor, while maintaining adequate seating between the fuel channel fastener and the fuel channel. It is a further objective of the current invention to provide a fuel channel fastener that will be less susceptible to damage compared to current fuel channel fastener designs. The objectives of the current invention are achieved as described and illustrated. The invention provides a fuel channel fastener. The fuel channel fastener comprises a washer, a body with a first hole, a spring with a second hole and with two perpendicular spring members configured to extend away from the body down sides of a fuel assembly fuel channel, wherein the spring is configured adjacent to the body such that the first hole and the second hole are concentric, a bolt inserted through the washer, the second hole and the first hole, at least two anti-rotation supports attached to the body, and at least two anti-rotation pads connected to the body, wherein the anti-rotation supports are configured to provide resistance to fastener rotation. Referring to FIG. 1, a fuel channel fastener 10 for a boiling water reactor fuel assembly is illustrated. The overall length of the fuel channel fastener 10 for the boiling water reactor fuel assembly may be, for example, approximately 5 inches (12.7 cm). The fuel channel fastener 10 is comprised of a bolt 26 that extends through lock washer 14, a body 38 and a spring 22. The fuel channel fastener 10 establishes a connection between a fuel channel of a nuclear fuel assembly and the internal structures of the fuel assembly. This connection allows the fuel channel of the fuel assembly to provide a rigid structure around the fuel assemblies. The fuel channel also allows for the channeling of moderator, such as reactor water, through the internal components of the fuel assembly including the nuclear fuel rods. The body 38 has a body first end 34 and a body second end 32, wherein the body 38 is configured to fit over an edge of a fuel channel of a nuclear fuel assembly. The body 38 may be inserted near a corner end of a fuel assembly fuel channel such that a surface of the body 38 contacts the exterior surface of the fuel channel. The body 38 is made of a metallic material, such as stainless steel, to account for adequate corrosion resistance. As illustrated, the bolt 26 of the current invention is torqued to about 6 pound feet. The body first end 34 may be configured at a top with a chamfered edge to provide an alignment feature for the spring form. The body second end 32 may be configured with a chamfered end as illustrated to provide a leading edge for interaction. The body 38 is configured with varying uniform thickness to provide adequate strength and form for interfacing with spring 22 and other channel fasteners. The bolt 26 is inserted through a washer 14. The washer 14 provides an arrangement to accept the bolt head 12 compressive forces and transfer those compressive forces to the spring 22 and the body 38. The washer 14 may be made of materials, such as stainless steel, for example, to limit corrosion or galvanic reaction. The washer 14 can be configured as a locking washer to prevent unintended removal or loosening of the bolt. The washer 14 can have a low profile such that vertical protrusion of the bolt 26 above the top of the fuel assembly is minimized. The spring 22 is also compressed by the combination of the bolt head 12 connection to the fuel assembly. The spring 22 is configured with a spring first end 16 which extends around the fuel channel to a spring second end 20. The spring 22 has an inflection point 24 such that the spring 22 extends away from the body 38 up to the inflection point 24. The spring 22 is configured with a bend 18 to allow the spring 22 to extend from a horizontal section to a vertical section. The spring 22 may be made of a metallic material which provides a sufficient spring constant so that adjacent fuel is separated during normal reactor operation. The spring 22 may be made of nickel Alloy 718, for example. The bolt 26 is arranged in a configuration such that the bolt 26 cannot be removed from the body 38 and the spring 22 in a detorqued condition. The body 38 may be staked during manufacturing such that the removal of the bolt 26 from the body 38 and the spring 22 is prevented due to bolt material exceeding bolt hole 52 dimensions. The staking procedure eliminates concerns for loose parts, thereby encouraging foreign material exclusion from sensitive areas of the nuclear reactor. Referring to FIG. 2, a top view of the fuel channel fastener 10 of FIG. 1 is illustrated. Bolt 26 and washer 14 are not illustrated for clarity. The fuel channel fastener 10 has a first anti-rotation support 40 positioned, as illustrated in FIG. 2, on an exterior portion of the fastener 10. The first anti-rotation support 40 is configured to engage the exterior of the fuel channel of the fuel assembly, such that when the bolt 26 is rotated in a clockwise direction, the resulting overall torque placed on the fuel channel fastener 10 will cause the first anti-rotation support 40 to contact the external surface of the fuel channel. The resulting contact between the first anti-rotation support 40 and the external surface of the fuel channel restricts movement of the fastener during torquing thereby allowing tightening of the fastener without fastener rotation. A second anti-rotation support 42 is located on an opposite corner of the fastener 10. The second anti-rotation support 42, similar to the first anti-rotation support 40, contacts the external surface of the fuel channel of the fuel assembly, such that when the bolt 26 is detorqued, the second anti-rotation support 42 contacts the fuel channel, thereby preventing rotation of the fuel channel fastener 10. As an illustrative example, the supports 40, 42 may be approximately 0.38 millimeters in thickness. The bolt hole 52 is positioned such that it is located away from the edge of the fastener and such that the hole 52 is positioned along an axis drawn from the joints formed from the intersection of anti-rotation support 42 and back adjacent edge 50 with anti-rotation support 40 and back adjacent edge 54. Other configurations of hole position may be chosen, wherein the bolt hole 52 is moved respectively toward the front edge 48 or the back edge 46. The configuration presented, therefore, is merely illustrative of the possible configurations, including bolt hole 52 positions near the edge of the back edge 46, for example. The first anti-rotation pad 60 is in between spring 22 and the first anti-rotation support 40. The second anti-rotation pad 44 is likewise in between spring 22 and the second anti-rotation support 42. The first anti-rotation pad 60 and second anti-rotation pad 44 contact the spring 22 during torquing and detorquing of the bolt 26 to prevent the spring from rotating. FIG. 3 presents a side view of a fuel channel fastener 10 installed on a fuel channel 64 in contact with another fuel channel 68. The fuel channel fastener 10 positioned on each fuel channel 64 has a spring 22 which provides a bearing surface between the individual fuel channels 64, 68. The springs 22 of the individual fuel channel fasteners 10 contact other adjacent springs 22. The springs 22 are configured to deflect a sufficient amount such that a space 66 is always maintained between the respective fuel channels 64, 68. Referring to FIG. 4, a top view of the fuel channel fastener 10 of the present invention is illustrated. A center point 72 of the hole 52 is located along a neutral axis 70 defined by a first body connection end point 74 and a second body connection end point 76. Other configurations are possible wherein the center point 72 of the hole 52 may be closer or further from the front edge 48. Referring to FIG. 5, a side perspective view of the fuel channel fastener 10 with the individual components separated is illustrated. The bolt 26 is inserted through the washer 14 and the spring 22 into the bolt hole 52. The spring 22 is placed over the body 38 and into depressions 78 located on the body 38. The present invention provides a fuel channel fastener 10 which provides for a connection between the fuel channel and a fuel assembly. The fuel channel fastener 10 provided in the current invention prevents rotation of the fuel channel fastener 10 during tightening and loosening. The fuel assembly channel fastener 10 of the current invention provides a more resistant structure for bending as compared to other designs. The anti-rotation capabilities of the fuel assembly channel fastener 10 are established without modifying fuel channels of fuel assemblies and without detrimental impact to overall core flow characteristics. In the foregoing specification, the invention has been described with reference to specific exemplary embodiments, thereof. It will be evident that modifications and changes may be made thereunto without departing from the broader spirit and scope of the invention as set forth in the appended claims. The specification and drawings are accordingly to be regarded in an illustrative rather than a restrictive sense.
	abstract	An absorber rod for a nuclear reactor having a rod cladding defining an internal volume, the rod cladding having an upper end and a lower end, an upper end fitting positioned in the upper end of the rod cladding, a rod internal arrangement configured in the internal volume of the rod cladding, an absorber rod lower end cap positioned at the lower end of the rod cladding, the lower end cap having an upper surface and a lower surface, a stack support with a stack support upper end and a stack support lower end, the stack support lower end placed in contact with the upper surface of the lower end cap, the stack support upper end configured to support the rod internal arrangement, and an annulus of material configured around the stack support and contacting the upper surface of the lower end cap.
047754958	claims	1. A process for disposing of a radioactive liquid waste, which comprises adding an alkaline earth metal hydroxide to a radioactive liquid waste containing sodium sulfate to convert said sodium sulfate into an insoluble alkaline earth metal salt thereof with the formation of sodium hydroxide as a by-product, adding a silicon oxide compound to the sodium hydroxide by-product to form water glass (sodium silicate), and solidifying said insoluble alkaline earth metal salt using said water glass. 2. A process for disposing of a radioactive liquid waste according to claim 1, wherein the radioactive liquid waste contains sodium sulfate as the main component. 3. A process for disposing of a radioactive liquid waste according to claim 2, wherein the alkaline earth metal hydroxide is barium hydroxide. 4. A process for disposing of a radioactive liquid waste, which comprises adding an alkaline earth metal hydroxide to a radioactive liquid waste containing sodium sulfate to form an insolubilized solid component and a remaining aqueous solution component containing sodium hydroxide, separating said insolubilized solid component from the aqueous solution component containing sodium hydroxide, adding a silicon oxide compound to the remaining aqueous solution of sodium hydroxide to form water glass, and solidifying said insolubilized solid component with a solidifier including said water glass. 5. A process for disposing of a radioactive liquid waste according to claim 4, wherein a mixture of the radioactive liquid waste and the alkaline earth metal hydroxide is kept at 40.degree. to 80.degree. C. and stirred to insolubilize the sodium sulfate. 6. A process for disposing of a radioactive liquid waste according to claim 5, wherein the mixture of the formed aqueous sodium hydroxide solution and the silicon oxide compound added thereto is stirred at a temperature of about 80.degree. C. to form water glass. 7. A process for disposing of a radioactive liquid waste according to claim 4, wherein the alkaline earth metal hydroxide is barium hydroxide. 8. A process for disposing of a radioactive liquid waste according to claim 4, wherein the radioactive liquid waste contains sodium sulfate as the main component. 9. A process for disposing of a radioactive liquid waste, which comprises the steps of adding an alkaline earth metal hydroxide to a radioactive liquid waste containing sodium sulfate to form a mixture of an insolubilized solid component and an aqueous sodium hydroxide solution; adding a silicon oxide compound to the liquid mixture while stirring the liquid mixture to form water glass mixed with the insolubilized solid compound; evaporating water contained in the liquid mixture comprising the insolubilized solid component and the water glass thus formed thereby to concentrate the liquid mixture; and adding a hardening agent to the concentrated liquid mixture to obtain a waste package. 10. A process for disposing of a radioactive liquid waste according to claim 9, wherein the radioactive liquid waste contains sodium sulfate as the main component. 11. A process for disposing of a radioactive liquid waste according to claim 10 wherein the alkaline earth metal hydroxide is barium hydroxide. 12. A process for disposing of a radioactive liquid waste according to claim 11, wherein the mixture of the radioactive liquid waste and the alkaline earth metal hydroxide is stirred at a temperature in the range of 40.degree. to 80.degree. C. to insolubilize the sodium sulfate. 13. A process for disposing of a radioactive liquid waste according to claim 12, wherein a silicon oxide compound is added to the formed aqueous sodium hydroxide solution and the mixture is stirred at a temperature kept at about 80.degree. C. to form water glass. 14. A process for disposing of a radioactive liquid waste according to claim 9, wherein the mixture comprising the formed water glass and the insolubilized solid component is dried and pulverized and then water and a hardening agent are added thereto to obtain the waste package. 15. A process for disposing of a radioactive liquid waste according to claim 9, wherein the ratio of silicon oxide (SiO.sub.2) to sodium oxide (Na.sub.2 O) in the water glass is in the range of 1 to 4. 16. A process for disposing of a radioactive liquid waste according to claim 15, wherein the ratio of silicon oxide to sodium oxide in the water glass is in the range of 2 to 3. 17. In a process for treating a radioactive liquid waste containing sodium sulfate for disposal comprising adding a silicon oxide compound to the radioactive liquid waste containing sodium sulfate to form water glass, and adding a hardening agent to said water glass to form a solidified body, the improvement which comprises: (1) adding an alkaline earth metal hydroxide to the radioactive liquid waste containing sodium sulfate to form an insoluble alkaline earth metal salt and an aqueous sodium hydroxide solution, (2) adding a silicon oxide compound to the resultant sodium hydroxide solution to form an insoluble alkaline earth metal salt-water glass mixture, and (3) adding a hardening agent to said mixture to form a solidified body. 18. A process for disposing of a radioactive liquid waste according to claim 17, wherein the mixture comprising the formed water glass and the insolubilized solid component is dried, pulverized and pelletized and then water and a hardening agent are added thereto to obtain the waste package. 19. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17 wherein the radioactive liquid waste contains sodium sulfate as the main compound. 20. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17, wherein the alkaline earth metal hydroxide is barium hydroxide. 21. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17, wherein the mixture of the radioactive liquid waste and the alkaline earth metal hydroxide is stirred at a temperature in the range of 40.degree. to 80.degree. C. to insolubilize the sodium sulfate. 22. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17, wherein the mixture comprising the formed water glass and the insolubulized solid component is concentrated. 23. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17, wherein the mixture comprising the formed water glass and the insolubulized solid compound is dried, pulverized, and rewetted. 24. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17, wherein the ratio of silicon oxide (SiO.sub.2) to sodium oxide (Na.sub.2 O) in the water glass is in the range of 1 to 4. 25. A process for treating a radioactive liquid waste containing sodium sulfate for disposal according to claim 17, wherein the ratio of silicon oxide (SiO.sub.2) to sodium oxide (Na.sub.2 O) in the water glass is in the range of 2 to 3. 26. A process for disposing of radioactive liquid waste, which comprises the steps of: adding barium hydroxide to a radioactive liquid waste containing sodium sulfate as the main component and stirring the liquid waste and the barium hydroxide added thereto to form a mixture of an insolubilized solid component of barium sulfate and an aqueous sodium hydroxide solution; adding a silicon oxide compound to the mixture while stirring the mixture to form water glass mixed with the insolubilized solid compound, whereby the mixture is free of the sodium hydroxide; evaporating water contained in the mixture of the insolubilized solid component and the water glass thus formed thereby to concentrate the liquid mixture; and, adding a hardening agent to the concentrated mixture while stirring the mixture and adding water to obtain a waste package.
	description	The invention relates generally to the field of diagnostic imaging and more particularly to apparatus and methods for providing a flexible antiscatter grid to reduce scatter effects in a radiographic image. Scattered radiation presents a particular challenge for radiographic imaging. In some cases, scatter can significantly reduce subject contrast, making it difficult to discern various anatomical features in the radiographic image. Linear grids have been devised in order to help correct this problem. Linear grids are antiscatter devices that are used to improve contrast and to improve the signal to noise ratio in radiographic images. A conventional antiscatter grid typically consists of a series of lead foil strips separated by spacers that are transmissive to x-rays. The spacing of the strips determines the grid frequency, and the height-to-interspacing distance between lead strips determines the grid aspect ratio. Grids can be oriented horizontally or vertically relative to the imaging medium. There are two general types of antiscatter grids: moving (Bucky-Potter configuration) and stationary. For moving type grids, the shadows of the lead strips are blurred out by motion, which can be either reciprocating or unidirectional (single stroke). For stationary grids, the shadows of the lead strips are imposed onto the radiographic image and can be reduced using programmed image processing methods. In general, antiscatter grids, equivalently termed “image contrast antiscatter grids”, are required for most types of “thick” tissue medical imaging procedures; i.e., procedures in which the screen is not located close (within about the thickness of the screen) to body tissue during medical imaging procedures. Image contrast antiscatter grids have been formed in a number of ways. Grids can be formed by laminating together foils of x-ray transparent material, such as aluminum and x-ray absorbing material, such as lead, to form an extended sandwich structure. The simplified schematic of FIG. 1 illustrates a known sandwich structure image contrast antiscatter grid 24 including aluminum foils 26 and lead foils 28 forming an alternating, parallel arrangement. Other methods of forming image contrast antiscatter grids have been described, for example, in U.S. Pat. Nos. 5,581,592 and 5,557,650, which are incorporated herein by reference in their entirety. The various methods proposed for forming antiscatter grids have proved to be cumbersome or unsatisfactory for a number of reasons, including: (i) high cost, due in large part to complex and expensive fabrication; (ii) significant weight, making the grid difficult to position; (iii) grid visibility in the acquired image; [Known image contrast antiscatter grids, such as the image contrast antiscatter grid 24 in FIG. 1, have a relatively coarse structure that produces grid lines in radiographic images. To reduce this problem, for example, the grids can be moved slightly back and forth in a direction approximately perpendicular to the normal (that is, slightly back and forth perpendicular to the direction of the x-rays) to blur the image of the grid lines formed on the receiver. This movement of the grids is known as the “Bucky system.” However, the Bucky system requires the imaging system to include additional components and, thus, increases the cost and complexity of the system.] (iv) incomplete scatter compensation; [Known image contrast antiscatter grids, such as the image contrast antiscatter grid 24 of FIGS. 1 and 2, only remove the Compton-scattered, non-normal (off-z-axis) photons in one dimension (i.e., along either the x-axis or the y-axis). In order to provide two-dimensional photon removal using these grids, two grids, such as two of the image contrast antiscatter grids 24, must be stacked with their respective foils oriented orthogonally with respect to each other. Although the combined use of two grids may improve Compton-scattered photon removal in a second direction, the cost of the imaging system is significantly increased by the added cost of the second grid. Thus, the value of improving the performance of the imaging system by using two image contrast antiscatter grids may not justify the associated added cost and space requirements to achieve the improved performance.] (v) fragile structure, readily damaged by mishandling; (vi) rigidity, making the grid unusable in some applications; [Flexibility of the grid can have value in particular imaging applications, but is not currently available.] and (vii) not readily adaptable to different imaging conditions. In addition, grid use increases the required patient exposure, because of the needed compensation for absorption of primary radiation by the interspace material that forms the grid. A greatly enlarged cross sectional portion of a simple, conventional image contrast anti-scatter grid 24 is schematically shown in FIG. 2. In the grid, slats of x-ray opaque lead foil 28 alternate with filler strips of x-ray transmissive aluminum foil 26 or fiber. The height of the grid is h, and the interspace width is w. The ratio r=h/w is known as the grid ratio. In practice, for this ratio, h/w=16/1 may be considered a maximum. To achieve this ratio without reducing a transmission magnitude of the grid requires a large number of slats (i.e., a small value of w), since the available h is limited by the current use and design of x-ray equipment to values of about two millimeters. Slats, generally formed of lead, add significantly to grid weight. Another type of grid, shown in U.S. Pat. No. 2,605,427 issued Jul. 29, 1952, to Delhumeau is a two-dimensional focusing grid, so called because the slats are aligned with the rays coming from the x-ray source. Two-dimensional anti-scatter grids can be nearly twice as heavy as one-dimensional grids due to the additional amounts of x-ray absorbent material that are needed. U.S. Pat. No. 6,408,054 to Rahn et al. describes a micromachined contrast grid having numerous tiny holes formed by etching and photolithography, for example. The holes can be of selected depths and angles. The holes are then filled with small amounts of lead or other x-ray absorbing material to form a grid pattern. Various coating processes are described for filling the cylindrical holes formed in the grid substrate. Among solutions proposed for grid fabrication is forming multiple sheets and aligning the sheets to each other to define the path of incident radiation through the grid. U.S. Pat. No. 4,951,305 to Moore et al. describes one approach to this problem using aligned sheets. Accurate alignment of multiple sheets to each other, however, proves to be difficult, even where extremely tight manufacturing tolerances are maintained. Moreover, subsequent handling of the grid can easily cause inadvertent misalignment of the successive sheets used in such an arrangement. Rigid grids do not adapt to nonplanar detector surfaces and allow only a very small focus range. Because grids have traditionally been formed using lead strips separated by aluminum spacing material, grids have been formed as rigid, planar devices that do not flex. However, there may be some applications for which some amount of grid flexure is of value. Existing approaches used for grid fabrication do not allow flexibility of the grid. It can thus be appreciated that there are advantages to grid design that helps to remedy one or more of the identified factors that make conventional grids cumbersome or unsatisfactory for use. An object of the present disclosure is to advance the art of radiography and to address the need for fabrication of a flexible grid. Features of the present disclosure include improved capabilities for alignment of sheets in a multiple-sheet grid. These objects are given only by way of illustrative example, and such objects may be exemplary of one or more embodiments of the invention. Other desirable objectives and advantages inherently achieved by the disclosed invention may occur or become apparent to those skilled in the art. The invention is defined by the appended claims. According to one aspect of the disclosure, there is provided an antiscatter grid for radiological imaging, the grid formed as a stack of two or more sheets of a flexible substrate, wherein each sheet has: (i) a plurality of spaced-apart opaque cavities, each opaque cavity containing a radio-opaque material and wherein the arrangement of opaque cavities within the stack of sheets defines a plurality of channels for ionizing radiation that extend through the sheets; and (ii) a plurality of magnets that are disposed along the sheet and that couple the sheet to one or more neighboring sheets within the stack. The following is a detailed description of the preferred embodiments, reference being made to the drawings in which the same reference numerals identify the same elements of structure in each of the several figures. Where they are used herein, the terms “first”, “second”, and so on, do not necessarily denote any ordinal, sequential, or priority relation, but are simply used to more clearly distinguish one element or set of elements from another, unless specified otherwise. In the context of the present disclosure, the terms “viewer”, “operator”, “viewing practitioner”, “observer”, and “user” are considered to be equivalent and refer to the viewing practitioner or other person who views and manipulates an x-ray image on a display monitor or other viewing apparatus. As used herein, the term “energizable” relates to a device or set of components that perform an indicated function upon receiving power and, optionally, upon receiving an enabling signal. The term “actuable” has its conventional meaning, relating to a device or component that is capable of effecting an action in response to a stimulus, such as in response to an electrical signal, for example. The term “subject” refers to the patient who is being imaged and, in optical terms, can be considered equivalent to the “object” of the corresponding imaging system. The term “set”, as used herein, refers to a non-empty set, as the concept of a collection of elements or members of a set is widely understood in elementary mathematics. The term “subset”, unless otherwise explicitly stated, is used herein to refer to a non-empty proper subset, that is, to a subset of the larger set, having one or more members. For a set S, a subset may comprise the complete set S. A “proper subset” of set S, however, is strictly contained in set S and excludes at least one member of set S. In the context of the present disclosure, the term “oblique” means at an angle that is not an integer multiple of 90 degrees. Two lines, linear structures, or planes, for example, are considered to be oblique with respect to each other if they diverge from or converge toward each other at an angle that is at least about 10 degrees or more away from parallel, or at least about 10 degrees or more away from orthogonal. In the context of the present disclosure, the term “coupled” is intended to indicate a mechanical association, connection, relation, or linking, between two or more components, such that the disposition of one component affects the spatial disposition of a component to which it is coupled. For mechanical coupling, two components need not be in direct contact, but can be linked through one or more intermediary components. Radio-opaque materials are those that absorb and thus attenuate the x-ray beam significantly enough for detection and radiographic imaging and are considered to be non-transparent to x-rays at given energy levels. A well known radio-opaque material typically used for grids is lead. Radio-lucent or transmissive materials do not significantly absorb or attenuate the x-ray radiation. Non-magnetic materials are materials that are negligibly affected by magnetic fields and that exhibit no perceptible magnetic attraction and are thus not perceptibly pulled toward a magnet. In general, non-magnetic materials have a low relative magnetic permeability, typically not exceeding 1.0 at room temperature. Some exemplary non-magnetic materials include copper, aluminum, standard stainless steel, and most metals and alloys; sapphire; various ceramics; wood and paper composite materials; glass; water; plastics and other polymers; fiberglass; and various composite materials such as phenolic materials. Magnetic materials have higher relative permeability and are considered to be “magnetically responsive”, exhibiting magnetic attraction that can be readily perceived without requiring instrumentation; this includes ferromagnetic materials and various compounds of rare earth materials, for example. There are two general classes of ferromagnetic materials. Permanent magnets are made from “hard” ferromagnetic materials such as alnico and ferrite that are subjected to special processing in a powerful magnetic field during manufacture, to align their internal microcrystalline structure to exhibit a magnetic flux field. Magnetically “soft” materials like annealed iron, on the other hand, can be magnetized for a period of time, but do not tend to stay magnetized. To demagnetize a saturated magnet, a magnetic field of a given threshold must be applied, and this threshold depends on coercivity of the respective material. “Hard” materials that behave as permanent magnets have high coercivity, whereas “soft” materials have low coercivity. By way of example, electrical steel, used as a flux carrier in many electrical devices, exhibits coercivity values in the range of about 0.5 oersteds; samarium cobalt, used for rare earth permanent magnets, has coercivity in the range of about 40,000 oersteds. In the context of the present disclosure, a flexible sheet is a sheet that can be bent from a substantially flat planar form over a bend radius of 100 millimeters or less and can be restored to substantially flat planar form without damage. Embodiments of the present invention address a number of problems that relate to grid fabrication and can be used to form an antiscatter grid that is flexible. The methods and apparatus of the present invention can be used to configure a wide range of grid patterns to allow for various radiographic imaging applications. The perspective view of FIG. 3 shows schematic representation of a grid 30 formed according to an embodiment of the present disclosure. Features are not shown to scale, but are exaggerated in size to represent their physical arrangement. Grid 30 is formed as a stack 48 of sheets 32 that are aligned in a suitable arrangement for a particular imaging application. There are nine sheets 32 according to an embodiment of the present disclosure shown in FIG. 3. The number of sheets 32 that are used can be varied, allowing grid 30 to be formed from a stack 48 of as few as two sheets 32, or more. Each sheet 32 is formed from a flexible material that has numerous radio-opaque cavities 34 that are filled with radio-opaque material 42. Successive sheets 32 in the stack 48 are registered to each other so that their respective radio-opaque cavities 34 align to provide a suitable pattern of clear channels bounded and defined by the radio-opaque cavities 34. Thus, each neighboring sheet 32 can be the same or can be separately configured. As shown in FIG. 4, alignment of radio-opaque cavities 34 defines how clear channels 46 are arranged in the stack 48. Groups of aligned radio-opaque cavities 34 in successive sheets 32 can each be formed along an axis, with the axes of the groups of radio-opaque cavities 34 converging toward a focal point that lies some distance outside the sheet 32. Sheets 32 stacked to form an antiscatter grid can have radio-opaque cavities 34 directed toward the focal point so that, when the full set of sheets 32 for an antiscatter grid are stacked, clear channels 46 are formed extending from the top of the stack 48 to the bottom, with the clear channels 46 inclined toward the focal point. Cavities 34 themselves can each be formed along an axis A that is normal (angle N) to the sheet surface, with focus provided by staggering respective axes A of each of the radio-opaque cavities 34 from one sheet 32 to the next, so that radio-opaque cavities 34 on each sheet 32 are offset slightly from those on the neighboring sheet 32, effectively forming a substantially clear channel 46 for radiation when the sheets 32 are stacked, as shown in FIG. 4. Cavities 34 can be tiny and may not be visible to the unaided eye, so that each sheet 32 can have as many as a few million radio-opaque cavities 34, each extending within sheet 32 in the depth direction (h in FIG. 2) from one major surface (upper or lower) to the other major surface (lower or upper). The exploded side view of FIG. 4 shows a portion of one possible stack 48 arrangement, in which radio-opaque cavities 34 are formed along axes A normal to a surface of each sheet 32, which sheets 32 are successively staggered in the horizontal direction with respect to each neighbor sheet 32, for the orientation shown. This type of arrangement may provide oblique clear channels 46 formed by the stack 48 for focusing of incident radiation to form a focused grid 30, for example. Other arrangements can be envisioned for special purpose imaging, taking into account factors such as source-to-image distance (SID), grid resolution, amount of scatter compensation that is needed, and the like. Sheet Substrate In order to provide a flexible grid 30, sheet 32 may be formed from a material that is substantially transparent to radiation and that is also flexible. Substrates 40 that can be used for forming sheet 32, such as shown in FIG. 5, include, but would not be limited to, various polymer materials including polyester and Polyethylene Terephthalate (PET), for example. The substrate sheet material can be dimensionally stable in the plane of the sheet 32 but flexible out of the plane. Forming Sheet Cavities Cavities 34 can be formed in any of a number of ways, depending on size and spacing requirements and properties of the substrate that is used. According to an embodiment of the present disclosure, radio-opaque cavities 34 and other features (e.g., holes) are formed by sheet treatment with a laser, such as a CO2 laser, to form holes or cavities through a sheet 32 initially. Lasers from Universal Laser Systems, Trotec Laser Systems, or other manufacturers can be used. FIG. 5 shows a cross section of finished sheet 32 formed with angled radio-opaque cavities 34. The side view of FIG. 6 shows one method for forming holes 33 through one or more sheets 32. Using this technique, one or more sheets 32 are bent about a radius having a preselected magnitude; then, a laser is used to form holes 33 through the one or more sheets 32. When a sheet 32 that has been treated in this way is straightened, the resulting holes 33 may be disposed at variably controlled angles. Holes 33 or radio-opaque cavities 34 in a plurality of stacked sheets 32 may also be drilled or otherwise formed simultaneously. Where this is done, the holes 33 may extend fully through several sheets 32. Other methods can be used for forming radio-opaque cavities 34, including molding, perforation, drilling, and etching, for example. Each radio-opaque cavity 34 or hole 33 may have a diameter of about 2/1000 of an inch or less. Cavities 34 can be uniformly sized and spaced. Alternately, radio-opaque cavities 34 may have variable size and/or spacing along the sheet 32. Filling Sheet Cavities The fabrication sequence shown in FIGS. 7A through 7D shows how sheet 32 can be formed, according to an embodiment of the present disclosure. A flexible substrate 40 is shown in FIG. 7A. A cavity-forming operation is then used to form an array of tiny cavities 34, as shown in FIG. 7B. As noted previously, each cavity 34 can have a central axis A that may be normal to the flexible substrate 40 surface or may be oblique with respect to the flexible substrate 40 surface. Then, as shown in FIG. 7C, a filling operation is performed, filling each cavity 34 with a radio-opaque material 42. FIG. 7D shows addition of an x-ray transparent sealant layer 44, which may be a thin sheet of the same substrate material, for example, or some other suitable sealing material to form a finished sheet 32. The additional x-ray, transparent sealant layer 44 can also be applied to the opposite side of flexible substrate 40. Where holes are drilled completely through the starting substrate sheet, sealing material or treatment is applied to both surfaces of the sheet, thereby transforming the filled hole into a filled cavity. The radio-opaque material 42 that is used for filling holes or cavities 34 may be any of a number of materials and can be provided in nanoparticulate or powdered form, or provided in solution, such as an emulsion or colloidal solution, for example. Finely ground metals such as tungsten (W) or lead (Pb) can be used. Tungsten particulate is an advantageous alternative to lead. One commercially available tungsten powder (Buffalo Tungsten, type SR) has apparent powder density 8.5-10 g/cm3. By way of reference, solid lead has a density of 11.3 g/cm3. According to an alternate embodiment of the present disclosure, as shown in the partial fabrication sequence of FIGS. 8A and 8B, a squeegee 62 is used to press a tungsten-based paste 64 as radio-opaque material 42 into holes 33. The high density emulsion can then be at least partially dried. Excess tungsten-based paste 64 is then scraped from the top and bottom of the filled substrate and a protective film 66 applied to both top and bottom surfaces to form the finished sheet. Alternately, particulate or nanoparticulate material could be forced into the holes 33 as radio-opaque material 42. According to an alternate embodiment of the present invention, radio-opaque material 42 is a particulate and is combined with a curing material that binds the particulate to the side walls of holes 33 (or cavities 34) so that no sealant layer 44 (FIG. 7D) or protective film 66 is needed. Once holes 33 have been filled and sealed, assembly of the sheet 32 with other sheets 32 to form the grid can be performed. Sealing can also be performed using heat to encapsulate material within the cavity 34 or hole 33, or by using an adhesive, a hardener, or other material or process to encapsulate the radio-opaque material 42 or to harden the material into place. Sheet Coupling and Alignment Grid 30 can be formed from a number of suitably prepared sheets 32. To provide close registration between neighboring sheets 32 and all of the sheets 32 for a particular grid 30, holes 33 for all of the sheets 32 used for a grid 30 can be cut through at the same time. Angled cavities 34 may be created in a sheet 32 by forming a flexible substrate 40 in radius, then cutting holes 33 through the flexible substrate 40, then re-flattening the flexible substrate 40. Specialized, focused grid patterns may be formed, such as patterns that are suitably randomized to reduce or eliminate fringe patterns in the image that is obtained using the grid 30. The area of the pattern of cavities 34 on the sheet 32 can be a standard grid size, such as 435 mm×435 mm [17.1″×17.1″], for example. The stack 48 of sheets 32 forming the grid 30 can be about 42 cm×34 cm or about 42 cm×42 cm. Sheet alignment is a particular challenge with cavities 34 of the number and size needed. Sheets 32 can be coarsely aligned by permanently fastening together, either at one end or in the middle of the sheet 32 or at the perimeter. According to an embodiment of the present disclosure, sheets 32 are coarsely aligned in some way, then precision aligned and held together or coupled using magnetic attraction. Referring to the partial cross-section view of FIG. 9A, a pattern of cavities 34 are filled with a magnetized material 38 that give the corresponding magnetic cavities, now numbered cavities 36 in FIG. 9A and following, a particular magnetic polarity arrangement. Mutual attraction between magnetic cavities 36 that have magnetized material 38 causes these magnetic cavities 36 to align with each other and couple the sheets 32 together as shown in FIG. 9B. According to an embodiment of the present disclosure, magnetized material 38 is selectively deposited into a patterned distribution of magnetic cavities 36, so that the pattern of magnetic cavities 36 is the same for adjacent sheets 32. The pattern can be a matrix of magnetic cavities 36, such as having a magnetic cavity 36 at every increment of a given number of millimeters from a given origin, for example. The magnetized material 38 is then magnetized (typically, permanently magnetized) so that magnetic poles north (N) and south (S) are oriented to correctly attract each sheet 32 to its adjacent neighboring sheet 32 for a magnetic coupling. The magnetization step can be performed using a fixture that is designed with magnetizing elements in accordance with the given pattern. This magnetization processing can be performed prior to filling cavities 34 or holes 33 with high density materials to form clear channels 46 (FIG. 4). According to an alternate embodiment of the present invention, the patterned distribution of magnetic cavities 36 includes a matrix having both magnets and magnetic materials 38, such as ferrous materials that are magnetically responsive but are not themselves magnetized. The exploded view of FIG. 10A shows an alternate arrangement in which magnets 50a, 50b, 50c that are embedded within or coupled to the sheet can be used to magnetically couple and align neighboring sheets that may be placed in order, such as to provide radio-lucent channels 46 that are orthogonal to the sheet 32 surface or to provide oblique radio-lucent channels 46 using staggered-pattern sheets that allow the channeling of radiation in a focused manner. The same principles described with reference to magnets 50a, 50b, and 50c in FIG. 10A apply when using magnetized material 38 that is embedded within cavities 36, as described previously. In the example of FIG. 10A, three sheets are shown, labeled as sheets 32a, 32b, and 32c. Neighboring sheets 32a and 32b have corresponding magnets 50a and 50b, respectively, that are attracted to each other for coupling and alignment. Similarly, adjacent or neighboring sheets 32b and 32c have corresponding magnets 50b and 50c that have magnetic attraction to support alignment. Where each sheet 32 in the stack 48 that forms grid 30 is similarly constructed and the order of sheets 32 does not matter, the same pattern of magnet placement can be used on each of the neighboring sheets 30. In fabricating grid 30, the individual sheets 32 that form the grid 30 can be sequentially fitted onto the stack 48 in a sequence, such as by first registering only the edge of each subsequent sheet 32 to the existing stack 48 of sheets 32 and then unrolling the surface of the new sheet 32 onto the existing stack 48 to increasingly engage magnets in succession on each added sheet 32 and to reinforce magnetic attraction along the sheet 32 as it is rolled into position. FIG. 10B shows an arrangement of grid 30 formed from a stacking of sheets 32, not drawn to scale. Magnetized materials 38 are provided in cavities 36 that align in columns 54. As is particularly shown in the enlarged view of FIG. 10C, columns 54 that provide magnetic coupling and alignment may be arranged along axes A1. Axes A1 may be at angles that are offset from the angle A2 of cavities 34 that collectively form channels 46. For example, angle A2 may be normal while angle A1 is oblique, or vice versa, for portions of a focused grid in which cavities 36 are angled at normal, or other than a normal, to the sheet 32 surface. Columns 54 are shown as cylindrical and are shown distributed at spaced intervals along the stacked sheets 32. FIG. 10B shows some details of grid 30 according to an embodiment of the present disclosure. In the exemplary embodiment shown, the section of grid 30 shown has approximate dimensions of about 10.0×5.0 mm There are eight layered sheets 32 shown, each sheet having approximately 0.20 mm thickness; there can be some incremental spacing between sheets due to cavity features, giving a total grid stack height of about 1.79 mm Protective film thickness of sealant layer 44 (FIG. 7D) is about 0.012 mm Cavity 34 diameter is approximately 0.05 mm Cavity 34 centerline is about 0.25 mm, nominal, so that 40 cavities/cm are provided in this example. Magnet column 54 diameter is approximately 0.10 mm. One or more electromagnets can alternately be used to supplement the positional registration and coupling provided by permanent magnet materials of magnets 50a, 50b, 50c (FIG. 10A) between two or more sheets 32. Magnet-assisted coupling and alignment between neighboring sheets 32 is advantageous because it allows some amount of flexure to the stack, so that rigidity is not a requirement for maintaining sheet 32 alignment. Instead, grid 30 can be flexibly bent or curved in order to adapt more closely to the path of incident radiation through a material object or a human or animal subject and to a receiver. As the stack 48 is bent, magnetic attraction still holds the sheets 32 together and supports restoring the grid 30 to a flat, planar state. The magnetic coupling between sheets 32 allows some measure of flexibility and movement between sheets, so that a roller of a given radius can be used to move the grid into and out from the image path. The flexible grid 30 can thus be transported about the radius of the roller, yet remain in the bucky 52, for example, as shown subsequently. Tensioned wires can be used to help separate sheets when bending or otherwise in flexure, such as about a roller. One exemplary application in which flexure can be particularly useful is for x-ray imaging of large pipes, such as utility or chemical processing piping. The capability to wrap around the curved surface of the pipe or other structure has a number of advantages for non-destructive testing, for example. Magnets can be formed by filling a subset of the cavities 36 with a liquid magnetizable material, then applying a blade or squeegee to the surface of the sheet, and then magnetizing the material after it sets. Sheets 32 can be polarized so that all cavities 36 are N-polarized on one side of the sheet and S-polarized on the opposite side. Alternately, magnetization itself can be patterned, so that, from the same side of the sheet 32, a portion of the filled cavities 36 are N-polarized and the remaining portion of the filled cavities 36 are S-polarized. Other methods can alternately be used for coupling sheets 32 to each other in the stack. These include use of adhesives, fasteners, frames and other holders, and other devices. Magnetic coupling using micro-magnet structures such as cavities 36 is particularly advantaged, due to its capability for effecting precision alignment between sheets, at multiple points along the sheet surface. The perspective view of FIG. 11 shows a portion of flexible grid 30 that is formed from individual sheets 32 according to an embodiment of the present disclosure. Numerous sheets 32 can be used, depending on the desired thickness and other characteristics. It can be appreciated that the use of individual sheets 32 allows grid 30 to have any of a number of desirable arrangements, for focused or unfocused grids. The schematic view of FIG. 12 shows a focused grid 30, wherein channels 46 that extend through the grid 30, from sheet to sheet as described previously, substantially converge toward a focal point, shown as x-ray source 60. The position of a patient 14 and an x-ray detector 56 are also shown in FIG. 12. The x-ray detector 56 that is used for obtaining an image can be a digital radiography (DR) receiver or can be a film or computed radiography (CR) receiver, for example. The schematic view of FIG. 13 shows an optional bucky 52 that has a transport apparatus 70 with rollers 68 or other mechanics that can be energized to move grid 30 from a first position in front of x-ray detector 56 to a second position, behind the x-ray detector 56. Because grid 30 can be formed from separable polymer sheets that can be magnetically coupled and aligned, grid 30 is flexible and can be transported around the radius of a roller or other curved surface. Grid channels 46 (FIG. 12) can be slightly out of alignment during transport, since registration of magnetic cavities 36 to each other can be restored automatically by the magnetic alignment feature. An actuator 72 rotates one or both rollers 68 in order to transport grid 30 between positions. Bucky 52 can also be energizable to use rotation in order to change grids. In such an embodiment, bucky 52 moves a second grid 30b (shown in dashed outline) having a different focal distance or arrangement of channels 46 in front of the x-ray detector 56 after the first grid 30 is moved out of position. Alternately, other types of transport apparatus can be used, providing U-shaped movement about a roller radius, for example. The top view schematic view of FIG. 14 shows a grid 30 in transport apparatus 70 wherein a separation mechanism 80 is used to temporarily separate sheets 32a, 32b, 32c, etc. in order to relieve grid 30 stiffness and allow grid 30 to be transported about a radius around an x-ray detector 56. According to the embodiment shown in FIG. 14, the separation mechanism 80 includes one or more combs 76 of wires 78 that extend through grid 30 orthogonal to the radius of curvature, shown as dots in the cross-section top view of FIG. 14. When grid 30 is moved past rollers 74 (to the right in the FIG. 14 view), combs 76 are manipulated by an optional actuator (not shown) or by spring tension that spreads the comb wires 78 outwards. This relaxes grid 30 thickness by causing sheets 32a, 32b, 32c, etc. to separate slightly from each other as grid 30 is driven around the turn radius. Wires 78 lie outside the corresponding image area of grid 30 and can be formed of stiff wire, such as piano wire or other stiff structure. According to an alternate embodiment of the present disclosure, combs 76 are held in place while grid 30 moves through the turn radius. Because the sheets 32a, 32b, 32c, etc. are magnetically attracted to each other, reversing the turn direction allows the grid 30 to reconstruct itself, with its cavities properly aligned for imaging when the grid 30 is transported back into imaging position. The pattern of radio-opaque cavities 34 can be adapted to suit particular imaging requirements, with variable cavity angle and width and depth dimensions. Cavity shape and size can be different within the same sheet or in adjacent sheets 32 in the stack that forms grid 30. The channels 46 that extend through the stacked sheets 32 can substantially converge toward a focal point, such as toward the location of the x-ray source, as shown in FIG. 12. Grids having different focal lengths can be formed by placing inter-sheet spacers of different thicknesses between each sheet 32 in a grid. An inter-sheet spacer can be an untreated sheet of the same substrate that is used for forming sheet 32, for example. The flexibility of grid 30 is a function of factors such as the substrate 40 that is selected for sheet 32 material, density and size of radio-opaque cavities 34, radio-opaque material 42 (FIG. 7C) used to fill radio-opaque cavities 34, sheet thickness, number of sheets 32 stacked together (FIG. 10B), and how the sheets 32 are coupled to each other. While the invention has been described with reference to exemplary embodiments for flexible grid fabrication and use, it will be understood by those skilled in the art that various changes may be made and equivalents may be substituted for elements thereof without departing from the scope of the invention. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the invention without departing from the essential scope thereof. Therefore, it is intended that the invention not be limited to the particular embodiment disclosed as the best mode contemplated for carrying out this invention, but that the invention will include all embodiments falling within the scope of the appended claims.
	abstract	The present invention relates to a spent fuel dry reprocessing method for directly obtaining a zirconium alloy nuclear fuel, comprising: determining components and a ratio of a molten salt composition used for melting a spent fuel according to a requirement of reactor design on a zirconium alloy fuel and contents of actinium series metals in the spent fuel; melting the spent fuel in the above molten salt composition; and selecting an electrode pair for electrodeposition so that zirconium in the molten salt composition and uranium ions in the spent fuel or uranium and other actinium series metal ions are subjected to co-deposition, thereby obtaining the zirconium alloy nuclear fuel meeting a design requirement. The spent fuel dry reprocessing method provided by the invention is suitable for oxide spent fuel and metal spent fuel, and is simple and controllable in process, low in energy consumption, low in cost and easy to industrialize.
051981823	description	DESCRIPTION OF EXAMPLES EMBODYING THE BEST MOST OF THE INVENTION A neutron-absorbing tube according to the invention is shown generally at 10 in FIG. 1. Preferably, as will be described in greater detail immediately below, the tube 10 is formed of a pair of thin, rigid neutron-absorbing sheets 12 and 14 which are identical to one another, one of which is shown in FIG. 3, after rolling, trim and prior to bending. Each of the sheets 12 and 14 is formed from a metal ingot 16, as shown in FIG. 2. The metal ingot 16 is formed in an elongate, generally rectangular fashion from aluminum, therefore having aluminum faces 18 and 20, and opposite aluminum side edges 22 and 24. The ingot can be formed in accordance with the disclosure of incorporated U.S. Pat. No. 4,027,377, or in accordance with any other process for formation of a similar ingot. Once the ingot has been formed, at least one elongate, metal divider 26 is installed in the interior of the ingot to form the ingot into at least two chambers 28 and 30. The divider 26 shown in FIG. 2 is actually formed of one or more metal elements, and preferably is made of the same material as the metal ingot 16. Alternatively, the divider 26, side edges 22 and 24, and aluminum faces 18 and 20 can be formed as a single unit. Also, each metal divider 26 may, alternatively, be formed of a single length of metal, and as explained above, more than one metal divider may be installed in the hollow interior of the ingot 16, in order to form additional chambers. For example, if two metal dividers are installed, the dividers, in combination with the metal ingot, would form three chambers, three dividers would form four chambers, and so forth. Once the ingot has been thus-formed, the chambers 28 and 30 are filled with a substantially uniformly dispersed mixture of a finely divided neutron-absorbing compound and a finely divided metal powder. The mixture preferably is in accordance with that disclosed in detail in incorporated U.S. Pat. No. 4,027,377. Other similar materials, however, may be formed as well so long as they exhibit the neutron-absorbing capacity after a sheet has been formed. Once the ingot has been formed and filled, it is soaked to an elevated temperature which is below the melting temperature of the metal powder. In accordance with the preferred form of the invention, the ingot 16 is soaked for a period of time sufficient to raise the temperature of the ingot to between 800.degree. and 850.degree. Fahrenheit throughout, for rolling. In the rolling operation, the ingot 16 is hot rolled to reduce its thickness. The thickness is preferably reduced in a series of passes through a rolling mill. During that process, the mixture of boron compound and metal powder become metallurgically bonded together or sintered to form the sheets 12 and 14. The resulting sheet has thin, rigid neutron-absorbing areas 32 and 34, opposite metal edge portions 36 and 38, and an elongated metal spacer portion 40 at the location of each metal divider 26 in the ingot 16. The edge portions of the sheet 12 are trimmed as necessary so that preferably the combined widths of the edge portions 36 and 38 approximates the width of each spacer portion 40, although the combined widths can be wider or narrower, as well. After the sheet 12 has been formed, it is longitudinally bent at each spacer portion 40, preferably to an L-shaped cross section, as illustrated for each of the sheets 12 and 14 forming the tube 10 of FIG. 1. At the same time, the wider edge portion 36 is also bent to an L-shaped cross section. The two sheets 12 and 14 are then longitudinally joined at their edge portions 36 and 38 by welding, as shown at welds 42 in FIG. 1. It is preferred that the welds 42 not be located at the bends of the edge portions 36, but rather be formed on an unbent portion of the edge portion 36 to assure the greatest possible strength for the formed tube 10. Alternatively, if the edge portion 36 is not bent, the edge portions 36 and 38 can be abutted and welded where they abut. The tube 10 is therefore formed with neutron-absorbing areas 32 and 34 of the sheets 12 and 14 being unbent, but with bends in the spacer portions 40 and edge portions 36 being wholly within solid metal areas. The bends therefore do not affect the integrity of the neutron-absorbing core portions or areas of the sheet. Although the tube 10 is preferably formed of two sheets 12 and 14 formed and welded in accordance with the process set forth above, the tube 10 can also be formed from a single sheet having three metal dividers 26 installed in a metal ingot 16 forming four chambers in the ingot. After the ingot is filled and soaked, it is hot rolled in the same manner as described above, resulting in a sheet having opposite edge portions 36 and 38 similar to the sheet 12, but having three of the metal spacer portions 40 dividing adjacent neutron-absorbing areas. The thus-formed sheet is bent at each of the spacer portions to an L-shaped configuration, and also the edge portion 36 is bent to an L-shaped configuration, forming a tube appearing the same in cross section as the tube illustrated in FIG. 1. Since the tube is formed from a single sheet, only one longitudinal weld 42 is required to weld the edge portions 36 and 38 together to complete the tube. Although the tube 10 is shown in its preferred form as being generally square in cross section, the tube can be formed with a rectangular cross section. In addition, the tube can be formed with more than four sides, in a pentagonal, hexagonal or other configuration. However the tube is formed, it is important that the tube be formed with longitudinal bends being located wholly within a metal spacer portion dividing adjacent neutron-absorbing areas of the sheet or sheets forming the tube. Each of the metal dividers 26 may be formed of a single length of metal, or may be formed of one or more lengths of metal. However the dividers are configured, the ultimately-formed sheet is as illustrated in FIG. 3, with one or more spacer portions 40 being formed of metal, so that the sheet 12 may be bent at the spacer portions. It is also preferred that the sum of the widths of the edge portions 36 and 38 equal that of the spacer portion 40, so that when each of the sheets 12 and 14 is bent, the resulting formed tube 10 is essentially symmetrical. Again, however, that relationship is not critical, and the sum of the widths can be greater or less than the width of the spacer portion 40. A square tube 10 is illustrated in the drawing figures, manufactured from two of the sheets 12 having a central spacer portion 40. If a rectangular tube is desired, it will be evident that the metal divider 26 can be offset in the ingot 16 toward one or the other of the side edges 22 or 24 so that one of the chambers 28 or 30 is larger in the other, resulting in a sheet 12 with the spacer portion 40 offset toward one or the other of the edge portions 36 or 38. It will be evident to one skilled in the art that other shapes can be effected as well by simple location of the ultimate spacer portion or portions 40 and bending of the sheet or sheets 12 thereat. The width of the spacer portion 40 is formed to the minimum extent necessary so that the sheets 12 and 14 can be bent without bending or affecting the integrity of the neutron-absorbing areas 32 and 34. The width of the various spacer portions 40 will be evident to one skilled in the art, given the disclosure of the present application. Various changes can be made to the invention without departing from the spirit thereof or scope of the following claims.
	abstract	An extreme ultraviolet light source apparatus for supplying extreme ultraviolet light to a processing unit for performing processing by using the extreme ultraviolet light. The extreme ultraviolet light source apparatus includes: a chamber in which the extreme ultraviolet light to be supplied to the processing unit is generated; a collector mirror for collecting the extreme ultraviolet light generated in the chamber to output the extreme ultraviolet light to the processing unit; and an optical path connection module for defining a route of the extreme ultraviolet light between the chamber and the processing unit and isolating the route of the extreme ultraviolet light from outside.
051026139	abstract	A radial brake assembly is disclosed for selectively preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The brake assembly includes a stationary housing, a rotor disc fixedly connected to the shaft for rotation therewith, and a brake member disposed adjacent to the perimeter of the rotor disc. The rotor disc includes at least one rotor tooth and the brake member includes at least one braking tooth. The brake member is selectively positioned in a deployed position for allowing the braking tooth to contact the rotor tooth for preventing rotation of the shaft in a first direction, and in a retracted position for allowing the rotor disc and shaft to rotate without restraint from the brake member.
	summary
	claims	1. An optical apparatus for guiding EUV radiation to a predetermined surface, comprising:a radiation source which supplies EUV radiation; an illumination optical system having a reflective integrator which forms a secondary radiation source having a predetermined shape based on the EUV radiation supplied from the radiation source; and a projection optical system which is arranged in an optical path between a reflective mask and the predetermined surface and which forms an image of the reflective mask onto the predetermined surface based on the EUV radiation from the reflective mask, wherein the secondary radiation source having the predetermined shape has a shape which is adjustable. 2. The optical apparatus of claim 1, wherein a numerical aperture of the projection optical system is changeable. 3. The optical apparatus of claim 2, wherein a numerical aperture of the illumination optical system is changeable. 4. The optical apparatus of claim 3, further comprising an annular radiation beam converting unit which is arranged in an optical path between the radiation source and the reflective integrator. 5. The optical apparatus of claim 3, further comprising a multipolar radiation beam converting unit which is arranged in an optical path between the radiation source and the reflective integrator. 6. The optical apparatus of claim 5, wherein the multipolar shape comprises a quadrupolar shape. 7. The optical apparatus of claim 3, further comprising a unit that changes a radiation beam which is incident on the reflective integrator. 8. The optical apparatus of claim 3, wherein the projection optical system comprises six mirrors. 9. The optical apparatus of claim 1, wherein the secondary radiation source having the predetermined shape is changed based on information about the reflective mask. 10. The optical apparatus of claim 1, wherein the illumination optical system and the projection optical system are mask side non-telecentric. 11. A method of guiding EUV radiation to a predetermined surface, comprising the steps of:supplying EUV radiation; forming a secondary radiation source having a predetermined shape using a reflective integrator, based on the supplied EUV radiation; guiding the EUV radiation from the reflective integrator to a reflective mask; forming an image of the reflective mask using a projection optical system based on the EUV radiation from the reflective mask; and adjusting a shape of the secondary radiation source. 12. The method of claim 11, further comprising the step of changing a coherence factor. 13. The method of claim 11, further comprising the step of converting the supplied EUV radiation to an annular beam. 14. The method of claim 13, wherein the converting step is performed before guiding the supplied EUV radiation to the reflective integrator. 15. The method of claim 11, further comprising the step of converting the supplied EUV radiation to a multipolar beam. 16. The method of claim 15, wherein the converting step is performed before guiding the supplied EUV radiation to the reflective integrator. 17. The method of claim 16, wherein the multipolar shape comprises a quadrupolar shape. 18. The method of claim 11, wherein the adjusting step changes the shape of the secondary radiation source based on information about the reflective mask. 19. The optical apparatus of claim 3, wherein the predetermined shape of the secondary radiation source has a shape which is selected from the group consisting of a substantially circular shape, an annular shape, and a multipolar shape. 20. The method of claim 12, wherein the adjusting step selects a shape of the secondary radiation source from the group consisting of a substantially round shape, an annular shape, and a multipolar shape.
039768341	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to FIG. 1, there is shown a nuclear reactor 10 including a reactor vessel 12 having an active core or fuel region 14 therein. The core 14 comprises a plurality of fuel assemblies 16 that are supported in position by a lower support plate 18 having apertures (not shown) therein for admitting fluid coolant to the core 14. A core support shroud 20 surrounds the core 14 and is secured to the lower support plate 18. The lower support plate 18 and the entire core assembly 14 are vertically supported by a lower support assembly 22 which is, in turn, supported by the core support barrel 24, the latter being suspended from the vessel flange 26 by an out-turned lip 28 that engages the flange. In general and as shown in FIGS. 1, 3 and 4, the fuel assemblies 16 each include a plurality of longitudinally extending fuel elements 30 and a plurality of hollow guide tubes 32, 34 interspersed between and within the array of fuel elements 30. In the arrangements shown, the fuel assemblies 16 each contain four vertically extending outer guide tubes 32 which are arranged adjacent the four corners of a fuel assembly 16 and which are attached to upper and lower end fittings 36, 38. A fifth guide tube 34 is centrally located within each of the fuel assemblies 16 and is also attached to the upper and lower end fittings 36, 38. The guide tubes 32, 34 all extend above the upper end fitting 36 and may also extend below the lower end fitting 38, or other means may be provided, to engage the lower support plate 18 to support the fuel assemblies 16 (see FIG. 1). The guide tubes 32, 34 and the upper and lower end fittings 36, 38 provide the structural framework for the fuel assemblies 16. A plurality of generally rectangularly arranged spacer grids 40 (see FIG. 3) are suitably secured to the guide tubes 32, 34 at spaced elevations therealong. The spacer grids 40 serve to support the plurality of parallel fuel elements 30 in vertical disposition. The grids 40 are of conventional design and include, as is well known, an array of mutually aligned rectangularly arranged openings through which the fuel elements 30 extend. Located above the core region 14 is a guide structure assembly 42 that serves to align the top end of the fuel assemblies 16 and also to guide control elements 44 into and out of the core region 14. The guide structure assembly 42 shown in the figures is generally of the type described in copending application Ser. No. 266,858, now U.S. Pat. No. 3,849,257 entitled "Guide Structure for Control Elements" and filed June 28, 1972 by Frank Bevilacqua, although other types of guide structures are equally applicable for use with the present invention. The guide structure assembly 42 comprises a pair of vertically spaced tube sheets or plates 46, 48, each of which substantially overlies the entire core 14 and which are rigidly interconnected by a plurality of longitudinally extending hollow tubes 50. The guide structure assembly 42 of the present invention differs from that of the above-noted copending application in that it includes an integral emergency core cooling injection manifold 52 to be described in more detail hereinbelow. The tubes 50 are disposed, as shown in FIG. 3, with their opposite ends located slightly above the upper tube sheet 46 and slightly below the lower tube sheet 48, respectively. The preferred means of securing the tubes 50 to the tube sheets 46, 48 is somewhat different from that disclosed in the copending application to facilitate attachment of the injection manifold 52 and will also be described hereinbelow. A support barrel 54 (see FIG. 1) is secured to the upper tube sheet 46 and suspends the guide structure assembly 42 within the core support barrel 24. Alignment of the respective fuel assemblies 16 is effected by the telescopic reception of the upper ends of the four outer guide tubes 32 within the lower ends of the tubes 50. Control elements 44 are provided for vertical reciprocal movement into and out of the core 14 for controlling the neutron flux therein. The hollow tubes 50 of the guide structure assembly 42 provide guidance for these control elements 44 and also provide protection from the coolant cross flow in the outlet region 56 between the tube sheets 46, 48. As more fully described in the copending application, the control elements 44 comprise rods containing neutron poison which are situated for telescoping movement within some of the hollow tubes 50 and the fuel assembly guide tubes 32 in alignment therewith. Above the upper tube sheet 46 the control elements 44 may be ganged together and connected to a first type of drive mechanism 58 mounted on the head 60 of the reactor vessel 12, or may be coupled individually to a second type of drive mechanism 62. Operation of drive mechanisms 58, 62 controls the vertical position of the control elements 44 relative to the core 14 which in turn controls the power level of the reactor 10. In the case of ganged control elements, shrouds 64 may be provided which surround the ganged control elements above the upper tube sheet 46. During normal operation of the reactor 10, liquid coolant, which is usually water, enters the reactor 10 through the inlet nozzles 66 (one of which is shown in FIG. 1) and flows downwardly around the outside of the core support barrel 24. The coolant then flows inwardly and up through openings (not shown) in the lower support assembly 22 and in the lower support plate 18. As the coolant flows upwardly through the reactor core 14, it extracts heat generated therein from the nuclear fission in the fuel assemblies 16. The heated coolant then flows up through openings 68 in the lower tube sheet 48 into the outlet region 56 located between the two tube sheets 46, 48. From the outlet region 56, this fluid is passed through the outlet nozzles 70 (one of which is shown in FIG. 1) and conducted to a vapor generator or the like (not shown), where it serves as the operating medium for heating vaporizable liquids supplied thereto. Also, during this operation, some of the coolant in the reactor 10 flows upwardly through the hollow tubes 50 of the guide structure assembly 42 into the region 57 above the upper tube sheet 46 and back down to the outlet plenum 56 through openings (not shown) in the upper tube sheet 46. This coolant path is necessary to maintain proper cooling of the control elements 44 and to provide proper mixing of the water in the plenum region 57 above the upper tube sheet 46. Essentially the coolant in this region 57 is stagnant relative to the coolant in the outlet region 56. This fluid path up through the hollow tubes 50 is not considered to be part of the main reactor coolant flow path as described hereinabove. Means are provided in the described reactor arrangement for supplying emergency core coolant to avoid the danger of overheating the component parts of the apparatus in the event of an occurrence of an emergency condition, such as for example, a loss of coolant accident wherein primary coolant is prevented from circulating through the reactor core 14. In particular this means comprises a source of emergency core coolant liquid stored externally of the reactor vessel, an injection manifold or plenum 52 integrally formed with the guide structure 42, and means for distributing the emergency core coolant liquid to the manifold 52 and then into the reactor core 14. In one embodiment of the present invention, as best seen in FIGS. 2, 3, 5 and 6, the injection manifold or plenum 52 is located between the lower tube sheet 48 of the guide structure assembly 42 and a plate 74 positioned in spaced relationship thereabove in the outlet region 56. As seen in FIG. 2, the plenum forming plate or third plate 74 of the guide structure assembly 42 has a first and second plurality of holes or openings 76, 78 therethrough to permit passage through the plate 74 of the hollow tubes 50 and main reactor coolant fluid, respectively. The holes 76, 78 of the first and second plurality in the third plate 74 are each aligned with similar holes or openings 80, 68 in the lower tube sheet 48. Annular spacers or sleeves 82, 84 for the aligned openings 76, 80 and 78, 68 are provided between the two plates 48, 74, the spacers 82, 84 each being co-axially aligned with one of the aligned openings 76, 80 or 78, 68. The annular spacers 82, 84 serve to maintain the axial spacing between the two plates 48, 74 defining injection manifold or plenum 52 and further to substantially seal the main reactor coolant fluid from the interior of the manifold 52. In the preferred embodiment, as shown in FIG. 5, the spacers 84 for the main coolant flow holes 78, 68 are maintained in co-axial alignment by having a reduced portion 86 which is received within the flow hole 78 in the third plate 74. Shoulders 88 formed in the outer surface of the spacers 84 act to support the plate 74 above the lower tube sheet 48. The lower ends of the spacers 84 engage the upper surface of a lower tube sheet 48 and are concentrically positioned over the flow hole 68 in the lower tube sheet 48. This offers little increase in fluid resistance to the coolant flowing therethrough during normal reactor operation. The spacers 82 for the hollow tube openings 76, 80 are maintained in co-axial alignment therewith by being concentrically positioned around the hollow tubes 50 between the two plates 48, 74. The hollow tubes 50 are secured to the upper and lower tube sheets 46, 48 and the third plate 74 secured to the guide structure assembly 42 by use of double concentric tubes similar to the method disclosed in the aforementioned copending patent application. The inner tubes 50 (i.e. the tubes which guide the control elements 44 and which receive the ends of the fuel assembly guide tubes 32) each extend through the two tube sheets 46, 48 with the opposite ends located slightly above the upper tube sheet 46 and below the lower tube sheet 48, respectively. Each of the inner tubes 50 has a flanged portion 92 at one end and a threaded portion 94 at the other end which is adapted to receive a nut 96. The second outer tubes 98 are concentrically disposed around each of the inner tubes 50 between the upper tube sheet 46 and the third plate 74. By tightening the nuts 96 on each of the inner hollow tubes 50, the concentric outer tubes 98 and the sleeves or spacers 82, 84 around the tubes 50 or in the flow holes 78 are placed in compression while the inner hollow tubes 50 are placed in tension. This arrangement provides a rigid construction for the guide structure assembly 42 while at the same time it rigidly secures and spaces the third plate 74 above the lower tube sheet 48. Furthermore, by placing the sleeves 82, 84 between the third plate 74 and the lower tube sheet 48 in compression, this provides a more efficient seal to prevent ingress of main coolant fluid into the injection manifold 52. It should be noted that it is not necessary to provide a completely effective seal between the manifold 52 and the main reactor coolant flow path. During normal reactor operation, leakage of main coolant fluid into the manifold plenum 52 will be minimal due to the higher flow resistance into the plenum 52 as compared to the flow resistance through the flow holes 68, 78 to the outlet plenum. This minimal leakage will have no adverse effect on reactor operation. During an accident condition, leakage will again be minimal and most emergency core coolant liquid will be dispersed through the distribution means as described hereinbelow. Any leakage of emergency core coolant liquid which does not occur out of the manifold 52 will merely aid in further cooling of the reactor 10. In the preferred embodiment emergency core coolant fluid introduced into the injecton manifold or plenum 52 is directed to the core 14 by means of the central guide tubes 34 of selected fuel assemblies 16. The lower tube sheet 48 of the guide structure assembly 42 is provied with openings 102 in the lower surface thereof which are located to receive the upper ends of the fuel assembly central guide tubes 34. Although it is contemplated that each of the openings 102 will communicate with the interior of the manifold or plenum 52 several of the openings could be plugged or otherwise not communicate completely through the tube sheet 48 with the plenum 52 if it is desired to achieve some other particular type of cooling pattern in the core 14. For those openings 102 which do communicate with the plenum 52, emergency cooling liquid will be directed therethrough and into and down the interior of hollow central guide tubes 34. Exit openings 104 in the sidewalls of the central guide tubes 34 are provided along the longitudinal length and around the circumference thereof to permit the coolant to exit in any desired region of the core 14. Although only two exit openings 104 ae shown for the central guide tubes 34 in FIGS. 3 and 4, additional openings could be provided at other vertical and/or circumferential positions to direct coolant to particular regions such as "hot spots" in the core 14. The particular path of emergency coolant distribution for any particular core, of course, depends on a number of factors including the locations and types of each of the various fuel assemblies 16 comprising the core. By spraying emergency coolant in the interior region of the core 14, the emergency coolant is located closer to each of the fuel elements 30 upon which the emergency coolant must operate. Consideration, of course, must given to insure that each of the fuel elements 30 will receive adequate cooling to prevent rupture thereof and release of contaminated fission products. Emergency core coolant fluid is introduced into the injection manifold or plenum 52 from a source of coolant stored externally of the reactor vessel 12. The coolant liquid is contained in a storage tank 106 which preferably includes a means (not shown) for maintaining the liquid under an elevated pressure within the tank 106. It is contemplated that a pressure between 300 and 1,200 psi will be suitable for this purpose; however, higher or lower pressures are not excluded. A flow line 108 connects the tank 106 in fluid communication with an emergency core coolant injection nozzle 118 located within the interior of the guide structure support barrel 54 through aligned openings 110, 112, 114 provided in the vessel wall 12, the core barrel 24, and the guide structure barrel 54, respectively. A valve 116 is provided in this line 108 and is activated in the event of an accident to supply emergency coolant to the injection nozzle 118. As best seen in FIG. 6, the injection nozzle 118 extends downwardly through the upper tube sheet 46 through the third plate 74 of the guide structure assembly 42 and is in communication at its lower end with the interior of the injection manifold or plenum 52. A spacer ring 120 is secured, as by welding, to the third plate 74 and the lower tube sheet 48 around the circumference of the tube sheet 48 to substantially seal the interior of the plenum 52 from the main reactor coolant fluid. It will be appreciated that the emergency coolant liquid containment systems as described herein, is for the purpose of illustration only, and that other and varied forms of cooling containment systems may be alternatively provided. For example, a plurality of tanks 106 and associated conductors 108 can be caused to communicate with a plurality of injection nozzles 118 at circumferentially spaced locations about the guide structure assembly 42. Preferably, to insure reliability and to provide redundancy and additional safety, it is contemplated that four emergency coolant liquid containment systems 106, 108 and injection nozzles 118 will be provided which would be spaced equally about the circumference of the reactor vessel 12. The location of one injection nozzle 118 for such an arrangement is shown in FIG. 2. Preferably the injection nozzle 118 is circumferentially spaced away from the entrance to the outlet nozzle 70 (which lies along the core centerline 122) so as not to be subjected to the high cross flow load thereat. For the remaining core quadrants, the other injection nozzles would be similarly located away from the outlet nozzles. FIG. 2 also shows a typical arrangement of the flow holes 68, 78, the hollow tubes 50 of the guide structure assembly 42, and the distribution holes 102 in the lower tube sheet 48 for one quadrant of the core 14, the fuel assemblies 16 therein being shown in dotted outline. The remaining three quadrants comprising the core 14 are symmetrical about the two core centerlines 122, 124. Four hollow tubes 50 in which the upper ends of the guide tubes 32 of the fuel assemblies 16 are received, are provided for each fuel assembly 16, some of the hollow tubes 50 being shown with control elements 44 situated therein. An emergency core coolant distribution hole 102 is provided in alignment with the central guide tube 34 of each fuel assembly 16 and into which the central guide tube 34 will be received. Also, axially aligned flow holes 68, 78 in plates 48 and 74 respectively are arranged to lie along the projected peripheries of the fuel assemblies 16. In the event of an emergency condition requiring the admission of supplementary coolant liquid to compensate for the loss of primary coolant and to remove the decay heat from the reactor core 14, the valve 116 in the conduit line 108 is actuated to an open position thereby placing the pressurized interior of the emergency coolant storage tank 106 in fluid communication with the emergency core cooling injection nozzle 118. Emergency coolant liquid under pressure is thereby caused to flow into the injection manifold or plenum 52 integrally formed with the guide structure assembly 42. In this region 52, the liquid passes down through the distribution holes 102 into the central guide tubes 34 whose upper ends are open and from there, downward toward the center of the core 14. From the central guide tubes 34, the liquid is directed through the radial openings 104 provided therein into the core 14 where it extracts the decay heat from the reactor fuel elements 30 to thereby maintain the temperatures of the apparatus within tolerable limits. One advantage obtained with use of the emergency core coolant system of the present invention is derived from the fact that the emergency coolant discharged into the core 14 is pressurized. Because the coolant is applied at an elevated pressure, the possibility of steam blockage within the upper region of the reactor core 14 is removed, thereby insuring that the coolant liquid will flow to all the affected regions of the core 14. With coolant fluid being transferred into the reactor vessel 12 at a pressure greater than that of the vapor generated therein in the event of an accident, the latter will simply be forced through the reactor or any openings in the vessel 12 such as the inlet or outlet nozzles 66, 70. Of course, as can be appreciated, the realization of this advantage will be dependent on the pressure maintained in the storage tank 106 and on the volume of the injection manifold 52 and distribution means as well as on other factors which are known to those skilled in the art. Another embodiment of the present invention is shown in FIG. 7. In this form of the present invention the plenum forming plate is indicated in the drawing as 74' and is located adjacent the upper tube sheet 46 of the guide structure assembly 42. The injection manifold, indicated as 52', is formed between the plate 74' and the upper tube sheet 46. This third plate 74' of the guide structure assembly 42 is provided with openings 130 therethrough for the hollow tubes 50 of the guide structure assembly 42 and is integrally attached to the guide structure assembly 42 in spaced relationship by means of spacer sleeves 132 and double concentric tubes 50, 134 in a manner similar to that shown for the embodiment in FIGS. 3 and 5. The guide structure inner hollow tubes 50 each have a flanged end portion 92 and a threaded end portion 94, and the spacer sleeves 132 and outer tubes 134 are concentrically positioned around the inner tubes 50 between the upper tube sheet 46 and the third plate 74', and the third plate 74' and the lower tube sheet 48, respectively. A nut 96 threaded onto the threaded end 94 of the inner tubes 50 places the outer tubes 134 and spacers 132 in compression and the inner tubes 50 in tension. Pressurized emergency coolant liquid is introduced into the injection plenum 52' in a similar manner to that described hereinabove (i.e. an emergency injection nozzle 118, in fluid communication with the storage tank 106, communicates at its lower end with the plenum 52' through the upper tube sheet 48). In this alternative embodiment the distribution means for discharging emergency core coolant fluid in the injection plenum 52' into the core 14 comprises a plurality of hollow distribution tubes 136 which extend through all three plates 46, 74', 48 and which have their opposite ends located slightly above the upper tube sheet 46 and slightly below the lower tube sheet 48, respectively. The distribution tubes 136, which each have a flanged end 138 and a threaded end 140, are attached to the guide structure assembly 42 by tightening nuts 142 onto the threaded end 140 to tension the tubes 136. The distribution tubes 136 are in alignment with and have their lower ends positioned so as to receive the upper ends of the central guide tubes 34 of selected fuel assemblies 16. The distribution tubes 136 each have radial openings 144 in the sidewall thereof at the elevation of the injection plenum 52' to provide fluid communication between the plenum 52' and the interior of these tubes 136. The upper ends of the distribution tubes 136 are each sealed by a cap 146 which is held in place by the nut 142 securing the tube 136 to the guide structure assembly 42. The central guide tubes 34 of the fuel assemblies 16 communicating with the plenum 52' are similar to those described hereinabove (see FIGS. 3 and 4) in that they have radial openings 104 in the sidewalls at desired vertical and circumferential positions along the length thereof to spray emergency coolant outward into the core 14 at desired locations and in desired directions. The distribution tubes 136, as shown, are of a smaller inner diameter than the inner diameter of the hollow tubes 50 of the guide structure assembly 42 to insure that the emergency core coolant liquid will remain pressurized to a sufficient level to be discharged into the core 14. Of course, this dimension may be varied and in fact, may have a larger inner diameter depending on a number of distribution tubes 136 and the pressure maintained in the externally located storage tank 106. It should also be noted that with the use of this alternative embodiment, the lower tube sheet 48 only is provided with flow holes 68 to permit circulation of the main reactor coolant fluid between the core 14 and the outlet region 56. While preferred embodiments of the invention have been shown and described, it will be understood that these are merely illustrative of the invention, rather than restrictive, and that changes may be made without departing from the scope of the invention.
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050826182	description	DESCRIPTION OF PREFERRED EMBODIMENT The plant shown in FIG. 1 is intended to modify the concentration of the soluble poison contained in the cooling fluid of the primary circuit 1 of a nuclear reactor, the cooling fluid consisting of boric acid diluted from 2000 to 100 ppm (parts per million) of boron. Generally, this plant comprises a circuit A for chemical and volumetric monitoring of the main primary circuit, a circuit B for modifying the boron concentration of the primary circuit, and a circuit C for processing the effluents and recycling the ammonia. The cooling fluid is removed from the primary circuit 1 by means of a duct 2 connecting a cold branch of a loop of the primary circuit 1 to the inlet on the hot side of a heat exchanger 3. The outlet of the heat exchanger 3 is connected, by means of pressure-reducing orifices 4, to a second heat exchanger 5 which further cools the cooling fluid to a temperature of between 30.degree. and 80.degree. C. The heat exchanger 3 and the pressure-reducing orifices 4 are located inside a sealed enclosure, the wall 6 of which is shown, while the remainder of the plant is located on the outside. A valve 7, if necessary, allows the circuit for modifying the concentration of the poison to be isolated from the primary enclosure. After passing into the exchanger 5, the cooling fluid is filtered at 8. The circuit A also comprises a dimineralizer 12 of the mixed bed type, the anionic resins of which are saturated with boron, and a volumetric monitoring tank 13. By means of a feed pump 14, the tank 13 is connected to the cold inlet of the exchanger 3 by a line 15 provided with a regulation valve 16. The tank 13 is also connected to the controlled leakage seals of the primary pumps by a line 17. The circuit B for modifying the boron concentration of the primary circuit comprises two cationic resin beds 19 and 19', a heat exchanger 23, two electrodialysis modules 20 and 21 and a reverse osmosis apparatus 27. The effluent processing and ammonia recycling circuit C comprises first of all a tank 24 which stores the primary liquid effluents such as those discharged from bleeders and vents, valve outlets, etc. of the primary circuit 1 (indicated in the drawings by the reference 51 and pipes to the tank 24 by the line 52). The tank 24 is connected by a line 23 to the inlet of each electrodialysis module 20 and 21. Circulation of the effluents in the line 23 is ensured by a pump 25. Furthermore, the circuit C has a line 34 leaving the tank 24 from which the effluents are removed by means of a pump 35. The line 34 conveys the effluents first of all into a mixed-bed demineralizer 36 in which the anionic resins are saturated with boron and the cationic resins are saturated with ammonium (NH.sub.4), and then into a module 37 for thermal dissociation of the ammonium borates, and a tank 38. This circuit C also comprises two reverse osmosis apparatuses 39 and 40 mounted in series, two pumps 41 and 42 and two tanks 43 and 44. The module 37 has leaving it a line 45 allowing readjustment of the ammonia solution content of the electrodialysis modules 20 and 21 and of the reverse osmosis apparatus 27. For this purpose, the line 45 comprises a compressor 46, a tank 47, a pressure-reducer 48, a tank 49 provided with a water inlet 53 and a pump 50. With reference now to FIGS. 2 to 8, the mode of operation of the plant will be described. If the boron concentration, which has been measured beforehand in respect of the primary circuit, is correct, the valve 10 remains closed whereas the valve 10' opens in order to allow the cooling fluid to enter into the demineralizer 12 and the volumetric monitoring tank 13. The cooling fluid therefore flows normally inside the circuit A as shown in FIG. 2. If the soluble-poison concentration of the cooling fluid is not correct, the valve 10' remains closed, whereas the valve 10 opens in order to allow the cooling fluid to enter into the circuit B for modifying the boron concentration of the primary circuit as shown in FIG. 3. Opening of the valve 18 allows the cooling fluid to pass over the bed of cationic resins 19 initially saturated with cations (NH.sub.4 + for example) and to extract from the latter ions which increase its base content. This promotes dissociation of the boric acid contained in the fluid before the latter enters into the electrodialysis modules 20 and 21. The efficiency of these modules is thus increased. Before entering into the electrodialysis modules 20 and 21, the temperature of the fluid is further lowered to a temperature of between 30.degree. and 50.degree. C. by the heat exchanger 22. The electrodialysis modules 20 and 21 are identical and each consist of compartments separated by membranes. Preferably, these membranes are of limited thickness and have a good mechanical strength, it being possible to use, for example, membranes which have a thickness of less than 1 mm and the surface area of which amounts to 1 m.sup.2. These membranes are alternately cationic and anionic. The cationic membranes allow only the cations to pass through, while the anionic membranes allow only the anions to pass through. The end compartments of the electrodialysis modules are provided with electrode plates 20a and 20b supplied by a direct current, and the polarity of these electrodes may, moreover, be reversed, as will be seen further on during the operational description of these apparatuses. The arrangement of one of these electrodialysis modules is shown schematically in FIGS. 4 and 5 where, for reasons of clarity, a small number of compartments has been shown. The primary cooling fluid circuit, after passing into the heat exchanger 22, is divided so that the latter circulates inside one compartment of two compartments (a). In the other compartments, (b), the fluid from the circuit for modifying the soluble poison concentration circulates, via the pipe 23. When it is necessary to increase the soluble-poison concentration in the primary cooling fluid, in this case the boron concentration in the example chosen, the electrodialysis modules 20 and 21 are supplied with direct current, as indicated in FIG. 4. Under the effect of this electric current, circulating between the anode and the cathode, a transfer of anions through the anion exchange membranes (AEM) and of cations through the cation exchange membranes (CEM) takes place. Owing to the organization of the fluid circuits inside the various compartments of the module, there results an increase in the ion, in particular BO.sub.2 anion and NH.sub.4 cation, concentration of the cooling fluid, and a corresponding reduction in the concentration of the fluid of the circuit for modifying the boron concentration. On the other hand, when the boron concentration in the cooling fluid is too high and must be reduced, the electrodialysis modules 20 and 21 are supplied with direct current, as indicated in FIG. 5, i.e., with the polarity of the electrode plates reversed compared to the mode of operation described above. In this case, there is a transfer of ions from the compartments (a) to the compartments (b), and in particular of BO.sub.2 anion through the anion exchange membranes (AEM) and of NH.sub.4 cations through the cation exchange membranes (CEM). This thus results in a reduction in the boron concentration of the primary cooling fluid and a corresponding increase in the concentration of the fluid of the circuit for modifying the boron concentration. After passing through the electrodialysis modules 20 and 21, the cooling fluid is directed towards the second bed of cationic resins 19', which fixes the ions which had been extracted from the bed 19. The cooling fluid is then ready to follow the return circuit towards the primary circuit 1. It is also possible to reverse the direction of circulation of the fluid in the ion exchange beds 19 and 19'. In fact, in the preceding description, it has been assumed that the cooling fluid passes through the bed 19 upstream of the electrodialysis modules 20 and 21, and the bed 19' downstream of the latter; at the end of a certain period of operation, the "upstream" bed from which the ions are extracted no longer contains an adequate amount thereof, while the "downstream" bed which fixes the ions becomes saturated. In order to determine the moment when this reversal must be performed, a pulse counter 26 totals the throughput of cooling fluid supplied to the electrodialysis module 20 and 21. When the counter 26 totals a certain throughput, reversal is performed automatically. When the boron concentration of the cooling fluid is sufficiently high, above 500 ppm, and when the variation in this concentration is not very high, less than 50 ppm per hour, the electrodialysis modules 20 and 21 are used only as shown in FIG. 3. However, when the boron concentration is less than 500 ppm or when the variation in concentration must be greater than 50 ppm per hour, the electrodialysis modules 20 and 21 may prove to be insufficient. In fact, under these conditions, modification of the boron concentration is possible, but lengthy. These conditions exist at the end of the life cycle of the fuel, when the quantity of boron reaches a low level or during operational transitions (restarting after stopping when hot, and maximum poisoning of the core by the xenon, etc.), or when the variation of the concentration is large. Under these conditions, the reverse osmosis apparatus 27 is used either alone or in combination with the electrodialysis modules 20 and 21. Start-up of the apparatus 27 is effected by a valve 28. Opening and closing of this valve may be performed automatically as a function of the boron concentration. The reverse osmosis apparatus 27 is a conventional apparatus comprising membranes which enable the cooling fluid to be separated into two solutions, one solution with a very low boron concentration which is removed from the reverse osmosis apparatus 27 via the line 29, and a second solution with a very high boron concentration which is removed via the line 30. If it is required to reduce the boron concentration of the cooling fluid (FIG. 6), the three-way cock 31 is opened so that the fluid removed via the line 29 is directed towards the bed 19' (or 19) so as to rejoin the primary circuit 1. In this case, the three-way cock 32 is open so that the fluid removed via the line 30 is directed towards the effluent tank 24. On the other hand, if it is required to increase the boron concentration of the cooling fluid (FIG. 7), the fluid in the line 29 is directed towards the effluent tank 24 and the fluid in the line 30 towards the bed 19' (or 19). In this case, concentrated boron solution contained in the tank 43 is injected via the line 60. The plant described above has numerous advantages. It allows large volumes of cooling fluid to be continuously and rapidly processed within a wide range of concentration of the soluble poison. The same plant is used to increase or decrease the concentration of the poison. Start-up and stoppage of the plant, as well as reversal of operation, may be automated and performed very rapidly. Moreover, processing of the cooling fluid does not lead to the formation of a large volume of effluents, since the electrodialysis modules 20 and 21 do not increase the volume of the effluents, but have the sole effect of concentrating or diluting the fluid contained in the closed circuit 23 connected to the effluent tank 24. However, if the reverse osmosis apparatus 27 is used solely or in combination with the electrodialysis modules 20 and 21, the volume of effluents stored in the tank 24 will increase. To this end, in order to process the effluents, the line 34 conveys the effluents removed from the tank 24 (FIG. 8) first of all into the mixed-bed demineralizer 36, then into the module 37 for thermal dissociation of the ammonium borates, and finally into the tank 38 inside which the effluents are stored until there is a sufficiently large volume thereof to start the processing operation. Processing is performed by the two reverse osmosis apparatuses 39 and 40 mounted in series, and the pressurization of these apparatuses and the circulation of the fluid is ensured by the two pumps 41 and 42. The apparatuses 39 and 40 enable the effluents to be separated into a concentrated solution of poison, stored in 43, and of water stored in 44. The concentrated solution of poison leaving the apparatus 40 is recycled towards the tank 38 so as to admit into the tank 43 only a concentrated solution with a constant high level, preferably in the region of 7000 ppm. Moreover, the line 45 (FIG. 8) makes up for the ammonia solution losses which inevitably occur in the electrodialysis modules 20 and 21, the ammonia solution escaping towards the effluent tank 24. The readjusting line 45 removes the effluents after degassing thereof, i.e., at the outlet of the module 37. The module 37 also enables the ammonia to be separated from the other gases dissolved in the effluents, such as hydrogen or xenon. The ammonia solution obtained at the outlet of the tank 49 is injected into the electrodialysis modules 20 and 21 when this is necessary, by means of the pump 50. The invention is not limited to the embodiment described above solely by way of example. Thus, the number of electrodialysis modules is not absolutely restrictive and depends in particular on the desired variations in concentration and the soluble poison concerned. The beds of cationic resins 19 and 19' may be replaced by zeolites, in the case where the additive intended to promote the dissociation of the boric acid is not ionized, the beds of the cationic resins being used solely for an ionized additive. The beds 19 and 19' may also be replaced by an ultrafiltration device; the additive injected in this case consists of basic compounds with large molecules, for example of the amine type, and, upon returning from the electrodialysis modules, the cooling fluid passes through an ultrafiltration module which retains the basic compounds. Moreover, the effluent processing circuit C may also use electrodialysis modules rather than reverse osmosis modules. It is sufficient to replace the modules 39 and 40 by one or more electrodialysis modules mounted in series or in parallel.
	description	In all the figures of the drawing, sub-features and integral parts that correspond to one another bear the same reference symbol in each case. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown a fuel assembly containing a multiplicity of fuel rods 4 to 6 which, in the operating state, extend vertically between a lower rod-holding plate 1 and an upper rod-holding plate 3. The rod-holding plates 1 and 3 are provided with non-illustrated coolant passages. The fuel rods 4-6 are disposed parallel to one another and are clamped in spacers 11 to 18. While the fuel rods of normal length do not rest or only rest loosely on the lower rod-holding plate 1, the part-length rods are securely anchored in the rod-holding plate 1 by their lower ends. A fuel assembly channel 2 (only partially illustrated in FIGS. 1 and 3) which is open at the top and bottom encloses the bundle of fuel rods 4 to 6 and forms a closed shroud for a liquid coolant which enters through the lower rod-holding plate 1. On its way through the fuel element channel 2, the coolantxe2x80x94preferably waterxe2x80x94is heated by the fuel rods 4 to 6 and begins to evaporate, so that a mixture of liquid coolant and coolant in vapor form takes up the heating capacity of the fuel rods 4-6 in an upper region of the fuel assembly. The mixture of liquid and vapor has a larger volume than the pure liquid. In order nevertheless to avoid an undesirably high flow velocity with a low mass throughput, it is known per se to shorten some of the fuel rods, so that the clear passage cross section in the upper region of the fuel assembly channel 2 is greater than in the lower region. In configuration terms, the spacers 11 to 18 are divided into a lower group A (12 to 15) and an upper group B (15 to 18), a distances between the spacers 12 to 15 in group A being identical to one another. It may be that just two rod lengths (full length and a single part length) will be sufficient, but two part lengths are more advantageous (and more complex). Accordingly, distances between the spacers 15 to 18 in the upper group B are shorter than in the lower group A, in particular becoming shorter the further up they are. To more precisely achieve values that are required in order to optimize a maximum transition power, the fuel rods 5 and 6 are shortened by different extents and in some cases end above the spacer 14 and in other cases end directly above the spacer 15 which forms the boundary between the spacer groups A and B. As a result, the configuration is further away from the maximum power for transition to boiling, so that the configuration according to the invention, in configuration terms, provides an additional optimization parameter. FIG. 2 likewise shows the fuel assembly with eight spacers 11 to 18 which, in configuration terms, are divided into a group A and a group B. For the sake of clarity, the fuel assembly channel 2 is not shown in FIG. 2. To provide a configuration situation which differs from that used for the configuration shown in FIG. 1, in this case fuel rods 7 are also provided. The fuel rods 7 are shortened to a lesser extent than the fuel rods 5 and 6. Further variations on fuel assemblies configured in accordance with the invention are shown in FIGS. 3 and 4. In these two solutions, in each case nine spacers 11 to 19 are provided, three different fuel rod lengths being used in FIG. 3, as in FIG. 1, and four different fuel rod lengths being used in FIG. 4, as in FIG. 2. FIGS. 1 to 4 in each case show only one of, for example, 9 to 11 rows of fuel rods positioned one behind the other, all of which can be differently equipped with fuel rods of different lengths. All the above measures together or on their own allow the maximum power for transition to boiling to be optimized over a wide range. The configurations shown in FIGS. 1 to 4 are advantageous in particular because they allow unrestricted measures for segregating the liquid phase and the vapor phase in the upper region of the fuel assembly. Devices on the spacers 11 to 19, which are illustrated in FIGS. 5 and 6, are used for this purpose. FIG. 5 shows, on a greatly enlarged scale, one of the spacers 11 to 19 which contains metal strips 20 which cross one another at right angles and penetrate through one another. The metal strips 20 form approximately square mesh openings for accommodating the fuel rods 4 to 7 which are clamped securely in mesh openings by lugs 21 and springs 22. As well as crossing points of the metal strips 20, in each case upwardly directed, laterally bent-off sheet-metal vanes 23 are provided, of which in each case those which are disposed next to the same crossing location 25 act in the same direction on a partial flow of the coolant which is flowing through the spacers 11 to 19 parallel to the fuel rods 4 to 7, so that a turbulent impulse D is imparted to a partial flow 26. The resultant rotary movement generates centrifugal acceleration in the partial flow 26, which forces the liquid phase of the coolant onto the fuel rods 4 to 7 and increases the cooling of these rods. In principle, the spacer as shown in FIG. 6, in which the mesh openings which are provided to receive the fuel rods 4 to 7 are formed by hollow cylindrical sleeves 24, which likewise bear the sheet-metal vanes 23 and impose the turbulent impulse D on the partial flow 26 of coolant flowing past them, acts in the same way. The vanes 23 shown are provided beneath the very top spacer on some (preferably all) spacers 23 belonging to group B, but are not present in group A or are only much smaller in this group. In this way, the hydraulic stability is increased, since compared to the upper part the pressure drop in the lower part of the fuel assembly should not be too low. It is also possible, with a view to achieving a low pressure loss in the upper part B, to dispense with springs, lugs or similar holding elements for the fuel rods on one or more spacers (e.g. at position 17 in FIGS. 1 to 4). FIGS. 5 and 6 show a spacer region, the mesh openings of which all have in each case one fuel rod passing through them (except for the positions which are taken up by a water tube). However, above the part-length fuel rods there are mesh openings through which no fuel rod passes. In this case, the configuration of the vanes 23 is advantageously unchanged, and at any rate the vanes still do not project into the region of the area which lies in a rectilinear continuation of the part-length fuel rods illustrated, i.e. into the area which is formed above the part-length fuel rods. However, springs, lugs or similar supports for the fuel rods may be absent in these mesh openings. In boiling water fuel assemblies, a configuration with at least one coolant tube is advantageous, in order to ensure that sufficient liquid moderator (cooling water) is present in the center of the fuel assembly even in the vapor/liquid zone of the fuel assembly. This is precisely the effect achieved by fuel rods whose length is shortened to from half to ⅔ of the normal fuel rod length. Hitherto, it has been assumed that, in the regular pattern in which the fuel rods are distributed across the cross section of the fuel assembly, all positions that are adjacent to the coolant tube configuration or a part-length fuel rod must be occupied by fuel rods of full length. The part-length fuel rods PL are disposed according to this rule in FIGS. 7 and 8. However, it is advantageous if at least a plurality of fuel rods which are directly adjacent to the coolant tube configuration are occupied by shortened fuel rods, as illustrated using fuel rods PLxe2x80x2 in FIGS. 7 and 8. FIG. 7 shows an example with two D-shaped tubes 30, 31 as the coolant tube configuration 30, 31, while FIG. 8 shows a single coolant tube 32 which is square in cross section as the coolant tube configuration. In both cases, the coolant tube configuration 30-32 covers a plurality of fuel rod positions and there is no fuel rod in the center of the fuel assembly. In a reactor core, a corner of a fuel rod channel 33 serves as a guide for a cross-shaped control rod 34, while the diametrically opposite corner is adjacent to an instrumentation tube 35 for measuring probes. This configuration of the measuring probes 35 and the control rod 34 in the gaps between outer surfaces of adjacent fuel assembly channels causes a relatively great width of the water-filled gaps, the gaps that carry the control rods often being wider than the other gaps. This leads to an uneven distribution of absorption material, moderator and fuel and therefore to inhomogeneity in the neutron flux and the power and burnup of the fuel rods. To achieve good utilization of the fuel at a sufficient distance from the power for transition to boiling, i.e. cooling which is adapted to the inevitable inhomogeneity, it is advantageous if the coolant tube configuration is not central, but rather is offset diagonally away from the control rod toward the opposite corner. This is achieved simply if the coolant tube configuration 30-32 is positioned correspondingly eccentrically in the pattern of the fuel rods. FIGS. 7 and 8 show some of the fuel rods 36 of full length and the webs of spacers 37, which are held at a predetermined distance from the inner surfaces of the fuel assembly channel 33 by suitable distancing elements 38, 39 which are disposed on the outer web of the spacers. FIGS. 7 and 8 show that advantageously the entire pattern of fuel rods disposed around the coolant tube configuration is also held in the same eccentric configuration in the channel. Accordingly, the distancing elements 38 produce a wider gap between the outer web of the spacer and the channel inner surface than the distancing elements 39. Furthermore, FIG. 8 shows that the shortened fuel rods are also advantageously distributed in a similarly eccentric manner across the fuel assembly cross section. A diagonal DG indicates the direction in which the coolant tube configuration and the entire pattern of fuel rods are offset with respect to the center axis of the fuel assembly channel. In the diagonal half of the channel cross section which is disposed symmetrically about the diagonal DG and is adjacent to the instrumentation tube 35 (i.e. is delimited by the second diagonal DGxe2x80x2 and includes at least the greater part of the coolant tube configuration), there are more shortened fuel rods PL and PLxe2x80x2 than in the other half which is disposed symmetrically about the diagonal DG and is adjacent to the control rod 34. Moreover, FIG. 8 shows that in the half which is delimited by the diagonal DGxe2x80x2, the walls of the water passage configuration 32 are advantageously closer to the adjacent fuel rods (e.g. PGxe2x80x2) compared to the fuel rods in the other half. In these figures, PL denotes fuel rods which are shortened to a lesser extent than the fuel rods PLxe2x80x2, i.e. the fuel rods PLxe2x80x2 have the shortest length. In accordance with FIG. 8, a plurality of fuel rods which are directly adjacent to the coolant tube configuration 32 are advantageously shortened to the shortest length, and at least most of the fuel rods with the shortest length are situated in the corresponding diagonal half of the channel cross section which is adjacent to the instrumentation tube 35. These rules relating to the configuration of the coolant tube configuration, of the fuel rods and in particular of the part-length fuel rods may advantageously be used even with boiling water fuel assemblies whose spacers are disposed at constant axial distances from one another. However, a particularly advantageous configuration results if the spacers in a lower region are at a constant distance from one another, but the spacers in the upper region are at a mean distance from one another which is shorter than the constant distance in the lower region.
054188237	description	DETAILED DESCRIPTION OF THE DRAWINGS In accordance with an exemplary embodiment of the present invention, the eddy-current probe arrangement and technique used to obtain impedance measurements from the outside of a cladding tube employs a pair of wire coil windings electrically interconnected to form a differential coil pair, as illustrated in FIG. 1. The differential coil pair are electrically interconnected into a impedance bridge and oscillator circuit arrangement that forms conventional electronic impedance measuring equipment (not shown). One of the coils (10) is used as the eddy-current generating "probe" in the testing of cladding tubes by passing the specimen cladding tube (11) undergoing test through the central opening (12) in the coil windings. The remaining coil (14) is contained in a separate magnetically shielded housing (indicated by dotted line 16) and surrounds a short exemplary segment (18) of a fuel rod cladding tube that is used as a "dummy" impedance load. Coils 10 and 14 are preferably constructed such that the diameter of central opening 12 is on the order of a few mils larger than the outside diameter of the nominal type of tube anticipated being tested. It is desired to match probe coil inner diameter as closely as possible with cladding tube outer diameter to minimize any "fill" errors (analogous to "lift-off" errors in conventional eddy-current surface probes) in the measured impedance. Accordingly, different specimen tube types may require a correspondingly different coil-pair size. Dummy impedance load tube segment 18 is maintained within the coil throughout the testing process and functions primarily as a nominal load impedance necessary to maintain the impedance bridge circuit (not shown) in a near-balanced condition. Accordingly, a dummy load segment is chosen which has dimensional and physical properties (e.g., electrical conductivity and magnetic permeability) that are substantially similar to the particular type of fuel cladding tube to be tested. Since the two coils are connected in a bridge circuit arrangement, any difference in electrical impedance between the two coils creates a measurable impedance imbalance which is indicative of a difference in the respective physical properties (e.g., a dimensional or compositional difference) of the dummy load cladding segment and the specimen cladding tube under test. Ideally, if the dummy load cladding tube segment is compositionally identical to the type of specimen fuel rod cladding tube undergoing a test, any impedance imbalance measured in the bridge circuit coils is caused predominately by dimensional differences between the dummy impedance load tube segment and the fuel rod cladding tube. To quantify the above dimensional difference, a carefully constructed "calibration standard" cladding tube is utilized to correlate measured impedance differences to actual physical dimensional differences. In an exemplary embodiment of the invention, primary calibration standard cladding tube 20 consists of sixteen distinct segments 22 as shown in FIG. 2. These segments are joined together to form a single long calibration standard tube. Each segment has approximately the same length as the dummy load reference segment (e.g., two inches) and is preferably not much larger than the length of coils 12 and 14. In addition, each segment 22 has the same liner thickness but different outside and inside diameters which are known to a high precision. Table 1 below lists the sixteen different segment numbers of the primary calibration standard along with exemplary outside (OD.sub.i) and inside (ID.sub.j) diameters for each segment. TABLE 1 ______________________________________ SEGMENT NO. OD ID ______________________________________ 1 .4425 .3790 2 .4400 .3790 3 .4375 .3790 4 .4350 .3790 5 .4425 .3815 6 .4400 .3815 7 .4375 .3815 8 .4350 .3815 9 .4425 .3840 10 .4400 .3840 11 .4375 .3840 12 .4350 .3840 13 .4425 .3865 14 .4400 .3865 15 .4375 .3865 16 .4350 .3865 ______________________________________ Table 2 below illustrates exemplary preferred inside and outside diameter combinations for each numbered segment of the primary calibration standard in the form of a dimensional matrix. Table 2 below illustrates exemplary preferred inside and outside diameter combinations for each numbered segment of the primary calibration standard in the form of a dimensional matrix. TABLE 2 ______________________________________ 1) OD.sub.1 /ID.sub.1 2) OD.sub.2 /ID.sub.1 3) OD.sub.3 /ID.sub.1 4) OD.sub.4 /ID.sub.1 5) OD.sub.1 /ID.sub.2 6) OD.sub.2 /ID.sub.2 7) OD.sub.3 /ID.sub.2 8) OD.sub.4 /ID.sub.2 9) OD.sub.1 /ID.sub.3 10) OD.sub.2 /ID.sub.3 11) OD.sub.3 /ID.sub.3 12) OD.sub.4 /ID.sub.3 13) OD.sub.1 /ID.sub.4 14) OD.sub.2 /ID.sub.4 15) OD.sub.3 /ID.sub.4 16) OD.sub.4 /ID.sub.4 ______________________________________ where, OD = outside diameter, ID = inside diameter, OD.sub.1 > OD.sub.2 > OD.sub.3 > OD.sub.4, and ID.sub.1 < ID.sub.2 < ID.sub.3 < ID.sub.4 Upon passing the primary calibration standard through the eddy-current probe coil, dimensional differences between the segments in the calibration standard and the dummy load reference tube segment will cause a unique measurable impedance imbalance for each of the sixteen segments. These impedance difference measurements are correlatable to the known dimensions of the primary calibration standard segments. Mathematically speaking, since electrical impedance consists both of real and imaginary components, namely the inductive reactance and the resistance, it may be represented diagrammatically by a point plotted in the real/imaginary complex plane. Consequently, a conventional eddy-current impedance measurement instrument having a cathode ray tube (CRT) display arrangement, which is configured to depict impedance values using horizontal and vertical axes of the display to represent the complex impedance plane, may be used to display the unique impedances of each of the sixteen segments of the calibration standard. A CRT display of measured impedances for all of the segments of calibration standard 20 in this manner forms an "impedance matrix", as illustrated in FIG. 3. The different impedance values for the segments appear as illuminated spots on the CRT display screen (represented in FIG. 3 by rectangular spots 32). For example, each displayed impedance corresponds to a particular inside and outside diameter combination listed in TABLE 2. Since the inside and outside diameter dimensions of a tube affect its measured impedance, a CRT display in the impedance plane of a particular cladding tube under test can be utilized to determine the inside and outside diameter of the tube by comparison (either visually or electronically) of its measured impedance with the previously measured impedances of the calibration standard. More specifically, the cathode ray tube display in a convention eddy-current instrument can be configured to display measured input voltages that are calibrated with respect to indicia on the face of the display in either or both of the horizontal and vertical directions. Consequently, the display of an impedance matrix can be oriented (e.g., rotated) so that horizontal axis voltage indicia correspond to differences in tube outside diameter and vertical axis voltage indicia correspond to differences in tube inside diameter or vice versa. In accordance with a preferred embodiment of the present invention, the unit difference in vertical positions of each displayed impedance spot 32 is calibrated to the unit difference in inside diameters between segments of the calibration standard. For example, referring again to FIG. 3, E.sub.1 represents a difference in CRT vertical display voltage produced by a difference in the measured impedances of two separate primary calibration standard segments. The impedance matrix, however, is purposely oriented on the display such that E.sub.1 also directly corresponds solely to a difference in inside diameters of the calibration segments. During the testing of cladding tubes, when a specimen tube is inserted into the circumferential measuring coil, the resulting detected impedance measurement will fall somewhere within the range of impedance values forming the calibration standard impedance matrix. Referring again to FIG. 3, circular mark 34 represents the impedance value of an exemplary specimen cladding tube measured at some arbitrary point along its length. Since each of the displayed impedance values, 32, making up the primary calibration standard impedance matrix are correlated to known outside and inside tube diameters of segments in the primary calibration standard, the outside and inside diameter of a specimen cladding tube can be uniquely determined for any desired point along its length from its impedance relative to the known calibration standard impedance values. However, this approach to determining diameter dimensional information is only accurate to the extent that the specimen cladding tube liner thickness is identical to the liner thickness of the calibration segments in the primary calibration standard. For example, due to relative differences in electrical resistivity of the Zircaloy-2 tube metal and the pure zirconium liner (72 micro-ohm-centimeters vs. 40 micro-ohm-centimeters, respectively) the overall measured conductance of induced eddy-current (and hence the impedance) is dependant on the relative proportions of Zircaloy-2 and zirconium in the cladding tube. Accordingly, if the cladding liner thickness is less than that of the calibration standard segments, the overall conductance (and correspondingly the overall impedance) will be less than that of the calibration standard and, therefore, will produce an erroneous measurement indicating an inside diameter greater than the actual inside diameter of the cladding tube. Conversely, a cladding tube liner thickness greater than that used in the calibration standard segments will cause an erroneous measurement indicating an inside diameter less than the actual value. However, in accordance with the present invention, if liner thickness can be separately calibrated to measured impedances and the "actual" cladding tube inside diameter determined precisely via ultrasonic (or other) techniques, then the difference between the actual inside diameter and the inside diameter as determined electromagnetically can be used to deduce the thickness of a cladding tube liner. Thus, in accordance with a preferred embodiment of the present invention, a second calibration standard cladding tube (not shown) is used to calibrate the effect that various liner thicknesses have on the overall measured impedance. Accordingly, the second calibration standard consists of multiple tube segments (e.g., seven in the presently preferred embodiment) having various known liner thicknesses but all having the same inside and outside tube diameters. FIG. 4 shows a representation of a CRT display of exemplary measured impedances (spots 32 and 42) for the two different calibration standards. The second calibration standard produces a vertical array of spots 42 on the CRT display distinct from spots 32 comprising the impedance matrix of the primary calibration standard. Spots 42, represented in FIG. 4 by seven small circles, indicate the separate measured impedances for each of the known different liner thicknesses of the seven segments of the second calibration standard. Since a separation in the displayed impedances of the second calibration standard correlates to liner thickness only, the unit difference in vertical positions of each displayed impedance spot 42 is, thus, calibrated to the unit difference in liner thicknesses. For example, referring again to FIG. 4, E.sub.2 represents a difference in CRT vertical display voltage produced by a difference in the measured impedances of two separate segments of the second calibration standard. The measured impedances of the second calibration standard, however, are oriented on the display such that E.sub.2 also corresponds to a difference in liner thicknesses of the calibration segments. In accordance with a method of the present invention, the actual liner thickness, T.sub.b, for a particular fuel rod cladding tube undergoing test, can be calculated using the known thickness of the liner in the sixteen segments of the first calibration standard, the difference between ultrasonic and electromagnetic measured values for inside diameter, and the impedance values (represented as voltages) from the two calibration standards, by using the following formula: EQU T.sub.b =T.sub.std +E.sub.1 /E.sub.2 .times.(ID.sub.ut -ID.sub.ec) (Equ. 1) where, T.sub.b =Liner (or "barrier") thickness PA1 T.sub.std =Thickness of liner in the segments of the primary calibration standard PA1 E.sub.1 =Change in vertical display voltage per unit change in inside diameter as measured by electromagnetic subsystem PA1 E.sub.2 =Change in vertical display voltage per unit change in liner thickness as measured by electromagnetic subsystem PA1 ID.sub.ut =Inside diameter determined ultrasonically, and PA1 ID.sub.ec =Inside diameter determined by electromagnetic subsystem. Referring now to FIG. 5, a block diagram illustrates the combined eddy-current and ultrasonic testing system in accordance with an exemplary embodiment of the present invention. Arithmetic unit (or computer) 501, obtains measurement information separately from both ultrasonic subsystem 503 and electromagnetic subsystem 502. Electromagnetic measuring subsystem 502 includes differential coil pair eddy-current impedance probe arrangement 508 (depicted in FIG. 1) connected to a data acquisition system (DAS) consisting of a conventional electrical impedance measurement bridge with associated signal processing and A/D (analog-to-digital) conversion circuits 510. A display device, for example, CRT 512, may also be connected to the DAS. DAS electromagnetic signal processing circuitry 510 may also include a computer or microprocessor programmed for performing conventional signal analysis and measurement interpolation calculations. For example, inputs to DAS 500 are interpolated from the correlation of "horizontal" voltage to outside diameter and "vertical" voltage to inside diameter using conventional LaGrainge interpolation calculations by using DAS processing circuitry 510. Alternatively, this correlation may be performed from digitized measurement data by computer 501. Ultrasonic subsystem 503 basically consists of conventional ultrasonic testing equipment 505 connected either directly or indirectly, through conventional a/d conversion and signal processing circuits 507, to computer 501. Ultrasonic testing equipment 505 may be any conventional system or apparatus for accurately measuring the wall thickness of a metal tube through ultrasonic techniques, such as, for example, the apparatus disclosed in commonly assigned U.S. Pat. No. 5,063,780, issued Nov. 12, 1991 to Landry. Digitized impedance data acquired from electromagnetic subsystem 503 and digitized diameter measurement data from ultrasonic measurement subsystem 502 is stored in memory 504 and subsequently retrieved by computer 501 to calculate thickness, T.sub.b, of a cladding tube liner in accordance with Equ 1 above. Computed liner thickness data may then be stored in memory 504 or printed or displayed via output devices 506. Referring now to FIG. 6, a schematic flowchart is shown that illustrates steps of an exemplary program executed by an arithmetic unit or computer 501 for acquiring measurement data and calculating liner thickness in accordance with a preferred embodiment of the invention. Steps S1 and S2 of the flowchart provide for the setting up of reference impedance values in memory 504 from the two calibration standards for subsequent use in computing Equ 1. Steps S3 through S8 provide for determining liner thickness from both impedance and ultrasonic data acquired from the specimen under test. More specifically, in Step 1, a known value of liner thickness, T.sub.std, for the primary calibration standard is input via conventional input device 506 and stored in memory 504. In addition, impedance values from primary calibration standard 20 are obtained from electromagnetic measurement subsystem 502 and also stored in memory 504. In step S2, impedance values from the second calibration standard are, likewise, measured and stored. (step S2 may be performed prior to step S1 or as part of the same step). In step S3, the input data from the eddy-current subsystem (502) is correlated (for example, by performing LaGrainge interpolations) to the measured values of outer diameter and inner diameter from the primary standard and the known liner thicknesses from the secondary standard. Next in step S4, a specimen is tested electromagnetically using subsystem 508 and the measured impedance is stored in memory 504. If desired, impedances of the calibration standards and the test specimen may also be displayed as measured during these steps via CRT 512. In step S5, conventional eddy-current analysis and computational techniques are performed to interpolate/calculate an inside diameter value for the tested specimen based on its measured impedance and the stored reference impedance values of the primary calibration standards. This value, ID.sub.ec, is then stored in memory 504. Next, in step S6, the inside diameter of the same specimen is tested ultrasonically using subsystem 503 and the resulting diameter data, ID.sub.ut, is also stored in memory 504. (Step S5 may be performed prior to step S3 instead, if desired). In step S7, the liner thickness, T.sub.b, is calculated according to Equ 1 from the stored impedance and ultrasonic data. This information is then stored in memory 504 and/or provided to a printer or other output devices 506, as indicated in step S8. Finally, in step S9, measuring of other cladding tube specimens to determine liner thickness is conducted in the same manner or else the testing process is terminated. 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. A person of ordinary skill in the art should appreciate that the requisite correlated reference impedance values could be obtained by utilizing a single "combined" calibration standard having the calibrated segments of both the primary and second calibration standard tubes instead of using two separate calibration tubes. Moreover, a person of ordinary skill in the art should realize that various utilizations of the invention may not require the use of all sixteen impedance references of the primary standard or all seven impedance references of the second calibration standard and that the calibration of reference impedance values would not necessarily have to correspond to any particular CRT display axis or orientation. In addition, although the presently preferred embodiment utilizes ultrasonic techniques for acquiring the inside diameter of a specimen, it is envisioned that other systems or means for acquiring "actual" inside diameter information could also be used without departing from the scope of the invention.
039376532	claims	1. A nuclear reactor diagrid combination for supporting the core of a nuclear reactor in which a liquid sodium coolant is circulated within a vessel containing the core and diagrid, said diagrid being constituted by a box structure having end-walls formed by two metallic plates braced by hollow cylindrical support columns which are suitably disposed at intervals within the box structure on a uniform pitch, the upper end plate being flat and horizontal and the axis of the reactor core being vertical, said diagrid being provided with means for applying the box structure against a support which forms part of the reactor vessel, wherein the lower end plate of said box structure has the shape of an upwardly convex spherical segment which terminates at its periphery in a flat rim and wherein the upper flat end plate has a circular flange which is parallel to the peripheral rim of the lower plate, said rim and said flange being braced with respect to each other, the flange of the upper end plate resting on a ring-girder supported by the vessel which provides the box structure with a peripheral side restraint and the flat rim of the lower end plate being supported by a flat circular flange formed on a shell element rigidly fixed to the reactor vessel, wherein the shell element has a conical shape and is in the line of extension of the spherical segment of the lower end plate of the box structure, and wherein the ring-girder is a hollow metallic torus opening laterally toward the box structure, said torus being a manifold for supplying the diagrid with the coolant sodium. 2. A combination in accordance with claim 1, wherein the flat rim of the lower end plate and the parallel flange of the upper end plate are braced by small vertical stiffening columns. 3. A combination in accordance with claim 1 wherein the ring-girder is provided with a top horizontal bearing surface of circular shape having substantially the same radius as the flange of the upper end plate, said flange being applied against said bearing surface by means of coaxial circular grooves, the interengagement of which forms a labyrinth seal against the coolant sodium. 4. A combination in accordance with claim 3, wherein the positioning of the flange of the upper end plate with respect to the circular bearing surface of the ring-girder is carried out by means of studs having the shape of sectors which are carried by the flange and engaged in recesses formed in the bearing surface. 5. A combination in accordance with claim 1, wherein a flexible seal is interposed between a shouldered portion of the flat circular flange of the shell element and the flat rim of the lower end plate of the box structure.
047675942	claims	1. In a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating said hot pool and said cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on said lower pressure sodium cold pool and an outlet to a reactor core; said reactor core for heating the sodium and discharging the sodium to said reactor hot pool; a heat exchanger for receiving sodium from said hot pool, and removing heat from the sodium and discharging the sodium to said lower pressure cold pool; the improvement across said reactor vessel liner comprising: a jet pump having a venturi installed across said reactor vessel liner, said jet pump having a lower inlet from said reactor vessel cold pool across said reactor vessel liner and an upper outlet to said reactor vessel hot pool; a pumping fluid inlet from the high pressure discharge of said main sodium reactor coolant pumps, said inlet having a high velocity outflow to said jet pump venturi whereby upon normal operation of said main sodium reactor coolant pumps, the jet pump maintains a pressure differential from said lower pressure cold pool to said hot pool and upon failure of said main sodium reactor coolant pump, the jet pump permits immediate sodium backflow from the hot pool to the lower pressure cold pool across the reactor vessel liner flow gap to establish immediate cooling of residual reactor heat through the reactor vessel wall. a reactor core for heating sodium discharged from said pumps, said reactor core having an inlet communicated to the outlet of said pump and an outlet to said reactor hot pools; a heat exchanger for receiving sodium from said hot pool, removing heat from the sodium and discharging the sodium to said cold pool; a jet pump having a venturi installed across said reactor vessel liner, said jet pump having a lower inlet from said reactor vessel cold pool and an upper outlet to said reactor vessel hot pool; a pumping jet having an inlet from the high pressure discharge of said main sodium reactor coolant pump and having an outflow to the jet pump venturi whereby upon operation of said main sodium reactor coolant pumps, the jet pump maintains a pressure differential from said lower pressure cold pool to said hot pool, and upon -oss of normal heat removal paths and associated shutdown of said main sodium reactor coolant pumps, the jet pump permits immediate backflow from the hot pool to the lower pressure cold pool across the reactor vessel flow gap to establish an immediate sodium cooling flow path for residual reactor heat removal through the reactor vessel. 2. The invention of claim 1 and wherein said jet pump outlet is parallel to and well below the surface of liquid sodium in said reactor. 3. The invention of claim 1 and wherein said sodium cooled reactor includes a plurality of said jet pumps. 4. The invention of claim 3 and wherein said sodium cooled reactor has a plurality of main sodium reactor coolant pumps and each of said pumps has a plurality of said jet pumps. 5. A sodium cooled reactor comprising in combination: a reactor hot pool; a lower pressure reactor cold pool; a reactor vessel liner separating said hot pool and said cold pool interior of a reactor vessel and immediate said reactor vessel side walls; a main sodium reactor coolant pump having a suction on said low pressure cold pool and an outlet; 6. The invention of claim 5 and wherein said jet pump discharge to said reactor vessel hot pools is parallel to the surface of said reactor vessel hot pool. 7. The invention of claim 5 and including a plurality of jet pumps connected to said sodium cooled reactor.
052020851	summary	BACKGROUND OF THE INVENTION The present invention relates to a fuel assembly, a reactor core, and a method for loading of the fuel assembly, especially, relates to the preferable fuel assembly being loaded in boiling water type nuclear reactors (hereinafter called BWR) for increment of reactor shut down margin, improvement of fuel economy, and maintenance of reactivity control, and relates to the preferable reactor core loaded with the fuel assemblies and the preferable method for loading of the fuel assembly. A conventional fuel assembly which is used in a BWR is generally composed of a plurality of fuel rods and one or a plurality of water rods which are arranged in a channel box by being supported at an upper end and a lower end of the fuel rod and the water rod with an upper tie plate and a lower tie plate. During operation of the reactor, slightly unsaturated cooling light water enters from a hole of the lower tie plate into an interval among the fuel rods in the fuel assembly, and flows out from a hole of the upper tie plate as vapor-liquid two phase flow after being heated by the fuel rods and boiled during flow from lower portion to upper portion of the fuel rod interval. As the result, void fraction of the coolant increases from 0% at the lower portion to about 70% at the upper portion of the fuel assembly. Consequently, the ratio of hydrogen atoms to heavy metal atoms; that is, the ratio of moderator to fuel (H/U ratio), which is an important factor for determining of nuclear characteristics of the fuel assembly alters remarkably depending on a position in an axial direction. On the other hand, it is necessary to install control rods and instrument tubes for neutron detectors exterior of the channel box in the BWR and, therefore, such intervals (hereinafter called water gap) as to enable the above mentioned rods and tubes be inserted are provided between the fuel assemblies. The water gap is filled with saturated water and, consequently, effects of the saturated water existing in the water gap to the fuel rods in the fuel assembly are different depending on whether the fuel rods locate at periphery of the fuel assembly (a region near the water gap) or central region of the fuel assembly. That is, the peripheral region of the fuel assembly near the water gap has larger H/U ratio than the central region. Accordingly, such H/U ratio as an important factor for determining the nuclear characteristics of the fuel assembly differs depending on the radial position in the fuel assembly. The H/U ratio is a parameter to determine an average energy of a neutron. As the ratio becomes larger, the average energy of the neutron becomes smaller (softer neutron spectrum), and the nuclear fission reaction with nuclear fissile material is enhanced. Concurrently, the softening of neutron spectrum increases the neutron absorbing reaction by the moderator (light water as coolant) as well as the nuclear fission reaction. Accordingly, there is an optimum H/U ratio in view of fuel economy. Moreover, fuel rod power generation which depends on the reactivity of nuclear fission is determined by the H/U ratio. That is, in view of thermal margin and controlability of excess reactivity of the fuel assembly, it is necessary to consider the H/U ratio. On the other hand, with related to conventional nuclear reactors, extension of an operation cycle of the reactor and high burn up of fuel are considered for increasing of a plant utilization factor and effective utilization of uranium resources. For increasing of discharged burn up of the fuel assembly, it is necessary to increase enrichment of the fuel assembly. The increment of the fuel enrichment influences the optimum H/U ratio. Further, the extension of loading period of the fuel assembly in the reactor means that the fuel is effected under different H/U ratios for a long period in the reactor, and the above mentioned influence of the H/U ratio is enhanced. In regard to improvement of distribution of the H/U ratio in a radial direction and an axial direction of the fuel assembly, there are such methods as enlarging of a saturated water region at a necessary portion and regulation of distribution of the nuclear fissile material. The former is a method for improving the H/U ratio by enlarging the saturated water region at the central and the upper region of the fuel assembly, wherein moderating effect of the neutron is deteriorated. And the latter is a method for improving the H/U ratio in the axial direction by regulation of loading quantity of the fuel. For example, in JP-A-62-211584 (1987), a method to increase horizontal cross sectional area at the upper region in the axial direction of the fuel assembly and to arrange a water rod having a horizontal cross section of cruciform at the upper region in the axial direction is proposed. Short length fuel rods are loaded beneath the cruciform protruded region of the cruciform water rod. And, in JP-A-52-50498 (1977), a method to arrange fuel rods having different length in order to form a flow channel of coolant having reversely tapered shape toward the upper region in the axial direction of the center of the fuel assembly is disclosed. In USP-4,968,479, a fuel assembly for achieving high burn up by increasing of fuel enrichment is disclosed. The fuel assembly is composed of a water rod having larger horizontal cross sectional area at the upper region in the axial direction than the area at the lower region and of fuel rods having three kinds of different length in order to reduce an increment of local power peaking accompanying with using of the highly enriched fuel with a burnable poison at the beginning of operation and to optimize a reactivity distribution at the upper and the lower region of the fuel assembly during a designated operation period. The shortest fuel rod is arranged at the position adjacent to the lower small diameter region of the water rod, and contains fuel having equal to or lower enrichment than the fuel assembly average enrichment, the medium length fuel rod contains fuel having equal to the fuel assembly average enrichment, and a part of the longest fuel rods contain fuel having the burnable poison (column 15, line 25-60, FIG. 22, 30B-30D). Further, in JP-A-63-311195 (1988), on a fuel assembly for achieving high burn up by increment of fuel enrichment, an improving method for increasing the reactor shut down margin in considering that the increasing of the enrichment at the upper region of the fuel assembly increases the reactivity of the upper region at the reactor shut down margin is disclosed. The fuel assembly improved by the above described method has two water rods each of which have a large diameter and uniform horizontal cross section in the axial direction and fuel rods, which are arranged adjacent to the two large diameter water rods, containing lower enriched fuel at least at the upper region of the fuel rod than the fuel in other next fuel rods. Other prior techniques relating to the increment of burn up are disclosed in USP-4,229,258, JP-A-63-21589 (1988), and JP-A-64-28587 (1989). In USP-4,229,258, a fuel assembly having higher enriched fuel at the upper region than at the lower region is disclosed. In JP-A-63-21589 (1988), a fuel assembly in which high enriched fuel rods are arranged at the outermost periphery in the horizontal cross section and the enrichment at the lower region in the axial direction of the fuel rods is higher than the enrichment at the upper region is disclosed. In JP-A-6428587 (1989), a fuel assembly in which enrichment of fuel pellets in fuel rods containing enriched uranium and gadolinium is the highest in the fuel assembly and the effective fuel length of the fuel rod is shorter than the length of fuel rods containing enriched uranium but not gadolinium is described. Further, in JP-A-53-43193 (1978), a conventional method in which the saturated water region at the upper region of the fuel assembly is increased by making the thickness of the channel box wall thin at the upper region of the fuel assembly is disclosed. Among above described prior techniques, the conventional method disclosed in JP-A-63-311195 (1988), wherein large water rods having uniform horizontal cross section in the axial direction are used, improves the distribution of the moderator to fuel ratio (H/U ratio) at the upper region of the fuel assembly. Nevertheless, the improvement of the distribution of the H/U ratio at the lower region of the fuel assembly is not considered in the conventional method. In accordance with the prior art wherein the improvement of the H/U ratio distribution in the axial direction of the fuel assembly is aimed at, the characteristics at the lower region of the fuel assembly is sacrificed for the improvement of the H/U ratio distribution in the axial direction and, consequently, the improvement of the H/U ratio distribution in the radial direction at the lower region of the fuel assembly is not sufficient. And the distribution of the moderator and fuel materials (fissile materials and parent materials) in the axial and the radial direction is not considered sufficiently in the prior art. That is, in the methods disclosed in JP-A-62-211584 (1987), JP-A-52-50498 (1977), and USP-4,968,479, the horizontal cross sectional area of water rod or moderator flow channel at the upper region in the axial direction of the fuel assembly is made larger than the area at the lower region in order to increase the H/U ratio at the upper region of the fuel assembly. But the methods have such problems that the cross sectional area of the water rod at the lower region is not sufficient, and flattening of thermal neutron flux distribution is not achieved sufficiently. The problems cause lowering of the fuel economy. Moreover, in the methods disclosed in JP-A-62-211584 (1987) and JP-A-52-50498 (1977), when the enrichment of the short fuel rods arranged in a region which is yielded by decreasing of H/U ratio at the lower region of the fuel assembly is excessively high, fissile materials are generated more at the lower region than at the upper region of the fuel assembly and, consequently, a large peak in power distribution is caused at the lower region of the fuel assembly. Accordingly, there are such problems that stability becomes insufficient and fuel economy is lowered by increasing of average void fraction in the axial direction of the fuel assembly. SUMMARY OF THE INVENTION One of the objects of the present invention is to provide a fuel assembly for improving fuel economy by making the moderator to fuel ratio, which alters depending on the position in the fuel assembly, close to the optimum value as possible everywhere including the lower portion of the fuel assembly, a reactor core using the fuel assembly, and a method of usage of the fuel assembly thereof. Another object of the present invention is to provide a fuel assembly for improving fuel economy and controlability of excess reactivity by optimizing the distribution of fuel materials and moderators in the axial and radial direction of the fuel assembly, a reactor core using the fuel assembly, and a method of usage of the fuel assembly thereof. The feature of the present invention is to provide a fuel assembly comprising a plurality of first fuel rods, a means for moderating which is surrounded with the first fuel rods and have larger horizontal cross sectional area at upper region in the axial direction than the area at lower region, and second fuel rods which are arranged at adjacent to the lower region of the means for moderating and have lower enriched fuel than the horizontal cross sectional average enrichment of the fuel assembly, characterized in that the horizontal cross sectional area at the lower region of the means for moderating is so determined that both of the minimum values of thermal neutron flux distribution and resonance neutron flux distribution in the vertical direction to the longitudinal axis of the fuel assembly are located at an exterior region to the second fuel rod in the vertical direction to the longitudinal axis. The horizontal cross sectional area of the means for moderating at the lower region is preferably larger than sum of the two first fuel rods. The enrichment of the fuel contained in the second fuel rod is preferably lower than 0.7 of the average enrichment at horizontal cross section of the fuel assembly, and more preferably, lower than 0.5 of the average enrichment at horizontal cross section of the fuel assembly. The second fuel rod contains, for example, natural uranium. The second fuel rod is a short length fuel rod which is preferably arranged adjacent to the lower region of the means for moderating, and the length of the fuel rod is preferably less than a half of the effective fuel length of the first fuel rod. Further, the means for moderating is preferably a water rod having wider horizontal cross sectional area at the upper region than the area at the lower region. And the means for moderating is able to be composed of a water rod having an uniform horizontal cross sectional area along the axial direction and a coolant flow channel which surrounds the upper region of the water rod, and is able to be composed of a water rod having an uniform horizontal cross sectional area along the axial direction and a plurality of solid moderating rods which surround the upper region of the water rod. In order to achieve the objects, the present invention provides a reactor core loaded with the above described fuel assemblies. The reactor core preferably has at least a central region and a peripheral region, and the fuel assemblies are arranged more in the central region than in the peripheral region. Further, for achieving the objects, the present invention provides a method of usage of the fuel assemblies characterized in that the fuel assemblies are loaded more in the central region than in the peripheral region at fuel exchange. The distribution of the moderator to fuel ratio in the axial direction of the fuel assembly and in the vertical direction to the axis at the upper region of the fuel assembly are improved respectively by arranging a means for moderating having larger horizontal cross sectional area at the upper region in the axial direction than at the lower region. And, increment of resonance neutron flux absorbing effect and flattening of thermal neutron flux in the radial direction are achieved by arranging the second fuel rods containing lower enriched fuel than the average enrichment in horizontal cross section of the fuel assembly adjacent to the means for moderating, and determining of the horizontal cross sectional area of the means for moderating at the lower region in the axial direction of the fuel assembly so that the minimum value of both the thermal neutron flux distribution and the resonance neutron flux distribution in the vertical direction to the axis are located at an exterior region to the second fuel rod in the vertical direction to the axis and, accordingly, the fuel economy and the controlability of excess reactivity are improved. By the present invention, as the moderator, the fissile material, and the fertile material are optimally arranged in the axial direction and vertical direction to the axis of the fuel assembly, the moderator to fuel ratio comes close to the optimum value at everywhere of the fuel assembly including the lower region and, consequently, the increment of resonance neutron flux absorbing effect and thermal neutron flux flattening in the vertical direction to the axis are able to be utilized, and the effects of improving the fuel economy, the controlability of the excess reactivity, and the thermal margin are realized.
	description	The present invention generally relates to the measurement of the neutron flux in the core of a nuclear reactor. More specifically, the invention according to a first aspect relates to a method for measuring the neutron flux in the core of a nuclear reactor. It is necessary to know the state of the core of a nuclear reactor, in order to guarantee safety (protection of the fuel assemblies) and proper operation of this reactor. For this purpose, it is possible to track several parameters related to the bulk power of the core: the linear power along the axis of the core, the CHFR (Critical Heat Flux Ratio), the axial and radial power disequilibrium, etc. Neutron detectors are used for reconstructing these parameters, because they depend on the neutron flux in the core. Several types of neutron detectors are used: detectors known as “excore chambers”, which are placed outside the core and which give a signal proportional to the mean flux in a quarter of the core; “incore” detectors which are located inside the core. Certain incore detectors are mobile. They are periodically inserted into the core in order to establish a specific image of the power distribution in the core. Other incore detectors are fixed and continuously give a signal representative of the local neutron flux in an area of the core of the reactor. The fixed detectors are permanently subject to irradiation, which in the long run causes a loss of sensitivity of these detectors and a degradation of the accuracy of the corresponding signal. The use of cobalt neutron detectors as fixed incore detectors is known. These detectors behave like passive current generators, the current being generated by nuclear reactions within the detector under the effect of the neutron flux. Activation of cobalt 59 into cobalt 60 in the detector has the long-run effect of deteriorating the useful signal/total signal ratio of this detector, which is detrimental to the accuracy of the measurement. An object of the invention is to provide a method for measuring the neutron flux in the core of a nuclear reactor by means of a cobalt neutron detector with which better accuracy may be obtained. More specifically, the invention provides a method for measuring the neutron flux in the core of a nuclear reactor, the method comprises several steps performed recurrently at instants separated by a period, the method comprising at each given instant the following steps: acquiring a total signal by means of a cobalt neutron detector placed inside the core of the nuclear reactor; assessing a calibration factor representative of the delayed component of the total signal due to the presence of cobalt 60 in the detector; assessing a corrected signal representative of the neutron flux at the detector from the total signal and from the calibration factor; assessing a representative slope of the time-dependent change of the calibration factor between the preceding instant and the given instant; the calibration factor at a given instant being assessed as a function of the calibration factor assessed at the preceding instant, of the slope and of the period separating the given instant from the preceding instant. The method may also have one or more of the characteristics considered below, individually or according to all the technically possible combinations: the neutron detector is a fixed detector; the slope at the given instant is assessed at least as a function of the corrected signal at the preceding instant and of the calibration factor at the preceding instant; the measurement method comprises at each given instant TN a step for assessing a load factor representative of the time-dependent change of the power of the reactor locally around the neutron detector between the preceding instant TN−1 and the given instant, the slope at the given instant being also assessed as a function of the load factor; the load factor at each given instant is assessed by calculating the average power of the reactor locally around the neutron detector between the preceding instant and the given instant, and by dividing said average power by the value of the power of the reactor locally around the neutron detector at the preceding instant; the measurement method comprises at each given instant a step for assessing a load factor representative of the time-dependent change of the power of the reactor between the preceding instant and the given instant, the slope at the given instant being also assessed as a function of the load factor. According to a second aspect, the invention provides a device for measuring the neutron flux in the core of a nuclear reactor, the measurement device comprising at least one cobalt neutron detector placed inside the core of the reactor, and a computer; the computer having means for acquiring recurrently at given instants separated by a period, a total signal by means of the neutron detector; the computer having means for assessing at each given instant a calibration factor representative of the delayed component of the total signal due to the presence of cobalt 60 in the neutron detector; the computer having means for assessing at each given instant a corrected signal representative of the neutron flux at the neutron detector from the total signal and from the calibration factor; the computer having means for assessing at each given instant a representative slope of the time-dependent change of the calibration factor between the preceding instant and the given instant, the means for assessing the calibration factor being capable of assessing the calibration factor at the given instant as a function of the calibration factor assessed at the preceding instant, of the slope, and of the period separating the given instant from the preceding instant. The device may also have one or more of the characteristics below, considered individually or according to all the possible combinations: the neutron detector is a fixed detector; the means for assessing the slope are able to assess the slope at the given instant at least as a function of the corrected signal at the preceding instant and of the calibration factor at the preceding instant; the measurement device comprises means for assessing at each given instant a load factor representative of the time-dependent change of the power of the reactor locally around the detector between the preceding instant and the given instant, the means for assessing the slope being capable of assessing the slope at the given instant also as a function of the load factor; the means for assessing the load factor are capable of assessing the load factor at each given instant by calculating the average power of the reactor locally around the neutron detector between the preceding instant and the given instant, and by dividing said average power by the value of the power of the reactor locally around the neutron detector at the preceding instant. The method, schematically illustrated in FIG. 1, is intended to measure the neutron flux in the core of a nuclear reactor, by means of at least one cobalt neutron detector placed inside the core of the reactor. This type of detector is known under the acronym of Co-SPND (Cobalt-Silver Self Powered Neutron Detector). As shown in FIG. 2, a cobalt neutron detector 1 includes an external sheath 3, and a central emitter 5 positioned inside the sheath 3. The external sheath 3 acts as a cathode, the central emitter 5 acting as an anode. The central emitter 5 before use consists of a material essentially including cobalt 59. The external sheath 3 consists of a electrically conducting material. The central emitter 5 has a diameter of about 2 mm and a length of about 21 cm. Under the effect of the neutron flux, nuclear reactions occur in the central emitter 5, causing transfer of electrons from the central emitter 5 to the external sheath 3. The central emitter 5 is connected to an amplifier device 7 (see FIG. 7) through an electrically conducting cable 8. The device 7 is capable of amplifying and digitizing the current from the emitter 5. The signal collected by the device 7 will be called a total signal (Itot) in the following. The detector 1 then acts as a DC current generator. It does not require any exterior power supply. It is therefore particularly adapted to operation as a fixed detector within a nuclear fuel assembly in the core of the nuclear reactor. The cobalt detectors continuously deliver a signal, the quality of which degrades slowly all along their use in the core. This degradation is expressed by the increasing occurrence of a quasi-static component in the total signal Itot delivered by the detector 1. The significance of this quasi-static component relatively to the useful signal increases over time. Here, by useful signal means the component of the total signal Itot which is proportional to the neutron flux at the detector 1. The quasi-static component, called a delayed component I60, is due to the presence of cobalt 60 in the central emitter 5. Cobalt 60 is formed by activation of cobalt 59 under the effect of the neutron flux. This delayed component I60 is quasi-static in the sense that it is constant over short time intervals, for example over the whole duration of a power transient in the reactor. Such a power transient lasts for a few hours to a few days. The delayed component I60 is therefore de-correlated from the neutron flux, in the sense that it does not vary when the neutron flux varies at the detector 1. The delayed component I60 of the neutron flux thus generates a substantial loss of accuracy of the neutron detector 1 during a power transient. In order to be able to use the signal continuously delivered by the neutron detector 1, it is therefore necessary to correct this signal a posteriori, as illustrated in FIG. 3. This correction is carried out by subtracting from the total signal Itot a calibration factor C_UCO representative of the delayed component I60 due to the presence of cobalt 60 in the detector 1. Thus, the corrected signal is calculated in the following way:I=Itot−C—UCO In this equation, Itot corresponds to the signal continuously delivered by the neutron detector 1, C_UCO is a calibration factor, and I is the corrected signal (useful signal), this corrected signal being representative of the neutron flux at the detector 1. C_UCO is periodically determined by a calculation, as explained later on. The invention benefits from the fact that, as illustrated in FIG. 4, the delayed component I60 of the total signal Itot due to the presence of cobalt 60 in the detector 1, changes slowly over time. FIG. 4 illustrates an exemplary time-dependent change of the delayed component I60, for a given detector 1. The axis of the abscissae is graduated in full power yearly equivalents. It is clearly seen from this FIG. 4 that the delayed component I60 increases versus time, but on the other hand the slope decreases versus time. The delayed component I60 increases slowly and its time-dependent change may be considered as substantially linear over time intervals of the order of one month. On the other hand, the slope has to be re-updated regularly. Moreover, in the invention it is considered that the time-dependent change in the calibration factor C_UCO versus time is governed by an equation formally analogous to the equation of the time-dependent change in the cobalt 60 content in the central emitter 5. This time-dependent change is discretized with a time step adapted so as to benefit from the quasi-linearity of the delayed component I60. More specifically, as illustrated in FIG. 1, the method comprises several steps, performed recurrently at instants TN, said instants being separated by periods ΔT. At each instant TN, the following steps are performed: acquiring a total signal ItotN by means of the cobalt neutron detector 1; assessing a slope SLOPEN−1 representative of the time-dependent change of the calibration factor C_UCO between the preceding instant TN−1 and the given instant TN; assessing a calibration factor C_UCON, representative of the delayed component I60 of the total signal ItotN due to the presence of cobalt 60 in the detector 1, the calibration factor C_UCON at the given instant TN being assessed as a function of the calibration factor C_UCON−1 assessed at the preceding instant TN−1, of the slope SLOPEN−1, and of the period ΔT separating the given instant TN from the preceding instant TN−1; assessing a corrected signal IN representative of the neutron flux at the detector from the total signal ItotN and from the calibration factor C_UCON. The total signal ItotN is the signal delivered by the neutron detector 1. The slope SLOPEN−1 and the calibration factor C_UCON are determined by means of the following equations:SLOPEN−1=α×FchargeN−1×IN−1−β×C—UCON−1 C—UCON=C—UCON−1+SLOPEN−1×ΔT wherein: the index N corresponds to the iteration N (instant TN), IN−1 corresponds to the corrected signal assessed at the preceding instant TN−1, α and β are predetermined constants, FchargeN−1 is a load factor representative of the time-dependent change of the power of the reactor between the preceding instant TN−1 and the given instant TN, ΔT is the duration of the period separating the instant TN−1 from the instant TN. α and β for example have the values of 3.310−9 s−1, and 4.1710−9 s−1 respectively ΔT may be constant, or on the contrary may be variable. The relevance of these equations is based on the following considerations. The time-dependent change of the number of cobalt 60 atoms in the central emitter 5 of the cobalt neutron detector 1 is governed by the following equation: ⅆ N 60 ⅆ t = N 59 · ( ∑ i ⁢ σ i 59 ⁢ ϕ i ) - λ 60 ⁢ N 60 wherein: N59 and N60 are the number of nuclei of the 59 and 60 isotopes of cobalt in the central emitter 5 of the neutron detector 1; σi59 is the absorption cross-section of cobalt 59 for neutrons of the group of energy i; Φi is the flux of neutrons of energy of the group i around the neutron detector 1; λ60 is the radioactive decay constant of cobalt 60, itself depending on the half-life of cobalt 60. This equation, rewritten by means of the components of the signal delivered by the neutron detector 1, becomes: ⅆ I 60 ⅆ t = α ⁡ ( I tot - I 60 ) - β ⁢ ⁢ I 60 wherein: Itot is the total signal delivered by the neutron detector 1; I60 is the delayed component of the total signal Itot due to the presence of cobalt 60 in the detector; α and β are the constants mentioned above. If an instantaneous equation of change having the same form as the equation above is adopted for the calibration factor C_UCO, the following equation is obtained: ⅆ C_UCO ⅆ t = α ⁡ ( I ) - β ⁢ ⁢ C_UCO wherein α and β are the parameters defined earlier, I is the corrected signal, i.e. the total signal generated by the neutron detector 1 from which the calibration factor C_UCO was inferred. FchargeN−1 is determined at each given instant TN by calculating the average power of the reactor (integral of the power of the reactor) locally around the detector 1 between the preceding instant TN−1 and the given instant TN, and by dividing said average power by the value of the power of the reactor locally around the detector 1 at the preceding instant TN−1. In other words, the load factor FchargeN−1 is calculated according to the following equation: Fcharge N - 1 = ( ∫ TN - 1 TN ⁢ P ⁡ ( t ) ) / ( P N - 1 · Δ ⁢ ⁢ T ) wherein P(t) is the local power of the reactor around the detector, and PN−1 is the local power of the reactor around the detector at instant TN−1. Thus, if the power of the reactor between the instants TN−1 and TN remains constant and equal to the power of the reactor at instant TN−1, the load factor FchargeN−1 will be taken equal to 1. On the contrary, if the power of the reactor changes between the instants TN−1 and TN, this time-dependent change will be taken into account by the load factor FchargeN−1, which will then be different from 1. Initialization of the algorithm is accomplished when the reactor is stopped, by measuring the signal delivered by the cobalt neutron detector 1. This signal is strictly equal to the delayed component I60 due to cobalt 60 in the detector 1, i.e. to the natural decay of cobalt 60 found in the central emitter 5 of the neutron detector 1. C_UCO0 is selected to be equal to the thereby measured value. After starting the reactor, at a plurality of instants TN the following steps are carried out: determination of the period ΔT between TN−1 and TN; determination of the load factor FchargeN−1 representative of the time-dependent change of the power of the reactor locally around the detector around the instant TN−1 and TN; determination of the slope SLOPEN−1, representative of the time-dependent change of the calibration factor between the preceding instant TN−1 and the given instant TN; determination of the calibration factor C_UCON representative of the delayed component I60 of the total signal ItotN due to the decay of cobalt 60 in the detector 1; acquisition of the total signal ItotN delivered by the cobalt neutron detector 1; assessment of the corrected signal IN and storage and memory of this parameter. The period ΔT between two instants TN−1 and TN is typically of the order of 1 month. This algorithm is recurrently repeated at different instants TN, until the next shut-down of the reactor. A new initialization is then carried out during the following shut-down, by measuring the signal delivered by the neutron detector 1 and assignment of the latter to C_UCO0. The algorithm described above has the advantage of being stable. Thus, any difference between the delayed component I60 and the calibration factor C_UCON at instant TN is not amplified in the following iterations. If at an instant TN, the calibration factor C_UCON is greater than the delayed component I60, at the instant following the time-dependent change of the calibration factor, it will be smaller than the time-dependent change of the delayed component I60, so that the difference will decrease. Indeed, if one has:C—UCO>I60 then Itot−C—UCO<Itot−I60. Further, ⅆ I 60 ⅆ t = ⁢ α ⁡ ( I tot - I 60 ) - β ⁢ ⁢ I 60 ⁢ ⁢ and ⁢ ⁢ ⅆ C - UCO ⅆ t = ⁢ α ⁡ ( I tot - C_UCO ) - β ⁢ ⁢ C_UCO α ⁡ ( Itot - I 60 ) > α ⁡ ( Itot - C_UCO ) ⁢ ⁢ and ⁢ - β ⁢ ⁢ I 60 > - β ⁢ ⁢ C_UCO whence ⅆ I 60 ⅆ t > ⅆ CUCO ⅆ t In FIG. 5 the time-dependent change of the difference between the calibration factor C_UCO and the delayed component I60 is illustrated for two methods for computing the calibration factor C_UCO. The difference is in ordinates in FIG. 5, it is expressed in %. It is calculated according to the following formula:Difference=(I60−C—UCO)/(Itot−C—UCO) The abscissa corresponds to the time, expressed as a full power yearly equivalent. The curve in solid lines corresponds to the method for assessing the calibration factor C_UCO discussed above, the calibration factor C_UCO is reinitialized after each full power yearly equivalent, i.e. at graduations 0, 1, 2, etc. . . . . The curve in dashed lines corresponds to a method in which the calibration factor C_UCO is initialized at each shut-down of the reactor, and is then maintained constant. Thus, the calibration factor C_UCO is reinitialized at graduations 0, 1, 2, etc. . . . , i.e. all the full power yearly equivalents. It is seen in FIG. 5 that the method described above is much more accurate, the difference being less than 1%. The curves of FIG. 5 were determined by carrying out a simulation of the operation of the core of the reactor, and of the behavior of the cobalt neutron detector 1. Such a simulation involves unwieldy computations and input data specific to the loading of the core and to the wear of each neutron detector 1. On the contrary, the method of the invention only involves simple computations, which may be easily and rapidly performed by a computer, with very little input data accessible by measurement. In FIG. 6, the time-dependent change of the delayed component I60 is illustrated (dashed lines) and of the calibration factor C_UCO computed by the method described above (solid lines). The time in abscissae is expressed in full power yearly equivalents. It is seen that C_UCO follows I60 with a very small difference. The device provided for applying the method described above is illustrated in FIG. 7. The device for example includes 72 cobalt neutron detectors 1, distributed in the core of the nuclear reactor. These neutron detectors 1 are fixed. The neutron detectors 1 are for example distributed in 12 columns, each column including six detectors. In each column, the neutron detectors 1 are positioned vertically, one above the other. They are regularly spaced out from each other. The 12 columns are distributed in different locations of the core of the reactor. The measurement device also includes a computer 9, symbolized by a frame in dash-dot lines in FIG. 7. The computer 9 includes the device 7 for amplifying and digitizing the total signals from the neutron detectors 1. Each neutron detector 1 is connected to a distinct channel of the device 7, which is specific to it. The device 7 therefore forms means for acquiring at given instants TN, the total signal ItotN provided by each neutron detector 1. The computer 9 moreover includes means 11 for assessing the load factor FchargeN−1, at each given instant TN, by computing the average power of the reactor locally around each cobalt neutron detector 1 between the preceding instant TN−1 and the given instant TN, and by dividing this average power by the value of the power of the reactor locally around said detector 1 at the preceding instant TN−1. The computer 9 recovers the data relating to the power in the operating system of the nuclear power station. Moreover, the computer 9 includes means 13 for calculating the elapsed time period from the preceding instant TN−1 to the given instant TN. The computer 9 also includes memories 15 provided for storing the 72 corrected signals IN−1 computed at the preceding instant TN−1 and the 72 calibration factors C_UCON−1 computed at the preceding instant TN−1. The computer 9 further includes means 17 for assessing the SLOPEN−1 at the given instant TN for each cobalt neutron detector 1 as a function of the corrected signal IN−1 stored in memory and of the calibration factor C_UCON−1 stored in memory. The computation is carried out by using the equation described above. The computer 9 further includes means 19 for computing the calibration factor C_UCON of each of the cobalt neutron detractors 1 at a given instant TN, as a function of the calibration factor C_UCON−1 stored in memory, of the SLOPEN−1 and of the period ΔT separating the given instant TN from the preceding instant TN−1. Further, the computer 9 includes means 21 for computing, for each cobalt neutron detector 1, independently of the others, the corrected signal IN, by difference between the total signal ItotN and the calibration factor C_UCON at instant TN. The corrected signals of each of the neutron detectors 1 may then be exploited by the monitoring and protection systems 23 of the nuclear power station. The method for measuring the neutron flux described above has multiple advantages. Because this method comprises several steps recurrently performed at instants spaced out over a period, the method comprising at each given instant the following steps: acquiring a signal Itot by means of a cobalt neutron detector 1 placed inside the core of the reactor; assessing a calibration factor C_UCO representative of the delayed component I60 of the total signal Itot due to the presence of cobalt 60 in the neutron detector 1; assessing a corrected signal representative of the neutron flux at the neutron detector 1 from the total signal Itot and from the calibration factor C_UCO; assessing a slope SLOPE representing the time-dependent change of the calibration factor C_UCO between the preceding instant and the given instant; the calibration factor C_UCO at the given instant being assessed as a function of the calibration factor C_UCO assessed at the preceding instant, of the slope SLOPE and of the time period separating the given instant from the preceding instant, the method allows very accurate measurement of the neutron flux without any significant deterioration of the ratio between the useful signal and the total signal Itot of the detector 1 in the long run. The delayed component I60 due the presence of cobalt 60 in the central emitter 5 and to its natural decay, is adequately corrected by a calibration factor C_UCO which is re-assessed at each iteration. The calibration factor C_UCO is advantageously re-assessed linearly by taking into account a slope SLOPE. Such a linear re-assessment nevertheless gives the possibility of guaranteeing good accuracy, since the time-dependent change of the delayed component I60 of the signal is linear over limited time intervals, of the order of one month. The slope SLOPE is itself re-assessed periodically, by using a simple equation, the form of which is derived from the time-dependent equation of cobalt 60 within the central emitter 5 of the cobalt neutron detector 1. Such an equation gives the possibility of obtaining excellent accuracy of the measurement method. The equation with which the slope SLOPE may be re-updated takes into account a load factor Fcharge, representative of the time-dependent change of the power of the reactor locally around the detector between the preceding instant and the given instant, which contributes to adapting the algorithm to an operation of the power station while tracking the load for which the power may vary between two instants TN. The method is very simple to use, since the input data are provided to the algorithm by direct read-out of the total signal Itot of the cobalt neutron detectors 1, and by the results of the preceding iterations. It only requires for initialization, measurements of the signal delivered by each of the neutron detectors 1 when the reactor is stopped, and the power is zero. These signals are considered as being the initial calibration factors C_UCO0 to be taken into account. The method only applies rapid and minor computations, so that the algorithm may be directly integrated into a computer calibration tool, used automatically. This reduces the risks of human error. An entirely theoretical computation, modeling the calibration factor C_UCO in a specific way depending on the wear of the detector 1 under a neutron flux, would require detailed knowledge of the values of the neutron flux locally around the detector 1 all along the operating cycle, and specific knowledge of the degree of wear of the detector 1 at the beginning of the cycle. These values are different for each of the detectors 1, are very dependent on the localization of the detector 1 within the core and depend on the history of the flux all along the use of the detector 1. Specific pieces of equipment and individual tracking of the detectors 1 would therefore be required. The method is applicable when the reactor operates with load tracking, or in prolonged operation at reduced power (PORP), because of the taking into account of the parameter Fcharge. Further, the method is stable, because any error in the calibration factor C_UCO, for example resulting from a measurement error, is naturally reduced by the algorithm. The method does not require any specific neutronic calculation for anticipating the time-dependent change of the retarded component I60 due to cobalt 60. It only requires knowledge of the calibration factor C_UCO from the preceding iteration, of the corrected signal I assessed at the preceding iteration, of the time having elapsed between two iterations, and in the case of operation with load tracking, of the local load factor Fcharge at the detector 1. The method uses data specific to each cobalt neutron detector 1 and thus allows individual processing of each detector 1 with a specific correction parameter taking into account the different wear of each of the detectors 1. Thus, the accuracy of the measurements of each cobalt neutron detector 1 is improved, which has advantages in terms of safety (better knowledge of neutron fluxes) and economic advantages (improvement in the operating margins of the reactor, better processing of the wear of the cobalt neutron detectors 1). The device described above may have multiple alternatives. The load factor FchargeN−1 may be assessed by considering not the time-dependent change in the reactor power locally around the detector but the time-dependent change in the overall power of the reactor between the preceding instant TN−1 and the given instant TN. In this case, the load factor FchargeN−1 is the same for all the detectors and the correction is less accurate. It is then assessed from the average power of the reactor between the preceding instant TN−1 and the given instant TN, and by dividing said average power by the value of the power of the reactor at the preceding instant TN−1. The average power of the reactor between TN−1 and TN may be assessed for example by dividing the number of full power equivalent days between TN−1 and TN by the number of days having elapsed between TN−1 and TN. The number of full power equivalent days may be recovered in a computer of the nuclear power station. The measurement device may include any number of cobalt neutron detectors 1, <72 or >72. The period ΔT separating two consecutive instants TN and TN+1 may be constant or may be variable.
	claims	1. A nuclear reactor fuel assembly comprising: a spacer grid including a plurality of intersecting straps having springs and dimples formed into the straps, the intersecting straps defining cells with the springs and dimples arranged to engage fuel rods passing through the cells, wherein at least one of the springs in the spacer grid includes a single base anchored to the spacer grid both in a plane of a corresponding strap and at a mid-plane of the spacer grid and the at least one spring is cantilevered with a bridge region disposed between a distal end of the cantilevered spring and the base of the cantilevered spring, wherein the base of the cantilevered spring is disposed in the plane of the corresponding strap. 2. The fuel assembly of claim 1 wherein the bridge region is parallel with a plane of the strap. 3. The fuel assembly of claim 1 wherein the cantilevered spring includes a contact surface and the bridge region includes a bump spaced apart from the contact surface of the cantilevered spring, the bump being disposed along the cantilevered spring between the contact surface and the base of the cantilevered spring. 4. The fuel assembly of claim 3 wherein the contact surface of the cantilevered spring is less stiff than the bridge region including the bump. 5. The fuel assembly of claim 1 wherein at least some of the cantilevered springs are arranged as dual cantilevered springs disposed on a strap. 6. The fuel assembly of claim 1 wherein an outer strap of the intersecting straps do not include springs. 7. The fuel assembly of claim 6 wherein the outer straps of the plurality of intersecting straps include dimples. 8. The fuel assembly of claim 1 wherein the spacer grid includes outer straps that do not include springs and inner straps that all include springs, wherein only the outermost of the inner straps include the cantilevered springs with bridge regions disposed between distal ends of the cantilevered springs and bases of the cantilevered springs. 9. The fuel assembly of claim 1 further comprising fuel rods comprising fissile material held in a spaced apart arrangement by a plurality of spacer grids. 10. A nuclear reactor comprising a pressure vessel containing a nuclear reactor core wherein the nuclear reactor core includes fuel rods comprising fissile material held in a spaced apart arrangement by a plurality of spacer grids set forth in claim 1. 11. A nuclear reactor fuel assembly comprising: a spacer grid including a plurality of intersecting straps having springs and dimples formed into the straps, the intersecting straps defining cells with the springs and dimples arranged to engage fuel rods passing through the cells wherein at least one of the springs in the spacer grid includes a single base anchored to the spacer grid both in a plane of a corresponding strap and at a mid-plane of the spacer grid and the at least one spring is cantilevered with a first contact surface and a secondary contact surface formed by a bump, the secondary contact surface having at least an order of magnitude higher stiffness than the first contact point and located between the base of the spring and the first contact surface, wherein the base of the cantilevered spring is disposed in the plane of the corresponding strap. 12. The fuel assembly of claim 11 wherein: springs are not formed into the outer straps of the spacer grid, and dimples are formed into the outer straps of the spacer grid. 13. The fuel assembly of claim 11 wherein the springs have a spring constant that is no larger than one-half of the spring constant of the dimples and the secondary contact surface has a spring constant that is at least an order of magnitude higher than the spring constant of the springs. 14. The fuel assembly of claim 11 further comprising fuel rods comprising fissile material held in a spaced apart arrangement by a plurality of spacer grids. 15. A nuclear reactor fuel assembly comprising: a spacer grid including a plurality of intersecting straps having springs and dimples formed into the straps, the intersecting straps defining cells with the springs and dimples arranged to hold fuel rods passing through the cells, wherein at least one of the springs in the spacer grid includes a single base anchored to the spacer grid both in a plane of a corresponding strap and at a mid-plane of the spacer grid and the at least one spring is a cantilevered spring with a contact surface and a bump spaced apart from the contact surface and disposed along the cantilevered spring between the contact surface and the base of the cantilevered spring, the bump limiting travel of the spring, wherein the base of the cantilevered spring is disposed in the plane of the corresponding strap. 16. The fuel assembly of claim 15 wherein: springs are not formed into the outer straps of the spacer grid, cantilevered springs with the bump are formed into the most outboard of the inner straps of the spacer grid, and springs without the bump are formed into the inner straps of the spacer grid other than the most outboard of the inner straps. 17. The fuel assembly of claim 15 wherein the cantilevered spring with travel limiting bump is disposed in the outermost straps that are inside the outer straps. 18. The fuel assembly of claim 15 wherein the travel limiting bump has a spring constant that is at least an order of magnitude higher than the spring constant of the contact surface. 19. The fuel assembly of claim 15 further comprising fuel rods comprising fissile material held in a spaced apart arrangement by a plurality of spacer grids. 20. The fuel assembly of claim 15 wherein the cantilevered spring is made of one of Zircaloy and Inconel.
046876316	claims	1. In a fuel assembly including a top nozzle with an adapter plate having at least one hole and at least one longitudinal structural member with a threaded upper end portion extendible through the adapter plate hole so as to project therefrom, a reusable fasterner device comprising: (a) attaching means threadable onto and from said upper end portion of said structural member between fastened and unfastened positions therewith for attaching and detaching said structural member to and from said top nozzle adapter plate; and (b) retaining means mounted on said top nozzle adapter plate so as to extend about said adapter plate hole and project outwardly from said adapter plate, said retaining means capturing said attaching means when in both said fastened and unfastened positions thereof and maintaining said attaching means with said top nozzle upon removal from said upper end portion of said structural member; (c) said attaching means including (d) said retaining means including (a) an attachment nut including (b) a retainer housing including (a) attaching means threadable onto and from said upper end portion of said structural member between fastened and unfastened positions therewith for attaching and detaching said structural member to and from said top nozzle adapter plate; and (b) retaining means mounted on said top nozzle adapter plate so as to extend about said adapter plate hole and project outwardly from said adapter plate, said retaining means capturing said attaching means when in both said fastened and unfastened positions thereof and maintaining said attaching means with said top nozzle upon removal from said upper end portion of said structural member; (c) said attaching means including (d) said retaining means including 2. The reusable fastener device as recited in claim 1, wherein said retaining means includes a a plurality of lower tabs connected to and extending radially outwardly from said lower end of said tubular body, said tabs resting upon and rigidly connected to said top nozzle adapter plate so as to generally align said central opening of said retaining means with said hole through said adapter plate. 3. In a fuel assembly including a top nozzle with an adapter plate having at least one hole and at least one longitudinal structural member with a threaded upper end portion extendible through the adapter plate hole so as to project therefrom, a reusable fastener device comprising: 4. In a fuel assembly including a top nozzle with an adapter plate having at least one hole and at least one longitudinal structural member with a threaded upper end portion extendible through the adapter plate hole so as to project therefrom, a reusable fastener device comprising: 5. The reusable fastener device as recited in claim 4, wherein said retaining means also includes means attached to one of said tubular body and upper sectors and being movable between open and closed positions for respectively allowing and preventing passage of said lower segments of said attaching means lower flange through said openings defined by said upper sectors of said retaining means. 6. The reusable fastener device as recited in claim 5, wherein said movable means take the form of auxiliary bendable tabs connected to and extending outwardly from said upper sectors, said tabs extending adjacent to said openings between said upper sectors when disposed in said open positions and extending in overlying relation to said openings when disposed in said closed positions.
055457940	claims	1. A method for removing radioactive contaminants from a metal surface comprising: (a) directing a stream of steam to said surface, (b) introducing a solution of nitric acid and cerium in a plus four oxidation state into said stream, (c) directing said stream and said solution to said surface. (a) directing a stream of steam to said surface, (b) heating said surface with said steam, (c) introducing a solution of nitric acid and cerium in a plus four oxidation state into said stream, (d) directing said stream and said solution to said surface. 2. The method of claim 1, wherein said steam is provided between 10 and 70 psi. 3. The method of claim 1, wherein said steam is provided between 110.degree. and 165.degree. C. 4. The method of claim 2, wherein said stream of steam is provided between 110.degree. and 165.degree. C. 5. The method of claim 1, wherein said solution is provided at a rate of at least 0.5 liters per minute. 6. The method of claim 2, wherein said solution is provided at a rate of at least 0.5 liters per minute. 7. The method of claim 3, wherein said solution is provided at a rate of at least 0.5 liters per minute. 8. The method of claim 4, wherein said solution is provided at a rate of at least 0.5 liters per minute. 9. A method for removing radioactive contaminants from a metal surface comprising: 10. The method of claim 6, wherein said steam is provided between 10 and 70 psi. 11. The method of claim 6, wherein said steam is provided between 110.degree. and 165.degree. C. 12. The method of claim 7, wherein said stream of steam is provided between 110.degree. and 165.degree. C. 13. The method of claim 6, wherein said solution is provided at a rate of at least 0.5 liters per minute. 14. The method of claim 7, wherein said solution is provided at a rate of at least 0.5 liters per minute. 15. The method of claim 8, wherein said solution is provided at a rate of at least 0.5 liters per minute. 16. The method of claim 9, wherein said solution is provided at a rate of at least 0.5 liters per minute. 17. The method of claim 6, wherein said surface is heated to a temperature of at least 90.degree. C.
	description	The present invention relates to compositions for the protection of living cells or organisms against electromagnetic radiation and more particularly to polymeric materials including a polyamide, barium sulfate and magnesium sulfate capable of preventing adverse effects associated with exposure to electromagnetic radiation and products formed therefrom. Electromagnetic radiation (EMR) is a self-propagating wave in space with electric and magnetic components. These components oscillate at right angles to each other. EMR is classified into types according to the frequency of the wave: these types include, in order of increasing frequency, radio waves, microwaves, terahertz radiation, infrared radiation, visible light, ultraviolet radiation, X-rays and gamma rays. EMR is emitted by every operating electrical and electronic device. The power of EMR emission varies depending on the size and electrical strength of the device and the electrical current it carries or employs. High voltage power lines are significant emitters of EMR, and field strengths sufficiently high to have the potential for causing adverse EMR effects in humans, animals and plants. Effects can be detected hundreds of feet away. Smaller devices such as computers, television sets, microwave ovens and the like emit lesser quantities of EMR, but the effect on humans can still be significant because people are in much closer proximity to such devices. Electromagnetic radiation carries energy and momentum, which may be imparted when it interacts with matter. Thus once struck, the matter can be affected. While the effect may vary depending on frequency and amplitude, there are biological effects that can be modulated by exposure to electromagnetic radiation. Among the effects believed to be associated with exposure to electromagnetic radiation include the disruption of hydrogen bonding. Thus, exposure to such radiation may disrupt the natural hydrogen bonding of compounds or molecules. This also affects the hydrogen bonding of water molecules. Therefore there is a particular concern regarding the exposure of areas of the body that are highly fluid, such as within the brain or blood stream. Thus while the effects of electromagnetic radiation are not widely accepted by all, it is nonetheless the desire of many prudent people to protect themselves, their animals and plants against whatever health risks might be involved by their exposure to electromagnetic radiation. Adverse human health effects that have been reported as attributable to long-term electromagnetic radiation exposure include but are not limited to occurrence of certain cancers, multiple sclerosis and autism. Adverse effects on animals have including stillbirths of young and reduction of milk production in cattle have also been reported. Unfortunately, effective and convenient devices for shielding against EMR are not generally available. Essentially the only defense against EMR has been removal of persons, animals and plants from proximity to the EMR-emitting devices. For major emitters such as power lines or electrical substations, this has usually meant that one has had to move to a different house or to a different job location away from the power line or substation, which commonly means substantial expense and inconvenience. The adverse costs and inconveniences are similar to farmers and ranchers who must move animals and crops to locations remote from the power lines or stations. For devices such as microwave ovens or computers, it has meant that a person must sit or stand at an awkward distance from the device, which can impair the person's ability to use the device in an optimum manner. The inventor of the present invention has proposed compositions for such protection in the past. U.S. Pat. No. 6,369,399 teaches compositions for the protection against electromagnetic radiation. The compositions include a material including an oxydated hydrocarbon emulsifier; a galvanic salt; an alkaloid; a dye or stain; and a polysaccharide. Although the compositions did demonstrate protection against electromagnetic radiation, the preparation of such a material was complex and therefore its widespread adoption was hindered. Thus there remains a need to develop compositions capable of protecting against electromagnetic radiation that are less complex and easier to adapt to a variety of uses. The present invention addresses the need to provide compositions for the protection against electromagnetic radiation and provides related benefits. Thus it is the primary object of the present invention to provide compositions that protect against adverse effects associated with exposure to electromagnetic radiation. It is another object of the present invention to provide housings or portions thereof for electronic devices that emit electromagnetic frequencies to reduce adverse effects associated with the use of such electronic devices. It is yet another object of the present invention to provide fabrics and protective garments capable of protecting against exposure to electromagnetic radiation. In one aspect of the present invention a polymeric material is provided to reduce adverse effects of electromagnetic radiation exposure. The polymeric material includes a polyamide such as nylon 6 or nylon 6, 6, barium sulfate and magnesium sulfate. The polymeric material upon exposure to incident electromagnetic radiation emits subtle electromagnetic oscillations at probiotic frequencies that counter or reduce adverse effects of incident electromagnetic radiation. The polymeric material may be used for the protection of humans, animals, plants, eukaryotic cells or organisms and the like. The polymeric material may be formed into a protective housing for electronic devices and may be formed into protective fabrics. In another aspect of the present invention, a housing for an electronic device that emits electromagnetic radiation is provided. The housing includes a polymeric material including a polyamide, barium sulfate, and magnesium sulfate in an amount suitable to reduce exposure to such radiation. The housing, upon exposure to incident electromagnetic radiation, emits subtle electromagnetic oscillations at probiotic frequencies that protect the user against incident electromagnetic radiation. In some embodiments the polyamide is nylon such as nylon 6 or nylon 6, 6. Examples of electronic devices that may benefit from the housing include wireless telephones, cordless telephones, audio players such as MP3 players and others, wireless headsets, headphones, computers, televisions and the like. In another aspect of the present invention a fabric or a protective garment constructed from fabric is provided to protect against exposure to electromagnetic radiation. The fabric includes a polymeric material including a polyamide, barium sulfate and magnesium sulfate. Fabrics of the present invention are believed to emit subtle electromagnetic oscillations at probiotic frequencies when exposed to incident electromagnetic radiation. As an introduction, the present invention provides polymeric materials and compositions formed therefrom to protect against exposure to electromagnetic radiation, such as electromagnetic frequencies emitted from electronic devices, power lines and the like. Compositions provided herein have been tested for their ability to protect against electromagnetic radiation by measuring a variety of biological indicators associated with brain chemistry in the interstitial fluid and have demonstrated the ability to reduce or counteract abnormal shifts identified upon exposure to electromagnetic radiation. In view of the present disclosure, one skilled in the art to which the present invention belongs will be able to provide a variety of protective housings, structures, fabrics and the like that provide as an element, a polymeric material including a polyamide in combination with barium sulfate and magnesium sulfate. Thus the polymeric compositions of the present invention can be adapted for use as a protective barrier against exposure to electromagnetic radiation; as protective housings or portions thereof for electronic devices that emit frequencies that adversely effect biological systems; and can be woven into fabrics for production of protective garments. The following description provides various preferred embodiments and uses for the polymeric material described herein. The various embodiments are intended to be nonlimiting since the polymeric materials may be varied or adapted for many protective uses. In one aspect of the present invention a polymeric material is provided to reduce adverse effects of electromagnetic radiation exposure. The polymeric material includes a polyamide such as nylon 6 or nylon 6, 6, barium sulfate and magnesium sulfate. Though nonlimiting, the polymeric material of the present invention is believed to emit subtle electromagnetic oscillations at probiotic frequencies that counter adverse effects of incident electromagnetic radiation. The polymeric material has been demonstrated as useful in reducing exposure to electromagnetic radiation and can be thus be provide to protect humans, animals, eukaryotic cells, plants and the like. Referring to the new combination of compounds and the beneficial results described herein, the present invention utilizes a polymeric material including a polyamide, barium sulfate and magnesium sulfate. The polymeric material is capable of reducing the effects of exposure from electronic devices and can be incorporated into a variety of housings, fabrics and protective structures. Though nonlimiting, the preferred polyamide is nylon-6 or nylon-6, 6. In the preferred embodiment the ratio by weight of the polymeric material is about ten parts by weight polyamide, about two parts by weight barium sulfate, and about one part by weight magnesium sulfate. However other embodiments include variations on these ratios. In some embodiments, the amount of one or more of the compounds varies by about 10%. In another embodiment, the amount of one or more of the compounds varies by about 15%. In still other embodiments, the amount of one or more of the compounds varies by about 20%. Thus the ratios provided herein correspond to preferred embodiments found during development but are not intended to limit the scope of the present invention. One may determine the particular desired ratio by varying ratios of each compound, forming a protective structure such as a housing or fabric and testing the ability to protect against electromagnetic radiation. In some embodiments, electro interstitial scan (EIS) analysis can be used to test for protective properties. In alternative embodiments a “phantom head” or “phantom body” study may be used to assess affects against electromagnetic radiation. The polyamide provides the primary polymer backbone to which the barium sulfate and magnesium sulfate interact or bind to form the polymeric material of the present invention. Polyamides are monomers of amides linked by peptide bonds. Although some polyamides occur naturally, such as those found in wool and silk, others are formed artificially. Polyamide polymers are frequently produced by condensation reactions between an amino group on one polymer and a carboxylic acid or acid chloride group on the opposing polymer. These reactions typically eliminate water, ammonia or hydrogen chloride thereby resulting in a polyamide chain. In the preferred embodiment of the present invention, the polyamide provided in the polymeric material is a nylon. Nylons are some of the most common polymers used as synthetic fibers and thus compositions of the present invention may be provided as substitutions for nylons for the preparation of fibers if the protective features of the present invention are desired. Nylons are commonly used in the clothing industry and the plastics industry. Most preferably, the polyamide of the present invention is nylon 6, 6 or nylon 6. Nylon-6, 6, which is also referred to those skilled in the present art as polyamide 6-6 or PA66, is a semicrystalline polyamide commonly used in fiber applications such as carpeting, clothing and tire cord. It is also used as an engineering material in bearings and gears due to its good abrasion resistance and self-lubricating properties. Nylon-6, 6 includes repeating units of the formula C12H22O2N, has a molecular weight of about 226.32 g/mol and can be formed by condensation reactions of a diamine and a dicarboxylic acid or acid chloride, such as hexamethylene diamine and adipoyl chloride, so that peptide bonds form at both ends of the monomers. The numerical indications within nylons indicate the number of carbons donated by the monomers; the diamine first and the diacid second. Thus nylon-6, 6, refers to the donation of 6 carbons from the diamine and 6 carbons from the diacid to form the polymer chain and is a repeating unit of alternating monomers, one after another. The polymer reaction is typically performed in an aqueous solvent. Nylon-6, also referred to as polyamide 6 or PA6, is a semicrystalline polyamide used most commonly in tire cord. Nylon-6 has a lower melting temperature compared to nylon 6, 6 and in general is believed to have better affinity towards dyes, tends to be more elastic and tends to be more resistant to weathering. Thus in some instances one may prefer to use nylon-6 depending on the resulting material, housing, structure, fabric and the like. The determination of which to use is well within the ability of one skilled in the present art. Referring back to the compound, nylon-6 is repeating unit of C6H11ON with a molecular weight per unit of 113.16 g/mol. Nylon-6 is not a condensation polymer but instead is formed by a ring-opening polymerization reaction of the monomer caprolactam. Like nylon 6, 6, the technique for preparing nylon 6 is well known in the art. Nylon-6 was developed by DuPont and may be obtained from a variety of sources such as Sigma-Alderich (St. Louis, Mo.). Barium sulfate is often provided as a fine white powder and has the chemical formula BaSO4. Generally it is poorly soluble in water and other traditional solvents but is soluble in concentrated sulfuric acid. Barium sulfate is commercially available through a variety of vendors including Sigma-Aldrich (St. Louis, Mo.). The preferred ratio of barium sulfate to polyamide is 20 grams barium sulfate to 100 grams nylon 6 or nylon 6, 6. The preferred embodiment is nonlimiting and thus more or less barium sulfate may also be used as long as protective properties are maintained. In one embodiment the ratio of barium sulfate to polyamide is about 20-25 grams of barium sulfate per 100 grams of polyamide. In another embodiment the ratio of barium sulfate to polyamide is about 25-30 grams of barium sulfate per 100 grams of polyamide. In another embodiment the ratio of barium sulfate to polyamide is about 15-20 grams per 100 grams of polyamide. In another embodiment the ratio of barium sulfate to polyamide is about 10-15 grams per 100 grams of polyamide. Thus the ratios are intended to provide various useful ranges, which may be considered by one skilled in the art for the particular use, and are intended to be nonlimiting. Magnesium sulfate is often provided as transparent crystals or a white powder and has the chemical formula MgSO4. It can also be found as a heptahydrate, MgSO4.7H2O. Magnesium sulfate is available through a variety of vendors including Sigma-Alderich (St. Louis, Mo.). The preferred ratio of magnesium sulfate to polyamide is 10 grams per 100 grams polyamide. In another embodiment the ratio of magnesium sulfate to polyamide is 10-15 grams per 100 grams polyamide. In another embodiment the ratio of magnesium sulfate to polyamide is 15-20 grams per 100 grams polyamide. In another embodiment the ratio of magnesium sulfate to polyamide is 7-10 grams per 100 grams polyamide. In another embodiment the ratio of magnesium sulfate to polyamide is 3-7 grams per 100 grams polyamide. Thus the ratios provided herein are useful as guidance for the formation of protective materials, housings, structures and fabrics but are intended as nonlimiting with respect to scope of the present invention. Compositions according to the present invention are formed by preparing the protective polymeric material then casting, molding or manipulating the material to form the desired product. In general, the polyamide is formed into a polymer chain then the barium sulfate and magnesium sulfate are added to the chain. The polyamide polymer may be purchased as single monomers or polymers and may be polymerized using chemistries that correspond to the particular polyamide or desired polymer. In one example, a condensation reaction is used to form a polyamide including nylon-6, 6. In another example ring opening polymerization is performed using caprolactam to form a nylon-6 polymer. After forming a polymer backbone, conventional chemistries can be used to form ester linkages or covalent bonds between the polymer backbone and the barium sulfate or magnesium sulfate. Once combined and allowed to react, a polymer incorporating the polyamide, barium sulfate and magnesium sulfate is formed. The resulting polymeric material is viscous slurry, which can be further processed to form desired protective housings, structures, fabrics and the like. As a nonlimiting exemplary embodiment, formation of the polymeric material may include mixing magnesium sulfate, barium sulfate and the polyamide at ratios provided herein and adding the mixture to a compounding machine. The operation of compounding machines for the preparation of polymeric materials is well known to those skilled in the present art and is intended to be nonlimiting. The mixture is heated to melt the polyamide and to absorb or combine with the magnesium sulfate and barium sulfate. Temperatures may vary depending on the melting temperature of the polyamide and may be about 250 degrees C. The mixture can then be forced through holes for the production of thread-like materials which can be cooled and cut into desired sized threads, pieces, granules and the like. Once cut the product may be collected for desired applications. The polymeric material may be further processed or formed as desired. In addition to the magnesium sulfate and barium sulfate, the polymeric material may also include compounds that affect the characteristics of the resulting composition according to the desires of the user. In some embodiments, one or more dyes are added to enhance or alter the coloring of the composition. In other embodiments, fillers are added to increase or decrease the density of the resulting polymeric matrix. In still other embodiments, compositions are coated with a coating to enhance sheen or reflective properties. As will become apparent to one skilled in the art to which the present invention belongs, the polymeric materials of the present invention may be cast or molded to form a variety of shapes and therefore a variety of protective housings. Thus it is another aspect of the present invention to provide a housing for an electronic device that is capable of protecting a user against electromagnetic radiation emitted from the electronic device. The housing includes a polyamide, such as nylon 6 or nylon 6, 6; barium sulfate; and magnesium sulfate. It is believe that the housing upon exposure to incident electromagnetic radiation emits subtle electromagnetic oscillations at probiotic frequencies that protect the user against the incident electromagnetic radiation. Casting and molding techniques are well are known in the plastic and polymer arts and are incorporated herein. Thus although the present invention provides increased protection against electromagnetic radiation, features such as viscosity and molding characteristics remain largely unchanged allowing conventional casting and molding techniques to be utilized. As with many nylons, the polymeric material of the present invention may also be provided as fibers or in a fibrous configuration for the preparation or weaving of protective fabrics. Thus one skilled in the art would readily acknowledge the present invention is not limited by a composition's size or configuration as the polymeric material may be formed in any suitable size or shape using known casting or molding techniques. The polymeric material of the present invention has particular utility as a protective housing for electronic devices. Many electronic devices emit electronic radiation. Thus in some preferred embodiments of the present invention the polymeric material is formed into a rigid housing to house an electronic device. It is believed that by encasing the electronics in a housing according to the present invention, the effect of frequencies generated by such devices on humans, plants and the like will be minimized. It is believed that the frequencies emitted from the electronic device will act as a carrier allowing the delivery of the subtle low frequencies emitted from the polymeric material to occur in combination with the harmful frequencies generated from the electrical device. Thus by delivering the protective frequency in combination with the harmful frequency, the overall effect from the electrical device is reduced or minimized. In some embodiments the polymeric material does not make up the entire housing but instead only a portion of the housing. In these embodiments, the polymeric material may be used in the front, back, top, bottom, side or any portion thereof. The examples demonstrate beneficial features of the present invention as a protective housing. More specifically, the examples describe experiments conducted where harmful effects were reduced or minimized by adapting an electronic device such as a cellular telephone with a polymeric material according to the present invention. The examples also demonstrate the ordinary use of traditionally housed cellular telephones effect the chemistry in the brain. Particular abnormal activity was found in the frontal and temporal lobes. In addition, abnormal shifts in minerals and hormones were also observed. However, when the housing was adapted with a polymeric material according to the present invention, activity in the frontal and temporal lobe was deemed normal or more normal than without. Also, the abnormal shifts identified in minerals and hormones were not observed when using a housing according to the present invention. Thus, the studies demonstrate through EIS analysis, that a polymeric material including a polyamide, barium sulfate and magenesium sulfate is effective at protecting humans against electromagnetic radiation. Housings of the present invention are not limited to cellular telephones but instead are intended for use with a variety of electronic devices that emit EMR, The polymeric material of the present invention may be formed into a housing or portion thereof for a number of household appliances including refrigerators, microwaves, blenders, coffeemakers, food processors and the like. Moreover the housings may be used for entertainment devices such as televisions, stereos, portable audio players such as MP3 players, and computers. Housings of the present invention may also be used for electronic devices such as telephones, cordless telephones, headphones, wireless headphones and the like. Thus any electrical device that emits a frequency similar to any of the devices provided herein may be adapted with a housing according to the present invention. In some embodiments, the housing of the present invention is provided to protect against frequencies in the MHz range. In other embodiments, housings according to the present invention are provided to protect against frequencies in the GHz range. If testing is desired, electro interstitial scanning is one method that may be used to detect changes in biological state after exposure to the electronic device with and without the protective housing. Another method is to test the protective capabilities using a “phantom head” or “phantom body” that mimics the conductivity or dielectric constant of the exposed region. The polymeric material of the present invention also provides a particular utility as a fabric in preparation of protective garments and the like. If exposed to electromagnetic radiation, the protective garment may help reduce or minimize adverse effects associated with exposure. Examples of particular garments are any known in the art and may include hats, jackets, shirts or blouses, pants, gloves, boots or shoes and the like. The garments may have particular utility in industries where electronic device manufacturing or testing occurs. It is therefore another aspect of the present invention to provide a fabric for the protection of a user against exposure to electromagnetic radiation. The fabric includes a polyamide, such as nylon-6 or nylon-6, 6; barium sulfate; and magnesium sulfate. The fabrics provided herein, upon exposure to incident electromagnetic radiation are believed to emit subtle electromagnetic oscillations at probiotic frequencies that protect the user against the incident electromagnetic radiation. Since the present invention retains many of the characteristics as conventional nylons, the methods used to form fibers and fabrics from nylons may also be used with the present invention. In particular the methods of forming fibers and fabrics from nylon-6, 6 and nylon-6 can be used with the present invention. As general guidance, once the polymer material including the polyamide, barium sulfate and magnesium sulfate is formed, the material may be extruded into fibers through pores, such as those provide in an industrial spinneret. During extrusion the individual polymer chains tend to align because of viscous flow. If subjected to cold drawing afterwards, the fibers align further, increasing their crystallinity, and the material acquires additional tensile strength. In practice, fibers incorporating the polymeric material of the present invention for fabrics are most likely to be drawn using heated rolls at high speeds. The resulting fibers may then be woven into fabric and thus used the preparation of garments having protective features. The preferred embodiments have described a variety of compositions useful for the protection against electromagnetic radiation. Though nonlimiting, the polymeric material of the present invention is believed to oscillate upon incident radiation. The oscillation is believed to generate a subtle, low frequency, non-coherent electromagnetic field (random field) that can affect the hydrogen lattice of the molecular structure of water and thus modify the electrodynamic properties of water. The low frequency oscillation is of a frequency lower than the incident radiation. It is believed these low frequency oscillations emitted from the polymeric material can be carried by higher frequencies generated by electronic devices, without adverse interaction and thus can be delivered in combination with the harmful frequency for desired protection. The biological effect of exposure to electromagnetic radiation is not fully understood however it is believed the electromagnetic radiation affects the water molecules and hydrogen bonds within the body. It is believed the oscillations generated by the compositions of the present invention protect against such effect by causing the reorganization of the water clathrate structures. This reorganization is believed to be beneficial and help prevent adverse reactions from exposure to the higher frequencies emitted from electronic devices. It will be evident to one skilled in the art that there are numerous embodiments of the present invention that are not expressly described herein, but which are clearly within the scope and spirit of the invention. The description is provided to demonstrate a variety of preferred embodiments only. EIS analysis was conducted at an independent testing facility to assess the biological effects of electromagnetic radiation from a cellular telephone on the human body and whether effects would differ if using a cellular telephone housed in a polymeric material of the present invention (herein referred to as MRET). In summary it was found that exposure to the cellular telephone without MRET caused significant shifts in brain chemistry within the right frontal and temporal lobe; whereas exposure to the cellular telephone with MRET did not show the adverse shifts. A representative example is provided as FIGS. 1A-1C. In addition levels of insulin, ACTH and TSH were also believed to be adversely effected after exposure to the cellular telephone without MRET. The adverse shifts in brain chemistry due to the exposure to the cellular telephone worsened over time. Although initial effects were difficult to detect, after 20 minutes from halting exposure, the biological effects continued to deteriorate, which suggest the effects from cellular phone usage continue beyond the initial exposure. The majority of the deleterious effects were lessened or mediated after use with the cellular telephone using MRET suggesting MRET plays an important role in preventing or correcting adverse effects from exposure to electromagnetic radiation. Materials and Methods The experiments detected changes in brain chemistry using Eletro Interstitial Scanning (EIS). Subjects were scanned at four time points. The first scan was conducted before any exposure. The second scan was conducted after 5 minutes of exposure to a TREO 650 cellular telephone (referred to herein as TREO RF) which operates at frequencies of about 1851.25-1908.75 MHz (PCS and CDMA frequencies). The TREO RF was placed next to the individual's right ear. The subject was then scanned after waiting 20 minutes. A TREO 650 cellular telephone adapted with a housing including the polymeric material of the present invention (referred to as TREO-MRET) was then used. The last scan occurred after 20 minutes of exposure to TREO 650 MRET. EIS gives a comprehensive overview of the reactions of the body. 3D models of the full body and various different parts of the body are created based on the electro interstitial gram (EIG). The models are color coded to indicate where areas of imbalance are hyper-functioning or hypo-functioning. In essence, EIS provides a functional assessment of the main organs, with report screens that show interstitial biochemical values and an evaluation of body composition including lean mass, fat mass and hydration data. Measurements are further extrapolated to provide report screens with hormone, electrolyte, neurotransmitter and oxidative stress analyses. More specifically, the EIS system operates as a biosensor, which analyzes the interstitial fluid locally in vivo by application of a D.C. current between cutaneous zones using electrodes. In use, the EIS introduces electric signals of low intensity (1.28V D.C.) through the human body via 6 electrodes. This is painless and has no negative effects to the patient. About 22 measurements are taken. The scanning results are recorded by EIS software, which analyzes and interprets the test results and produces a variety of informative models, graphs and data for interpretation by a medical practitioner. Results from Subject 1: 42 yr Old Female Initial EIS showed reduced conductivity (hypo-activity) in the right and left frontal lobes, intra-cranial vessels and right temporal lobe before exposure. This was believed to be stress-related. Thus the effect of 5 minute exposure to the TREO RF was initially not conclusive. Further analysis showed endogenous chatecholamines sharply decreased after exposure to TREO RF, which corresponds to low adrenal medullary hormone and thus TREO RF appears to adversely affect neurotransmitter activity. Dopamine levels after TREO RF also dropped. Despite abnormal values for frontal lobes, temporal lobes, intra-cranial vessels and amygdalas from measurements taken after 20 minutes from the earlier scan, positive effects after TREO-MRET exposure included: decreased cranial blood pressure; decreased cranial blood viscosity, decreased carbon dioxide levels, and decreased intra-cranial blood pressure. The values, which were statistically below the norm for the general population, were deemed positive in proportion to the subject's low values overall. An increase in phosphorous and a decrease in calcium was detected suggesting mineral balance may be slightly affected by the TREO-MRET however no shift in hormone levels was identified. Results from Subject 2: 48 yr Old Female The initial scan showed reduced oxygen levels and increased carbon dioxide levels believed to be associated with a fast paced lifestyle. Immediately after exposure to TREO RF, reductions in elevated values for the frontal lobe were identified. Blood pressure, H2O content, and ATP levels were adversely increased in the right temporal lobe. These adverse effects are believed to be associated with exposure to TREO RF. After waiting 20 minutes and before exposure to TREO-MRET, EIS showed significant abnormal values in the right temporal lobe, left temporal lobe, hypothalamus and left amygdala. Abnormal values in insulin, ACTH and TSH were also identified. In addition, measurements of the vertebral column suggesting nerve supply worsened. It is believed the negative effects associated with TREO RF continued over time. After exposure to the TREO-MRET, positive effects were identified for blood pressure, blood viscosity, ATP values and mitochondrial activity, oxygen levels and carbon dioxide levels. In addition, positive effects were detected in values of insulin, ACTH, cortisol, thyroid hormone and TSH. Improvement in the vertebrae was also identified. Results from Subject 3: 42 yr Old Male The initial scan showed abnormal levels in the right frontal lobe prior to testing. In addition, elevated intra-cranial blood pressure and hyperactivity of the temporal lobes was also shown. Because of the initial heightened values it was difficult to assess whether some of the changes in brain chemistry immediately after 5 minute exposure to the TREO RF occurred. After waiting 20 minutes and before exposure to TREO-MRET, EIS showed abnormal values in potassium, ACTH, insulin and cortisol. Abnormally high values were observed for dopamine. Abnormally low values were observed for catecholamine and serotonin. After exposure to TREO-MRET, insulin levels were improved but still below normal. Levels of cellular potassium, ACTH, catecholamine, dopamine and serotinin were normal after exposure to TREO-MRET. The present example demonstrates the ability of the polymeric material of the present invention to reduce the effects of electromagnetic radiation on a “phantom head,” which mimics the human head muscle and brain tissue composition. The intensity and localization of electromagnetic intensity was measured. The results showed a significant decrease in electromagnetic radiation intensity but no significant shift in localization indicating the polymeric material successfully reduces potential harmful effects on brain chemistry due to electromagnetic exposure. The study was performed using a variety of wireless mobile phones and is described in more detail below. To assess the protective effects of using the polymeric material of the present invention against exposure to radiation, a “phantom head” was formed to mimic the brain and muscle composition within the head. The “phantom head” was produced using a combination of hydroxyethylcellulose (FEC) gelling agent and saline solution. The mixture was calibrated to obtain proper dielectric constant (permittivity) and conductivity of the simulated tissue. The dielectric constant at about 835 MHz was about 40 and at about 1900 MHz, was about 39. The conductivity at about 835 MHz was about 0.88 mho/m, and the conductivity at about 1900 MHz was about 1.43 mho/m. An APREL Laboratories ALSAS system with a dosimetric E-field probe E-020 was used for measurements. The dipole was oriented parallel to the body axis. The investigation was conducted on cellular phones including Sanyo Model PM-8200(S), Kyocera Wireless Model 2325 and LG Model VX6000. Wireless mobile phones were evaluated in this experiment for localized specific absorption rate (SAR) for controlled environment/occupational exposure limits specified in ANSI/EEE Std. C95.1-1992 and had been tested in accordance with the measurement procedures specified in IEEE 1528-2003 and OET Bulletin 65. The RF phone was placed into simulated transmit mode using the manufacturer's test codes. Such test signals offer a consistent means for SAR and are recommended for evaluating of SAR data. Each SAR measurement was taken with a fully charged battery. In order to verify that each phone was tested at full power, conducted output power measurements were performed before and after each SAR test to confirm the output power. SAR measurement results were obtained, analyzed and compared to provide the scientific conclusion of the experiment. The protective polymers were prepared with and without the polymeric material according to the present invention. In the experimental polymer, about 1 gram of polymeric material was used (referred to as MRET polymeric material), whereas the control contained no MRET polymeric material. The resulting polymers were placed in an exposed jack then positioned next to the phantom head for measurement. Control and the experimental conditions were compared to determine differences in electromagnetic radiation intensity and localization of signal. The results were displayed as a heat map, which demonstrates the positioning and intensity of signal as hot spots and cool spots. Referring to FIGS. 2A and 2B, the analysis of “Hot Spot” Area Scan data provides evidence that the incorporation of 1 gram of MRET polymeric material in the protective polymer for the RF phones affects the amplitude of emission but does not change location of the “Hot Spot”. More specifically, the incorporation of 1 gram of MRET polymeric material protected the “phantom head” against the intensity of the electromagnetic radiation, while showing that the signal remained in substantially the same location as without the MRET polymeric material. Thus intensity was largely affected, whereas localization was not. The intensity of electromagnetic radiation when incorporating the MRET polymeric material decreased the amplitude in 80% of the data points. 60% of the data points were observed to have a significant decrease in SAR values in the range of 10% to 50%. Thus the incorporation of the MRET polymeric material in the “phantom head” leads to the reduction of the majority of SAR values. 12 SAR values out of 16 meaningful SAR values in this experiment were reduced in the range of 16.5%-32.6%, and only 3 SAR values increased by 1.0%-5.6%.
	summary
	abstract	An apparatus (1) for treatment with radiation for personal care, comprising a housing (2) provided with a UV source (3) and a wall (4) made of UV-transparent material which covers the housing (2) and which has a central area (7) and side areas (8) adjoining the central area (7). The wall (4) has a lower transmission to radiation with a wavelength less than 320 nm near the central area (7) than near the side areas (8). Thus the effective radiation energy of radiation with a wavelength less than 320 nm reaching the side portions of the irradiation plane (9) approximates a value of 0.14 W/m2 (European standard EN 60335-2-27), while the effective radiation energy of radiation with a wavelength less than 320 nm reaching the central portion of the irradiation plane (9) also approximates this value. This results in a practically uniform distribution of effective radiation energy of radiation with a wavelength less than 320 nm over the radiation surface (9) as a whole. A tanning result which is as uniform as possible is thus achieved in a time period which is as short as possible.
047013915	claims	1. A mask for forming patterns in lacquer layers by x-ray lithography comprising: a multilayer diaphragm being highly transparent to X-ray radiation, said multilayer diaphragm including a layer of magnesium and at least one additional layer, and a patterned layer on one major surface of said multilayer diaphragm for absorbing said X-ray radiation, said patterned layer contacting said layer of magnesium, wherein said at least one additional layer includes a layer of gold and a layer of titanium oxide (TiO.sub.2), said layer of titanium oxide (TiO.sub.2) contacting said layer of magnesium at a side opposite to said patterned layer.
040452824	abstract	Thermal monitoring of a reactor core is carried out by utilizng coolant temperatures detected at the outlets of a plurality of fuel assemblies. The mean value of the "hot" and "cold" core outlet temperatures and the difference between these latter are established at each instant. An analog signal corresponding to the temperature difference between the coolant temperature at the outlet of said assembly and the mean core outlet temperature is delivered in respect of each fuel assembly. The signal is processed in order to initiate the appropriate safety actions.
	claims	1. A radiation detector, comprising:a substrate;a plurality of control lines provided on the substrate and extending in a first direction;a plurality of data lines provided on the substrate and extending in a second direction crossing the first direction;a plurality of detection parts arranged in a matrix, each of the plurality of detection parts including a thin film transistor and a conversion part, the thin film transistor electrically connected to one of the plurality of control lines and one of the plurality of data lines, the conversion part electrically connected to the thin film transistor, the conversion part configured to convert radiation or light into electricity;a control circuit configured to, for each detection part of the plurality of detection parts, switch an on state and an off state of the thin film transistor included in the detection part;a signal detection circuit configured to, for each detection part of the plurality of detection parts, read out image data in the on state of the thin film transistor included in the detection part; andan incident radiation detection part configured to detect a start of radiation incidence, based on a value of the image data read out in the on state of the thin film transistor included in one of the plurality of detection parts,wherein the control circuit is further configured to, during a process of scanning each of the plurality of control lines extending in the first direction:select a first control line of the plurality of control lines by switching the thin film transistors electrically connected to the first control line to the on state,select a second control line of the plurality of control lines by switching the thin film transistors electrically connected to the second control line to the on state, andselect a third control line of the plurality of control lines by switching the thin film transistors electrically connected to the third control line are switched to the on state,wherein the signal detection circuit is further configured to, during the process of scanning each of the plurality of control lines extending in the first direction:read out the image data by reading out signals of each of the plurality of data lines electrically connected to the thin film transistors switched in the on state, andread out correction data by reading out signals of each of the plurality of data lines between selecting the first control line and selecting the second control line and between selecting the second control line and selecting the third control line, wherein each of the thin film transistors of the plurality of detection parts is in the off state during the read out of the correction data. 2. The radiation detector according to claim 1, further comprising an image processing part configured to process a radiation image based on the image data, wherein the image processing part is configured to correct the image data using the correction data. 3. The radiation detector according to claim 2, wherein the signal detection circuit is further configured to, during the process of scanning the plurality of control lines extending in the first direction, provide image indexes identifying the image data and the correction data read out at a time of reading out the image data, after reading out the image data, or both before reading out the image data and after reading out the image data. 4. The radiation detector according to claim 3, wherein the image processing part is further configured to extract the image data and the correction data based on the image indexes, process the correction data, and correct the image data by using the processed correction data. 5. The radiation detector according to claim 1, wherein the signal detection circuit is further configured to, during the process of scanning the plurality of control lines extending in the first direction, convert a difference between the read out image data and the read out correction data into a digital signal and output the digital signal. 6. The radiation detector according to claim 1, whereinthe control circuit is further configured to, in response to the incident radiation detection part detecting the start of radiation incidence, scan the plurality of control lines repeatedly. 7. The radiation detector of claim 1, wherein the signal detection circuit is further configured to read out both of the image data and the correction data every time a control signal for scanning one of the plurality of control lines extending in the first direction is output from the control circuit. 8. The radiation detector of claim 1, wherein the signal detection circuit is configured to read out the correction data immediately before and immediately after reading out the image data, during the process of scanning the plurality of control lines extending in the first direction.
058964312	abstract	Systems, methods, and apparatus for preventing steam leakage between a drywell and a wetwell in a nuclear reactor are described. In accordance with one embodiment of the present invention, the vacuum breaker of the nuclear reactor is coupled to a vacuum breaker condensing system which includes a condenser and a steam inlet pipe. The steam inlet pipe is substantially hollow and includes a first end, a second end, and a loop seal between the first and second ends. The first end of the pipe is positioned adjacent the drywell and the second end of the pipe is coupled to the vacuum breaker. The condenser is positioned proximate the steam inlet pipe and includes an inlet, an outlet, and a plurality of condenser tubes. The condenser inlet and condenser outlet are each coupled to a pool of water, e.g., the Gravity Driven Cooling System pool, and configured to draw water from the pool of water and through the condenser tubes to substantially condense steam flowing through the steam inlet pipe.
	description	This invention relates generally to a method for refueling a nuclear reactor, and in particular, to a method of refueling a small modular reactor with a compact containment vessel. Nuclear power plants which employ light water reactors require periodic outages for refueling of the reactor. New fuel assemblies are delivered to the plant and are temporarily stored in a fuel storage building, along with used fuel assemblies which may have been previously removed from the reactor. During a refueling outage, a portion of the fuel assemblies in the reactor are removed from the reactor to the fuel storage building. A second portion of the fuel assemblies are moved from one support location in the reactor to another core support location in the reactor. New fuel assemblies are moved from the fuel storage building into the reactor to replace those fuel assemblies which were removed. These movements are done in accordance with a detailed sequence plan so that each fuel assembly is placed in a specific location in accordance with an overall refueling plan prepared by the reactor core designer. Refueling activities are often on the critical path for returning the nuclear plant to power operation, therefore the speed of these operations is an important economic consideration for the power plant owner. Furthermore, the plant equipment and fuel assemblies are expensive and care must be taken not to cause damage or unnecessary radiation exposure due to improper handling of the fuel assemblies or fuel transfer equipment. The precision of these operations is also important since the safe and economical operation of the reactor core depends upon each fuel assembly being in its proper location. The typical pressurized water reactor needs to be refueled every eighteen to twenty-four months. During refueling, the reactor is disassembled and the core is off-loaded into the storage location typically known as a spent fuel pool. In a traditional pressurized water reactor, fuel is accessed by removing the reactor vessel closure head and upper internals. These components are stored within the containment building while a specialized refueling crane, supported from an operating deck above the reactor vessel flange, moves fuel assemblies one at a time from the reactor vessel to a fuel transfer canal. The transfer canal connects the spent fuel storage area of the plant to the reactor containment building. The fuel is downended (laid on its side) before it is moved through the transfer canal. The process is reversed to load the fuel back into the reactor vessel. The physical configuration of some pressurized water reactor designs, including an integral reactor being developed for small modular reactor plants, prevents this traditional approach to refueling from being applied directly. FIGS. 1 and 2 illustrate such a small modular reactor. FIG. 1 shows a perspective view of the reactor containment, partially cut away, to show the pressure vessel and its internal components. FIG. 2 is an enlarged view of the pressure vessel shown in FIG. 1. The pressurizer 22 is integrated into the upper portion of the reactor vessel head 28 and eliminates the need for a separate component. A hot leg riser 16 directs primary coolant from the core 14 to a steam generator 18 which surrounds the hot leg riser 16. A number of coolant pumps 26 are circumferentially spaced around the reactor vessel 10 at an elevation near the upper end of the upper internals 24. The reactor coolant pumps 26 are horizontally mounted axial flow canned motor pumps. The reactor core 14 and the upper internals 24, except for their size, are substantially the same as the corresponding components in an AP1000® reactor, offered by Westinghouse Electric Company LLC, Cranberry Twp., Pa. From the foregoing, it should be apparent that employing the traditional refueling method by flooding the reactor well above the area of the vessel flange 30 and transferring the fuel assemblies under water to a spent fuel pool by way of a transfer canal 32 that extends through the containment would not be practical with this type of containment and compact design. Accordingly, a new refueling method is desired that can accommodate compact, integral reactor designs. Furthermore, such a method is desired that can refuel such a compact containment and integral reactor design efficiently without damaging the transferred components or causing unnecessary radiation exposure. These and other objects are achieved by a method of refueling a nuclear reactor comprising a reactor vessel having an open upper end with a flange; the reactor vessel housing a core including a plurality of fuel assemblies and an upper internal structure supported above the core. A reactor vessel head with a mating flange seals off the open upper end of the reactor vessel. The refueling method comprises the steps of removing the reactor vessel head and placing the head in a first storage location outside of a path above the reactor vessel. The upper internals structure is then lifted out of the reactor vessel to a second storage location outside of the path above the reactor vessel. A cylindrical tank having an open lower end and an open upper end is installed on the reactor vessel flange and the lower end of the cylindrical tank is then sealed to the reactor vessel flange. A penetration on the side of the cylindrical tank is connected to a refueling canal that communicates the inside of the containment to a spent fuel pool outside of the containment and within the reactor building. The level of reactor coolant within the reactor vessel is raised to at least partially fill the cylindrical tank substantially to a level equal to a level of a coolant within the spent fuel pool. The refueling canal is then opened and a refueling machine supported above the cylindrical tank is employed to transfer a number of the fuel assemblies from the core and through the penetration and the refueling canal to a storage location in the spent fuel pool. In one embodiment, the step of lifting the upper internals structure includes the steps of lowering a radiation shield within the cylindrical tank, above the upper internals structure. The upper internals structure is then raised within the radiation shield and the radiation shield with the upper internals structure inside is lifted and moved to a second storage location. Preferably, the step of lifting the upper internals structure further includes the step of lowering the upper internals structure from the radiation shield to a shielded stand at the second storage location. Desirably, the shielded stand is located in a pool of coolant. In another embodiment, air is drawn into the radiation shield, filtered before it is exhausted from the radiation shield and exhausted from the radiation shield after it has been filtered. Additionally, the step of lifting the upper internals structure may include using the reactor building main crane for that purpose. Further, the method may include the step of supporting the refueling machine from the reactor vessel and preferably from above the cylindrical tank. The method may also include the step of indexing the refueling machine off of the reactor flange to locate the fuel assemblies to be moved. Preferably, the step of raising the level of reactor coolant is accomplished with an existing reactor vessel penetration and the cylindrical tank is sealed to the reactor vessel flange. The steps of this embodiment are sequentially illustrated in FIGS. 3-14. This embodiment uses a temporarily installed refueling machine 36 attached directly to the reactor vessel 10. The machine 36 can use the reactor vessel stud holes 38 or a similar feature to secure and align itself to the reactor vessel 10. The machine preferably includes a shielded tank 40 that is open at the top and bottom. The tank seals to the flange 30 of the reactor 10 by contacting the mating surface of the vessel. An O-ring or similar soft seal can be used to limit leakage. Pressure applied to make the seal is either provided by the weight of the tank or by mechanical fasteners such as the studs 42. Any leakage would be detected by the existing reactor vessel leak-off lines provided between the O-ring seals used to seal the reactor vessel during plant operation. The tank 40 has a penetration 44 perpendicular to the center line of the tank. This penetration provides a means of transferring the fuel assemblies in the core 14 from the reactor vessel 10 to a fuel storage area, i.e., the spent fuel pool 46 of the plant. This penetration forms at least part of the fuel transfer canal 48. The temporary refueling machine assembly 36 with its tank 40 and penetration 44 are lowered onto the reactor vessel 10 using the reactor building main crane 50 after the closure flange 52 on the reactor vessel head 28 and the upper internals 24 have been removed from the reactor vessel 10 to provide access to the fuel assemblies in the core 14. In the case of many integral reactor designs, the steam generator 18, pressurizer assembly 22 and upper internals are removed from the reactor vessel to gain access to the fuel. The tank penetration 44 is connected to a mating penetration 32 in the containment vessel 12 that communicates with the spent fuel pool 46. The water level in the temporary tank 40 is then raised by introducing additional water inventory to the reactor vessel through an existing reactor penetration such as the chemical and volume control system penetration. When the water level in the temporary tank 40 is substantially even with the water level in the spent fuel pool 46, the transfer canal 48 can be opened to allow fuel to be passed from the reactor to the spent fuel pool. A transfer cart 54 is used to move the fuel through the transfer canal 48. The cart 54 is extended through the canal to the reactor vessel to accept the fuel assembly from a temporary refueling machine 56 which is supported on top of the shielded tank 40. As in traditional plants, the cart 54 has a rotating basket to allow the fuel assembly to be downended, i.e., rotated to a horizontal position, to minimize the required diameter needed for the transfer canal. Once the cart 54 has passed from the reactor vessel 10 to the spent fuel pool area 46, the basket is upended again and a traditional fuel handling machine 58 removes the fuel from the cart. This traditional fuel handling machine 58 places the fuel into a temporary storage rack 60 until it can be transferred to dry storage or a reprocessing facility. Thus, this invention addresses a number of design challenges associated with integral pressurized water reactors and small modular reactors. The compact high pressure containment vessels 12 such as the one shown in FIG. 1, used in the design of small modular reactors, do not have room to include a refueling pool above the vessel which is typical for traditional operating pressurized water reactors. It is not possible to fill the containment vessel with water during refueling due to contamination concerns and sensitive equipment within the containment that cannot be designed for submergence. Instead, in accordance with the foregoing embodiment, a temporary refueling pool is provided by securing a tank to the reactor vessel that seals to the reactor vessel flange mating surface. The water within the temporary tank 40 provides shielding and is a filter should a fuel element leak develop. The tank 40 itself (including the structure of the transfer canal) provides additional shielding due to the thickness of the wall material. Since the containment vessel cannot be filled with water, the upper internals 24 cannot remain under water during removal to storage. A specifically designed lifting rig that is both shielded and positively vented is used to remove the upper internals 24 from the reactor vessel 10. A shielding bell 64 fits over a flange on the upper internals and a portion of the lifting rig structure passes through holes in the shielding bell to engage the features provided on the upper internals for lifting. To prevent airborne contamination, a combination of fan and HEPA filter draw air into the shielding bell 64 from the bottom and filter the air in the bell before it is discharged. The lifting rig 62 is used to place the upper internals in a shielded stand in the reactor building outside of the containment 12. The internals may be shielded in the storage location by being submerged in water or borated water. The distance between the reactor building operating deck and the fuel assemblies in the core 14 is much larger in small modular reactor designs than is experienced in conventional pressurized water reactor plants. Modifications to traditional refueling machines to operate at such a distance would not be practical due to dimensional control, ability to monitor visually and seismic considerations. This embodiment secures a temporary refueling machine 56 to the reactor flange which moves the machine much closer to the reactor core. The fuel is raised into a mask 70, traveling a distance similar to that of a traditional refueling machine. The reactor vessel provides a very stable attachment point that is indexed to the fuel allowing for precise alignment. FIGS. 3 through 14 are schematics of a reactor plant sequentially showing the different stages of the refueling method described heretofore. FIG. 3 shows a small modular reactor plant with a compact containment 12 during normal operation. For refueling externally flooded plant building containments, the flooded area figuratively represented by the area 74 in FIG. 3, the water level is first lowered and the upper portion 34 of the containment vessel 12 is removed and stored which opens the containment 12 as shown in FIG. 4. The reactor head 28, containing the steam generator 18 and pressurizer 22, is then removed using the main building crane and stored off to the side of the reactor building as illustrated in FIG. 5. The shielded upper internals lifting rig 62 is lowered into place with the reactor building main crane 50 and secured to the upper internals 24. As shown in FIG. 6, the upper internals 24 are drawn up into the shielded bell 64 of the upper internals lifting rig 62. During the lift, the fan and HEPA filter 66 prevent airborne contamination from being released into the containment building atmosphere. The upper internals are then placed into the upper internals storage stand 68 as shown in FIG. 8. A temporarily installed refueling machine 36 is then attached to the reactor vessel flange 30 as shown in FIG. 9. The shielded tank 40, which is part of the temporarily installed refueling machine 36, is sealed to the reactor vessel flange 30 at the mating surface and is attached to the transfer canal 48. The water level in the tank 40 is then raised to substantially the level of the spent fuel pool 46 using an existing reactor penetration as shown in FIG. 10. The fuel transfer cart 54 travels through the transfer canal 48 as fuel is being raised in the mast 70 of the temporary refueling machine 36 as shown in FIG. 11. A basket on the fuel transfer cart 54 is rotated and fuel is placed into the basket as shown in FIG. 12. The basket is then rotated to a horizontal position and the fuel downended as it enters the transfer canal 48 as shown in FIG. 13. After passing through the transfer canal 48, the basket is rotated back to a vertical position and the fuel is removed by the spent fuel handling machine 58 and placed into temporary storage racks 60 as shown in FIG. 14. The process is repeated as required to remove fuel from the reactor vessel 10. The process is reversed to bring fuel from the spent fuel pool 46 to the reactor vessel 10. When refueling the core has been completed the transfer canal 48 can be closed and the water level within the temporary refueling machine 36 can be lowered to within the reactor vessel 10 and the temporary refueling machine 36 can be removed by the main building crane to a storage location. The shielded upper internals lifting rig can then be used to raise the internals into the bell and lowered into the core. After the internals are secured the main building crane can be used to replace the reactor head 28 on the vessel 10 and the top of the containment 34 can then be restored to prepare the reactor system for operation. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	claims	1. A canister for the transportation and storage of nuclear fuel assemblies, comprising a basket assembly receivable into a canister shell, the basket assembly comprising:a plurality of interlocking structural plates that are disposed in spaced parallel relationship to each other in a first direction and a plurality of structural plates disposed in a second direction transverse to the first direction;said structural plates comprising portions defining spaced apart transverse slots formed along the structural plates, with the slots of the structural plates disposed in a first direction engaging with the slots of the structural plates disposed in the second transverse direction; andwherein the structural plates include a plurality of layers comprising a first layer composed of structural material, a second layer composed of a heat conducting material selected from the group consisting of aluminum and copper, and a third layer composed of a neutron absorbing material, wherein the third layer is between the first and second layers, wherein each of the first, second, and third layers include aligned transverse slots, and wherein the first, second, and third layers of the structural plates extending in the second transverse direction extend continuously from a first proximity with the inner surface of the canister shell to a second proximity with the inner surface of the canister shell, wherein the second proximity is distanced from the first proximity. 2. The canister according to claim 1, wherein the first and second layers of the structural plates are formed to encase the inner layer of the structural plates. 3. The canister according to claim 2, wherein the margins of at least the first layer of the structural plates tends over the edges of the third layer and is joined to the first layer of the opposite side of the structural plate. 4. The canister according to claim 1, wherein the first layer is composed of material selected from the group consisting of a high strength steel, a low alloy steel, a high strength and low alloy steel, a carbon steel, and stainless steel. 5. The canister according to claim 1, wherein the heat conducting layer is composed of at least one material selected from the group consisting of aluminum and copper. 6. The canister according to claim 1, wherein the neutron absorbing layer of the structural plate is a metallic, ceramic, or composite material containing an element that absorbs thermal neutrons. 7. The canister according to claim 1, wherein the layers of the structural plate are fastened together in face-to-face relationship to each other. 8. The canister according to claim 7, wherein the layers of the structural plate are fastened together by fasteners selected from the group consisting of: threaded fasteners, rivets, and weld pins. 9. The canister according to claim 1, further comprising an elongated locking key extending along and engaging with adjacent edge portions of adjacent structural plates to lock said adjacent edge portions together and to align the adjacent edge portions of the structural plates together. 10. The canister according to claim 9, wherein grooves are formed along the edge portions of the structural plates, said grooves are sized to closely receive the locking key therein, said locking key closely receivable within the grooves of the adjacent edge portions of adjacent structural plates. 11. The canister according to claim 9, wherein through holes are formed in the structural plates whereby the locking key passes through the through holes of the structural plates that extend transversely to the length of the locking keys. 12. The canister according to claim 1, wherein the exterior surface of the structural plates are treated to enhance radiative heat transfer from the fuel assemblies stored in the canister. 13. The canister according to claim 1, wherein the structural plates are treated with a hydrophobic coating to facilitate drying of the structural plates. 14. The canister according to claim 1, further comprising transition rails extending lengthwise of the canister at the outer perimeter of the basket assembly to interconnect the structural plates. 15. The canister according to claim 14, wherein the transition rails have an outer curvature in the direction transverse to the length of the transition rails that correspond to the circumference of the canister. 16. The canister according to claim 15, wherein the transition rails are at least in part hollow to receive a stiffening structure engageable within the hollow interior of the transition rails to enhance the structural integrity and rigidity of the transition rails. 17. A canister for the transportation and storage of nuclear fuel assemblies, comprising a basket assembly receivable into a canister shell, the basket assembly comprising:a plurality of interlocking structural plates that are disposed in spaced parallel relationship to each other in a first direction and a plurality of structural plates disposed in a second direction transverse to the first direction;said structural plates comprising portions defining spaced apart transverse slots formed along the structural plates, with the slots of the structural plates disposed in a first direction engaging with the slots of the structural plates disposed in the second transverse direction, wherein the structural plates extending in the second transverse direction extend continuously from a first proximity with the inner surface of the canister shell to a second proximity with the inner surface of the canister shell, wherein the second proximity is distanced from the first proximity; andwherein the structural plates include a plurality of layers comprising a first layer composed of material selected from the group consisting of a high strength steel, a low alloy steel, a high strength and low alloy steel, a carbon steel, and stainless steel, a second layer composed of a heat conducting material selected from the group consisting of aluminum and copper, and a third layer including a neutron absorbing material, wherein the third layer is between the first and second layers, wherein each of the first, second, and third layers include aligned transverse slots.
	description	This application relates to systems and methods for achromatically bending a beam of charged particles by about 90° during radiation treatment. External beam radiation therapy is one of the available non-invasive methods to treat a pathological anatomy (e.g., tumor, lesion, vascular malformation, nerve disorder, etc.). In one type of external beam radiation therapy, an external radiation source directs a sequence of x-ray beams at a target volume, e.g., a tumor, from multiple angles, with the patient positioned so the target volume is at the center of rotation (isocenter) of the beam. As the angle of the radiation source changes, every beam passes through the target volume, but passes through a different area of healthy tissue on its way to and from the target volume. Image-guided radiation therapy (IGRT) systems include gantry-based systems and robotic-based systems. In gantry-based systems, the radiation source, e.g., a linear accelerator (LINAC), is mounted on a gantry that moves the source around a center of rotation (isocenter) in a single plane. The radiation source may be rigidly attached to the gantry or attached by a gimbaled mechanism. Each time a radiation beam is delivered to a target volume during treatment, the axis of the beam passes through the isocenter. Radiation beam delivery is, therefore, limited by the rotation range of the radiation source mounted on the gantry, the angular range of the gimbaled mechanism (if present), and by the number of degrees of freedom available on a patient positioning system. Additionally, the shape of the radiation can be modified using a multileaf collimator. Alternatively, the treatment system has the radiation source mounted on a robotic arm with at least five degrees of freedom to enable non-coplanar delivery to a target volume. One example of such a system is the CYBERKNIFE® Robotic Radiosurgery System manufactured by Accuray Incorporated. (Sunnyvale, Calif.). One practical limitation for both gantry-based systems and robotic-based IGRT systems is space. Specifically, hospitals or other sites wishing to install such a system may have a specific room in which it is to be placed. However, the system may be too large for the room, thus requiring structural modification of the room. If the room cannot be modified within a specified budget (or at all), then it may be necessary to relocate the system; alternatively, the system's use may be entirely precluded at the site. Thus, it would be useful to reduce the size of an IGRT system. One way in which to do this is to reduce the amount of space required to generate the radiation beam (e.g., X-ray, electron, or proton beam). One example of a prior art robot-based IGRT system 100 is illustrated in FIG. 1. System 100 includes robot-based support system 110, robot-based linear accelerator (LINAC) system 120, X-ray imaging sources 131, and detectors 132. Robot-based LINAC system 120 includes LINAC 121 and robotic arm 122. Robot-based support system 110 includes patient treatment couch 111, robotic arm 112, track 114, and column 115. Responsive to instructions from a controller (not shown), robot-based support system 110 moves robotic arm 112 in any suitable direction, e.g., along track 114 and/or column 115, so as to adjust the position and/or orientation of patient treatment couch 111 and thus appropriately position the patient before and/or during the radiation treatment, in accordance with a treatment plan. Also responsive to instructions from the controller, robot-based LINAC system 120 moves LINAC 121 to a desired position and orientation using robotic arm 122, and generates radiation of the desired type, energy, field, and dose using LINAC 121, again in accordance with the treatment plan. X-ray imaging sources 131 and detectors 132 are configured to obtain x-ray images of the patient or nearby anatomical structures responsive to instructions from the controller, e.g., at appropriate times before and during the radiation treatment. Each of x-ray imaging sources 131 is arranged at a predetermined angle relative to vertical, e.g., at 45° from vertical, such that x-ray radiation generated by that source passes through the target volume and is received by corresponding detector 132. Based on the received radiation, each of detectors 132 obtains an x-ray image of the target volume. The pair of thus-obtained images may be referred to as “stereoscopic x-ray images,” and is provided from detectors 132 to the controller for use in guiding irradiation of the patient with LINAC 121. As is familiar to those skilled in the art, LINACs are designed to accelerate charged particles along a linear pathway. Generally, a LINAC includes a charged particle source, e.g., a source of electrons, protons, or ions, and an evacuated chamber along which the particles are accelerated. Depending on the type of charged particle, the evacuated chamber may be relatively long. For example, chambers for the acceleration of electrons may be between 0.5 and 1.5 meters long. Orienting such a chamber generally perpendicularly to the patient treatment couch, as is the case for LINAC 121 illustrated in FIG. 1, may substantially increase the overall height of the system. As such, the space requirements for installing and operating the system may increase correspondingly, thus potentially presenting practical problems for installing the system at space-constrained sites. Embodiments of the invention provide systems and methods for achromatically bending a beam of charged particles by about 90°. Under one aspect of the present invention, a system for achromatically bending a particle beam by about 90° includes first, second, third, and fourth bending magnets serially arranged along a beam path of the particle beam. The first and fourth bending magnets may be configured to generate a positive field gradient that defocuses the particle beam in a bend plane. The second and third bending magnets may be configured to generate a negative field gradient that focuses the particle beam in the bend planes. The first, second, third, and fourth bending magnets collectively bend the particle beam by about 90°, and in some embodiments, they each bend the particle beam by about 22.5° about an approximate center of curvature. The first and fourth bending magnets may have substantially the same construction as one another, and the second and third bending magnets also may have substantially the same construction as one another; however, the construction of the second and third bending magnets may be different than that of the first and fourth bending magnets. The first and fourth bending magnets each may include an iron cored, dipole electromagnet having pole faces that are symmetrically disposed about the bend plane and are inclined relative to each other and shaped so as to generate a positive field gradient. A positive field gradient is such that the magnetic induction decreases with increasing distance from the approximate center of curvature. The second and third bending magnets each may include an iron cored, dipole electromagnet and having pole faces that are symmetrically disposed about the bend plane and are inclined relative to each other and shaped so as to generate a negative field gradient. A negative field gradient is such that the magnetic induction increases with increasing distance from the approximate center of curvature. The pole faces of the first, second, third, and fourth bending magnets may be shaped such that the positive field gradient is substantially weaker than the negative field gradient. The pole faces of the first, second, third, and fourth bending magnets may each be approximately hyperbolically shaped. In some embodiments, the pole faces are shaped so as to introduce higher-order magnetic field components so as to control geometric and chromatic aberrations in the beam which may be produced by the simple linear field gradients heretofore described. In one embodiment, a midpoint between the second and third bending magnets defines a mirror plane. The first and fourth bending magnets may be positioned substantially symmetrically across the mirror plane from one another, and the second and third bending magnets may be positioned substantially symmetrically across the mirror plane from one another. Under another aspect of the present invention, a method for achromatically bending a particle beam by about 90° may include bending the particle beam with a first bending magnet that defocuses the particle beam in a first plane with a positive field gradient, and then bending the particle beam with a second bending magnet that focuses the particle beam in the first plane with a negative field gradient, and then bending the particle beam with a third bending magnet that focuses the particle beam in the first plane with a negative field gradient, and then bending the particle beam with a fourth bending magnet that defocuses the particle beam in the first plane with a positive field gradient, wherein the first, second, third, and fourth bending magnets collectively bend the particle beam by about 90°. Embodiments of the invention provide systems and methods for achromatically bending a beam of charged particles by about 90° during radiation treatment. By “achromatic” it is meant that particles having energies that lie within a certain range, e.g., within plus or minus 10% of a central energy value, or even within plus or minus 15% of a central energy value, will all be redirected in approximately the same direction as one another and in approximately the same position as one another. Such systems and methods may significantly reduce the space requirements for installing and operating radiation treatment systems that include a linear accelerator (LINAC) for generating radiation to be used in treating a patient. Specifically, whereas prior art radiation treatment systems such as system 100 illustrated in FIG. 1 may include a long LINAC that is oriented generally perpendicularly to the patient during treatment, which can significantly increase the space requirements for installing and operating the system, the systems and methods provided herein allow a LINAC to be oriented generally parallel to the patient and achromatically bend the output of the LINAC by about 90°. Because the LINAC is generally significantly longer than it is wide, such a parallel orientation may result in significant space savings, thus allowing the radiation treatment system to be installed in smaller spaces than may otherwise be possible, and/or to allow movement of the radiation source to otherwise inaccessible locations within the treatment space. Additionally, while some existing radiation treatment systems may include systems that achromatically bend a particle beam by about 270°, such bending systems may require about three times more material than the present systems. For robot-mounted bending systems, such a weight differential may significantly impact practical implementation of the bending system. For example, FIG. 2A schematically illustrates a plan view of a LINAC 221 and beam bending system 210, according to some embodiments of the present invention. LINAC 221 is mounted on robotic arm 222, which may be similar to robotic arm 122 illustrated in FIG. 1. LINAC 221 emits a particle beam 20 in a direction that is substantially parallel to the orientation of LINAC 221. Beam bending system 210 is coupled to LINAC 221 and receives particle beam 20 from LINAC 221. Beam bending system then bends the particle beam 20 by about 90° from an axis approximately parallel to the orientation of LINAC 221, and emits the bent beam at output 213. The bent beam is thus emitted substantially perpendicular to the orientation of LINAC 221. Note that robotic arm 222 and beam bending system 210 need not necessarily be in the same plane as one another. For example, beam bending system 210 may be oriented in the x-y plane, as shown, and robotic aim may be oriented in the y-z plane, e.g., orthogonally to beam bending system 210. As can be seen from FIG. 2A, beam bending system 210 allows LINAC 221 to be mounted so as to allow for a relatively compact arrangement. Specifically, LINAC 221 has a length D1 and a width D2 that is substantially smaller than length D1, and beam bending system 210 has a bending radius of D3. Together, LINAC 221 and beam bending system 210 have a dimension in the x-direction about equal to D1 plus D5, where D5 is the lateral dimension of the beam bending system in the x-direction. LINAC 221 and beam bending system 210 have a dimension in the y-direction about equal to D2 plus D4, where D4 is the lateral dimension of the beam bending system in the y-direction that extends past the edge of the LINAC. Usefully, the sum of D2 and D4 is significantly smaller than is D1, e.g., less than half of D1, thus resulting in significant space savings in the y-direction and potentially allowing the system to be installed at more space-constrained sites and/or to be moved to otherwise inaccessible treatment orientations. As used herein, the terms “about” and “approximately” mean plus or minus 10% or less. For example, beam bending system 210 may bend particle beam 20 by between 81° and 99°, or between 82° and 98°, or between 83° and 97°, or between 84° and 96°, or between 85° and 95°, or between 86° and 94°, or between 87° and 93°, or between 88° and 92°, or between 89° and 91°, or between 89.5° and 90.5°, or between 89.8° and 90.2°, or between 89.9 and 90.1°, or even exactly 90°. FIG. 2B schematically illustrates one embodiment of a beam bending system 210 suitable for use with LINAC 221. Beam bending system 210 receives particle beam 20 from LINAC 221 (not shown in FIG. 2B), and bends the particle beam by about 90°, resulting in redirected beam 20′. In one embodiment, redirected beam 20′ is a substantially round beam of electrons of selected energy, e.g., 6 MeV or 10 MeV, that irradiates X-ray target 250 (e.g., a tungsten target) to generate X-rays for use in radiation therapy. Other suitable charged particles may be used, with or without a target, as one of skill in the art will appreciate. Beam bending system 210 includes first, second, third, and fourth bending magnets 201, 202, 203, and 204, respectively, that are serially arranged along the beam path of particle beam 20, and are coupled to a suitable mount 211. In the illustrated embodiment, each of the first, second, third, and fourth bending magnets 201, 202, 203, 204 each bend the particle beam by about 22.5° with respect to an approximate center of curvature, e.g., center of curvature 220 for magnet 204 (centers of curvature for the other magnets are not shown). However, it should be appreciated that each magnet may bend the particle beam by any desired angle, such that collectively the magnets 201-204 bend the beam by about 90°. For example, the first and fourth magnets 201, 204 may each bend the beam by about 15°, and the second and third magnets may each bend the beam by 30°. In embodiments where the first and fourth magnets bend the beam by about the same angle as one another, and the second and third magnets also bend the beam by about the same angle as one another, the collection of magnets may be approximately symmetrical across mirror plane M and may provide imaging at approximately 1:1 magnification of the particle beam from the first to the fourth magnet. In embodiments where the first and fourth bend the beam by a different angle as one another, and the second and third magnets bend the beam by a different angle than one another, the collection of magnets may be asymmetrical across mirror plane M and may provide imaging at a ratio of other than 1:1 magnification of the particle beam from the first to the fourth magnet. Any suitable arrangement of magnets may be used, providing any desired imaging ratio. However, it will be appreciated that imaging at an approximately 1:1 magnification ratio may be useful in some circumstances because the particle beam will have similar spatial profiles both before and after the bend. In some embodiments, an aperture 230 may be provided between the second and third bending magnets and shaped so as to spatially filter particles that stray beyond an acceptable distance from beam path 20, e.g., particles that have energies outside of a pre-determined energy spread. As illustrated below with respect to FIGS. 3A-3B, the beam waist of the particle beam may be relatively large in the bend plane and relatively small in the non-bend plane. Where the charged particles are electrons, any electrons that strike aperture 230 may cause the emission of X-rays in the direction in which those electrons had been travelling, e.g., as denoted by the dash-dot line 240. As illustrated in FIG. 2B, the first and second bending magnets 201, 202 are separated by a lateral center-to-center drift distance D5. The second and third bending magnets 202, 203 are separated by a lateral center-to-center drift distance D6. The third and fourth bending magnets 203, 204 are separated by a lateral center-to-center drift distance D7 that in some embodiments may be substantially the same as D5. Note that if drift distance D7 is different than drift distance D5, then the bend angle of fourth bending magnet 204 may be different from the bend angle of the first bending magnet 201 by the inverse ratio (D5/D7) to achieve achromaticity. In some embodiments, a midpoint between the second and third bending magnets 202, 203 defines a mirror plane M, across which the positions of the bending magnets may be substantially symmetrical. That is, the first and fourth bending magnets 201, 204 may be positioned substantially symmetrically from each other across mirror plane M, and the second and third bending magnets 202, 203 also may be positioned substantially symmetrically from each other across mirror plane M. Thus, in some cases, the first and fourth bending magnets may be referred to as the “outer” bending magnets, and the second and third bending magnets may be referred to as the “inner” bending magnets. The distance D8 from beam path 20 to the top edge of mount 211, plus the distance D9 from beam path 20 to the bottom edge of mount 211, defines the lateral dimensions of beam bending system 210. Note that other embodiments need not necessarily include mirror symmetry across mirror plane M, and may in some circumstances image the beam with other than 1:1 magnification. In some embodiments, the first and fourth bending magnets 201, 204 may be substantially the same as one another, e.g., have substantially the same construction, materials, and configuration as one another. Similarly, the second and third bending magnets 202, 203 may be substantially the same as one another; however, the construction of the first and fourth bending magnets 201, 204 may be different from the construction of the second and third bending magnets 202, 203. In one embodiment, each of the first, second, third, and fourth bending magnets 201-204 are rectangular, laminated magnets, where the first and fourth magnets have pole faces that are the same as one another, and the second and third magnets have pole faces that are the same as one another but different than those of the first and fourth magnets. Such laminated magnets may be particularly useful where the field in the bend magnets may be changed relatively rapidly by a fast regulating power supply, thus facilitating achromatic bending of particle beams having central energies and/or energy spreads that differ significantly from one another. In an alternative embodiment described in greater detail below with respect to FIG. 11, trapezoidal and/or wedge-shaped magnets instead may be used to bend the particle beam. Preferably, the first, second, third, and fourth magnets 201-204 are configured so as to achromatically bend, in a bend plane, a full emittance particle beam 30 having a finite energy spread such as illustrated in FIG. 3A, in which the bending magnets are arranged along a straight line merely for simplicity of illustration and so angles and transverse positions relative to a central reference particle can be magnified. Specifically, the first and fourth bending magnets 201, 204 each may be configured to defocus the particle beam 30 in the bend plane, and have a positive field gradient. The magnitude of the field gradients of the first and fourth bending magnets 201, 204 may be substantially the same as one another. In contrast, the second and third bending magnets 202, 203 each may be configured to focus the particle beam 30 in the bend plane and have a negative field gradient. The magnitude of the field gradients of the second and third bending magnets 202, 203 may be substantially the same as one another, and substantially higher than the field gradients of the first and fourth bending magnets 201, 204, as discussed in greater detail below. The net result, illustrated in FIG. 3A, is that in the bend plane beam 30 diverges between the first and second bending magnets 201, 202, in the drift region defined by D5; converges to a waist and then diverges between the second and third bending magnets 202, 203 in the drift region defined by D6; converges between the third and fourth bending magnets 203, 204 in the drift region defined by D7; and is approximately collimated in the drift region after it passes through magnet 204, with imaging at 1:1 magnification of the beam from the first to the fourth magnet in the bend plane. FIG. 3B schematically illustrates the envelope of the same full emittance, finite energy spread particle beam 30 as shown in FIG. 3A, but instead in the non-bend plane. The first and fourth bending magnets 201, 204 each may be configured to provide imaging at 1:1 magnification of the particle beam 30 in the non-bend plane, yielding a narrow beam waist positioned approximately mid-way between the second and third bending magnets 202, 203. Note that in such a design, the use of four magnets is particularly useful for controlling the focusing and defocusing of the particles in the bend and non-bend planes. Preferably, the first, second, third, and fourth magnets 201-204 are also configured so as to achromatically bend a zero-emittance particle beam 31 having a finite energy spread, in the bend plane, in the manner illustrated in FIG. 3C, in which the bending magnets are arranged along a straight line merely for simplicity of illustration. Beam 31 includes particle 301 having the same energy as the central energy of the spread and relative to which magnified angles and transverse positions of the other particles are shown, and particles 302 and 303 having energies that are different than one another but equal and oppositely spaced from the central energy. First bending magnet 201 causes particles 302 and 303 to diverge from one another and from particle 301. Second bending magnet 202 redirects particles 302 and 303 so that their trajectories become parallel to one another and to that of particle 301. Third bending magnet 203 causes the trajectories of particles 302 and 303 to converge toward that of particle 301 at fourth bending magnet 204. Fourth bending magnet redirects particles 301, 302, and 303 such that their trajectories recombine with one another, again forming a zero emittance particle beam with finite energy spread 31′. The parallel and collinear paths of particles 301, 302, and 303 in the exit beam 31′ evidence the achromaticity of the system. Preferably, the first, second, third, and fourth magnets 201-204 are also configured so as to achromatically bend a full emittance monochromatic particle beam 32 in the bend plane in the manner illustrated in FIG. 3D, in which the bending magnets are arranged along a straight line merely for simplicity of illustration. First bending magnet 201 defocuses beam 32, then second bending magnet 202 focuses beam 32 to a relatively small beam waist that appears as a crossover at the midpoint between the second and third bending magnets. Drift distance D6 is selected to achieve this condition, which also provides the system with beam imaging at 1:1 magnification from input to output in the bend plane, as compared to that which could be achieved with a three-magnet system. Third bending magnet 203 focuses beam 32, and then the fourth bending magnet 204 collimates the beam, resulting in collimated beam 32′. Preferably, the beam path illustrated in FIG. 3A is the root mean square (RMS) of the beam paths illustrated in FIGS. 3C and 3D. Note that charged particles within the actual particle beam that is input to beam bending system 210 may have finite emittance and a finite energy spread, that is, the beam is not necessarily mono-directional and/or monochromatic. For example, particles within the particle beam may have energies that are within plus or minus 10%, or even within plus or minus 15%, of a central energy value; that is, the beam may have a 20% or even a 30% full width energy spread. Preferably, the first, second, third, and fourth bending magnets are constructed and arranged such that beam bending system 210 is substantially achromatic to first order. By “substantially achromatic to first order” it is meant that particles having energies within a specified range of energies will be bent in substantially the same direction as one another and that their position will also be substantially independent of their energies. In one embodiment, beam bending system 210 bends in substantially the same direction particles that deviate from a central energy by plus or minus 7% of a central energy value. Additionally, the magnets also may be constructed and arranged to image the beam in the both the bend and non-bend planes from the first to the fourth magnet with a magnification of one. Such beam shaping characteristics and achromaticity may be achieved, for example, by (1) selecting the field index and/or focal length of the inner (second and third) magnets so as to image the beam in the bend plane from approximately the center of the first magnet to the center of the fourth magnet (achromaticity condition); (2) selecting the field index and/or focal length of the outer (first and fourth) magnets so as to produce imaging at 1:1 magnification in the non-bend plane; and (3) selecting the drift length D6 between the inner (second and third) magnets so as to produce imaging at 1:1 magnification in the bend plane. In one embodiment, such imaging at 1:1 magnification is achieved for full emittance finite energy-spread beams, zero emittance finite energy-spread beams, as well as full emittance monochromatic beams, by constructing the system to have mirror symmetry about a mirror image plane M. Note, however, that such mirror symmetry is not required, and imaging at magnifications other than 1:1 may be provided. System 210 illustrated in FIGS. 2A-2B may be used to deflect any suitable beam of charged particles by about 90°. It will be appreciated, however, that the specific dimensions of system 210 and configurations of bending magnets 201-204 may be adjusted based on the masses and energies of the particles in particle beam 20. For example, electrons have relatively low masses and so may be deflected over a shorter distance than protons or heavier ions. Exemplary dimensions of the system of FIGS. 2A-2B, as configured for use with electrons of energies 6 MeV or 10 MeV, are set forth below in Table 1. TABLE 1Dimension6 MeV Electrons10 MeV ElectronsD5, D7 25 mm 25 mmD6 48 mm 48 mmD8154 mm165 mmD9 29 mm 30 mm The system 210 illustrated in FIGS. 2A-2B may be used to bend a beam of charged particles by about 90° during radiation treatment, according to method 400 illustrated in FIG. 4. The skilled artisan will readily appreciate modifications that may be made to the method of FIG. 4 for use with other types of radiation treatment systems, such as the gantry-based system illustrated in FIG. 12 and described in greater detail below. Referring to FIG. 4, a beam of charged particles is generated, using any suitable techniques known in the art (step 410). For example, LINAC 221 illustrated in FIG. 2A may be used to generate a beam of charged particles such as electrons or protons, or even heavier particles such as boron, carbon, or neon. The beam may have full or finite emittance and a finite energy spread. Then, the particle beam is bent by about 22.5°, and defocused in the bend plane with a positive field gradient (step 420 of FIG. 4). For example, as illustrated in FIG. 2B, beam 20 may pass through first bending magnet 201, which imparts about a 22.5° bend to beam 20 and defocuses the beam. After passing through first bending magnet 201, beam 20 diverges in the bend plane and converges in the non-bend plane as it passes through the drift region of dimension D5 between magnets 201 and 202, as illustrated in FIGS. 3A-3B. Then, the particle beam is bent by about 22.5°, and focused in the bend plane with a negative field gradient (step 430 of FIG. 4). The field gradient used in step 430 may have a larger magnitude than that used in step 420. As illustrated in FIG. 2B, beam 20 may pass through second bending magnet 202, which imparts about a 22.5° bend to beam 20 and focuses the beam. After passing through second bending magnet 202, beam 20 converges, reaches a waist in both the bend and non-bend planes at a mid-point of the drift region between the second and third magnets, and then diverges as it passes through the second half of the drift region of dimension D6 between magnets 202 and 203, as illustrated in FIG. 3A. Optionally, beam 20 also passes through aperture 230, which is positioned approximately at the midpoint between magnets 202 and 203, and which spatially filters off energy particles that stray by a predetermined distance in the bend plane from the desired path of beam 20, as illustrated in FIG. 2B. In one embodiment, to achieve a substantially round output beam 20′, collimation is done before or after system 210 because in the bend plane, the beam size is determined by the energy spread in the drift regions between the bend magnets. Then, the particle beam is again bent by about 22.5°, and focused in the bend plane with a negative field gradient (step 440 of FIG. 4). For example, as illustrated in FIG. 2B, beam 20 may pass through third bending magnet 203, which imparts about a 22.5° bend to beam 20 and focuses the beam. After passing through third bending magnet 203, beam 20 is convergent in the bend plane and divergent in the non-bend plane as it passes through the drift region of dimension D7 between magnets 203 and 204, as illustrated in FIGS. 3A and 3B. Then, the particle beam is again bent by about 22.5°, and defocused in the bend plane with a positive field gradient (step 450 of FIG. 4). For example, as illustrated in FIG. 2B, beam 20 may pass through fourth bending magnet 204, which imparts about a 22.5° bend to beam 20 and defocuses the beam in the bend plane. After passing through fourth bending magnet 204, beam 20 has been deflected by a total of about 90° and is approximately collimated, resulting in deflected beam 20′ as illustrated in FIG. 2B. Optionally, in embodiments where deflected beam 20′ is an electron beam, the deflected beam then may be used to irradiate an X-ray target (step 460 of FIG. 4). For example, as illustrated in FIG. 2B, beam 20′ may for example be an electron beam that impinges on X-ray target 250 and thus generates an X-ray beam. The X-rays thus generated may then be used for radiation therapy, e.g., using methods known in the art. The construction of bending magnets 201, 202, 203, and 204 will now be discussed in greater detail with references to FIGS. 5-6. FIG. 5 schematically illustrates a cross-sectional view of outer bending magnets 201, 204, past which particle beam 20 travels. In some embodiments, the first and fourth bending magnets each may include an iron cored, dipole electromagnet having pole faces that are symmetrically disposed about the bend plane and are inclined relative to each other and shaped so as to generate a positive field gradient. A positive field gradient is such that the magnetic induction decreases with increasing distance from the approximate center of curvature. It should be understood that the first and fourth bending magnets 201, 204 may in some embodiments have substantially the same construction as one another. However, in some embodiments, fourth bending magnet 204 may be arranged as a “mirror image” of first bending magnet 201. The description below applies equally to both arrangements. In FIG. 5, axis 583 lies in the median plane of bending system 210, i.e., in the bend plane. An approximately cylindrical surface 581 has axis 582 as its approximate axis. The approximate center of curvature 220 of FIGS. 2B and 11 lies on this axis, at axis 583. The beam path travels along a curve defined by the intersection of the median plane and surface 581. Outer bending magnets 201, 204 each include electromagnet 560 and core 570. Electromagnet 560 includes metal windings 561 and first and second cooling plates 562, 563 disposed above and below windings 561, which are configured to maintain windings 561 at a suitable temperature. Electromagnet 560 is substantially toroidal, with a preselected amount of current passing through windings 561 in the direction denoted by “+” and “−.” Windings 560 have a thickness DW. Note that this method of coil (winding) construction is only one of several that may be used. Other suitable constructions include wire wound, air cooled and hollow conductor with cooling water channeled through the hollow conductor. In either of these cases, there are no “cooling plates.” The specific winding technology may be determined by the specific parameters of the system, as well as practical considerations. In one illustrative embodiment, windings 561 are hollow copper conductor. Core 570 is formed of a ferromagnetic material such as iron, has an overall thickness DC, and includes three portions 571, 572, 573. First core portion 571 is disposed outside of electromagnet 560; second core portion 572 is disposed over electromagnet 560; and third core portion 573 is disposed inside of the toroid defined by electromagnet 560. The lower surface 574 of the third core portion 573 is disposed at a spaced distance from particle beam path 20, and is shaped so as to generate a magnetic field gradient effective to bend the charged particles traveling along that path by a desired angle, in one embodiment about 22.5°. Specifically, shaped lower surface 574 is inclined relative to the median plane, shaped so as to enhance the strength of the magnetic field to the left of axis 481 relative to the field to the right of the axis, yielding a positive magnetic field index n and causing the charged particles to defocus in the bend plane. In one embodiment, shaped lower surface 574 is approximately hyperbolic. FIG. 6 schematically illustrates a cross-sectional view of inner bending magnets 202, 203 past which particle beam 20 travels. The second and third bending magnets each may include an iron cored, dipole electromagnet and having pole faces that are symmetrically disposed about the bend plane and are inclined relative to each other and shaped so as to generate a negative field gradient. A negative field gradient is such that the magnetic induction increases with increasing distance from the approximate center of curvature. It should be understood that the second and third bending magnets 202, 203 may in some embodiments have substantially the same construction as one another. However, in some embodiments, third bending magnet 203 may be arranged as a “mirror image” of second bending magnet 202. The description below applies equally to both arrangements. In FIG. 6, axis 583 lies in the median plane of bending system 210, i.e., the bend plane. An approximately cylindrical surface 581 has axis 582 as its approximate axis. The approximate center of curvature 220 of FIGS. 2B and 11 lie on this axis at axis 583. The beam path 20 travels along a curve defined by the intersection of the median plane and surface 581. Note that in FIG. 6, beam path 20 is illustrated as a horizontal ellipse, reflecting the beam convergence in the non-bend plane such as illustrated in FIG. 3B. Inner bending magnets 202, 203 each include electromagnet 660 and core 670, which may have dimensions DW and DC as set forth above. Electromagnet 660 includes metal windings 661 and first and second cooling plates 662, 663 disposed above and below windings 661, which are configured to maintain windings 661 at a suitable temperature. Electromagnet 660 is substantially toroidal, with a preselected amount of current passing through windings 661 in the direction denoted by “+” and “−.” Core 670 is formed of a ferromagnetic material such as iron, and includes four portions 671, 672, 673, and 675. First core portion 671 is disposed outside of electromagnet 660; second core portion 672 is disposed over electromagnet 560; third core portion 673 is disposed inside of the toroid defined by electromagnet 560; and fourth core portion 675 is disposed beneath electromagnet 560. The lower surface 674 of the third core portion 673 is disposed at a spaced distance from particle beam path 20, and is shaped so as to generate a magnetic field gradient effective to bend the charged particles traveling along that path by about 22.5°. Specifically, shaped lower surface 674 is declined relative to the median plane 583 and shaped so as to enhance the strength of the magnetic field to the right of axis 581 relative to the field to the left of the axis, yielding a negative magnetic field index n and causing the charged particles to focus in the bend plane. In one embodiment, shaped lower surface 674 is approximately hyperbolic. The decline of shaped lower surface 674 illustrated in FIG. 6 may also be seen to be significantly greater than the incline of shaped lower surface 574 illustrated in FIG. 5, thus yielding a greater field index n. It should be understood that the dimensions and materials used in the inner and outer bending magnets, as well as the shape of lower surfaces 574, 674, may be modified based on the particular type and energy of charged particles to be bent, as well as the desired angle through which the particles are to be bent and the amount of focusing desired in the bend and non-bend planes. Exemplary dimensions and parameters of the outer and inner bending magnets illustrated in FIGS. 5-6, as configured for use with electrons of energy 10 MeV, are set forth below in Table 2. TABLE 2Outer Magnets Inner Magnets Dimension/Parameter(201, 204)(202, 203)DW30 mm30 mmDW/D30.042860.04286DC82 mm82 mmWinding Current4000 A4000 AB-Field at Origin (0,0)5199.5 Gauss5635.5 GaussBo0.5 Tesla0.5 TeslaField Index n at Origin (0,0)+2.357−6.801 The performance characteristics of inner and outer bending magnets configured as illustrated in FIGS. 5-6 and having the dimensions and parameters set forth in Table 2 were modeled using two-dimensional POISSON modeling for an electron beam 20 of energy 10 MeV, the results of which are set forth in FIGS. 7-10. Specifically, FIG. 7 is a magnetic flux plot of a POISSON two-dimensional model of outer bending magnets 201, 204, in which the magnetic field lines are denoted 590. FIG. 8 is a plot of the y-direction magnetic field component 591 and the z-direction magnetic field 592 component versus transverse position near the origin (0,0), e.g., near beam path 20, from the model of FIG. 7. As can be seen in FIG. 8, the y-direction magnetic field component 591 is substantially zero in this region, while the z-direction magnetic field component 591 varies smoothly and increases from right to left, with an inflection point around −0.8 cm. FIG. 9 is a magnetic flux plot of a POISSON two-dimensional model of inner bending magnets 202, 203, in which the magnetic field lines are denoted 690. FIG. 10A is a plot of the y-direction magnetic field component 691 and the z-direction magnetic field component 692 versus transverse position near the origin (0,0), e.g., near beam path 20, from the model of FIG. 9. As can be seen in FIG. 10A, the y-direction magnetic field component 691 is substantially zero in this region, while the z-direction magnetic field component 691 varies smoothly and declines from right to left, with an inflection point around +0.8 cm. Although the magnitude of the z-component of the magnetic fields near the origin is relatively similar in FIGS. 8 and 10A (e.g., approximately 5200 Gauss and 5630 Gauss, respectively), it can be seen that the magnetic field gradient n in the z-direction i.e., the slope of the z-direction magnetic field component, is significantly greater for the inner bending magnets and of opposite sign, as shown in FIG. 10A, as compared to that for the outer bending magnets, as shown in FIG. 8. In some embodiments, the pole faces are shaped so as to introduce higher-order magnetic field components so as to control geometric and chromatic aberrations in the beam which may be produced by the simple linear field gradients heretofore described. That is, system 210 illustrated in FIGS. 2A-2B also may be achromatic to second order, or to even higher orders. Such second order achromaticity may be useful, in facilitating substantially uniform bending of particle beams in which the particles have energies that vary about a central energy value, e.g., by about 10% above and below a central energy value, or even by about 15% above and below a central energy value. FIG. 10B is a plot of the x-direction magnetic field component 1091 and the z-direction magnetic field component 1092 versus transverse position near the origin (0,0), e.g., near beam path 20, for a model similar to that illustrated in FIG. 9 but configured so as to introduce second-order terms into the z-direction magnetic field component 1092. The z-direction magnetic field component 692 from FIG. 10A, which lacks such second-order terms, is also shown for comparison. In the example illustrated in FIG. 10B, the particle beam has an energy spread of plus or minus 10% about a central energy value 1093, the lower end of this spread designated 1093′ and the upper end of this spread designated 1093″. Absent the second-order terms included in z-direction magnetic field component 1092, particles having energies at the lower end 1093′ of the energy spread may deviate from beam path 20 as a result of overfocusing by the inner magnets 202, 203, caused by field strength greater than that needed to bend those particles by the same amount as are particles closer to the central energy value 1093. Similarly, particles having energies at the upper end 1093″ of the energy spread may deviate from beam path 20 as a result of underfocusing by the inner magnets 202, 203, caused by too low a field strength to bend those particles by the same amount as are particles closer to the central energy value 1093. In some embodiments, the second-order curve 1092 illustrated in FIG. 10B compensates for these bending errors by configuring inner bending magnets 202, 203 so as to generate a z-direction magnetic field component that quadratically curves upwards on either side of the central energy value 1093. Specifically, particles having energies at the lower end 1093′ of the energy spread experience a magnetic field that is decreased from ΔB1 to 0.9ΔB1, while particles having energies at the upper end 1093″ of the energy spread experience a magnetic field that is increased from ΔB2 to 1.1ΔB2. Note that such a curved profile does not distort the image of a monochromatic full emittance beam such as illustrated in FIG. 3D, because the beam crosses over between inner magnets 202, 203, and as a result electrons that are overfocused at magnet 202 may be underfocused at magnet 203, and electrons that are underfocused at magnet 202 may be overfocused at magnet 203, resulting in cancellation of the effect. Second order magnetic field terms such as illustrated in FIG. 10B may be achieved, for example, by adding an appropriate second order term to the gradients generated by the inner (second and third) bending magnets, for example by finely adjusting the shape of pole face 674 so as to generate second order field gradients. In one exemplary embodiment, fourth core portion 675 of inner bending magnets 202, 203, illustrated in FIG. 6, is removed or changed in position relative to that shown so as to generate higher order achromaticity. The particular shape of second-order curve 1092 is merely exemplary, and any number of magnets 201-204 may be configured to introduce any suitable higher-order terms into the magnetic fields to which particle beam 20 is exposed. In one illustrative embodiment, the z-component of the magnetic field generated by the inner magnets 202, 203 has a substantially quadratic profile over a location corresponding to the energy spread of the particles, e.g., over a 20% full width energy range, or even over a 30% full width energy range. Note that although gradient magnets having shaped pole faces 574, 674 are illustrated in FIGS. 5-6, other types of magnets may be used to achieve a similar effect, e.g., to each achromatically bend charged particles by a desired angle, in one embodiment by about 22.5°, resulting in a net bend angle of about 90°. For example, quadrupole magnets may be appropriately configured to provide comparable magnetic fields and field gradients to those shown in FIGS. 7-10, and performance comparable to that shown in FIGS. 3A-3D. In one illustrative embodiment, magnets 201-204 illustrated in FIG. 2B are rectangular, laminated electromagnets having pole faces respectively shaped as illustrated in FIGS. 5-6. As is known in the art, the core of an electromagnet may be laminated, e.g., using thin sheets of iron, to impede the circulation of induced currents that would otherwise resist rapid changes in magnetic field. As such, the use of laminated cores in magnets 201-204 may allow the fields generated by those magnets to be changed relatively quickly, for example on the millisecond timescale, for example to bend particle beams of different energies in quick succession. Alternatively, magnets 201-204 are rectangular, but are not laminated. In an alternative embodiment, illustrated in FIG. 11, non-rectangular first, second, third, and fourth magnets 1101-1104 may be used in alternative beam bending system 1100. These magnets may be laminated, as described above, or alternatively may be non-laminated. As illustrated in FIG. 11, outer (first and fourth) magnets are shaped as parallelograms, and inner (second and third) magnets are shaped as trapezoids. Such shaped magnets may be referred to as “wedge” magnets. As noted above, the systems and methods of the present invention are also compatible with radiation treatment systems other than the robot-based system 100 illustrated in FIG. 1. For example, FIG. 12 schematically illustrates a gantry-based system 500. System 500 includes patient positioning system 307 and gantry-based radiation system 501. Gantry-based radiation system 501 includes a gantry 502, a LINAC 503, and a portal imaging device 504. LINAC 503 is arranged substantially horizontally, and is coupled to a 90° achromatic bending system such as described herein (not illustrated in FIG. 12). Gantry 502 is configured to move LINAC 503 in a fixed plane about the patient 310. LINAC 503 may include a multi-leaf collimator. Patient positioning system 307 may be a robotic system for moving patient 310 relative to the gantry 502, as shown, or any other suitable patient support system as known to the skilled artisan. Gantry-based radiation system 510 and patient positioning system 307 are in operable communication with a controller (not shown) configured for operation with the particular radiation system and patient positioning system being used. While various illustrative embodiments of the invention are described above, it will be apparent to one skilled in the art that various changes and modifications may be made therein without departing from the invention. For example, although approximately 90° bends have predominantly been described, bends of other angles also may be made by suitably modifying the systems and methods described herein. The appended claims are intended to cover all such changes and modifications that fall within the true spirit and scope of the invention.
	summary
053533232	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS According to an embodiment shown in FIG. 1, an X-ray window (beryllium foil) 2 is provided between a vacuum chamber 1 where X-rays emitted from an X-ray source XS enter, and a chamber 3. The chamber 3 is filled with high-purity gaseous helium to provide an X-ray low attenuation atmosphere. An X-ray mask 4 having a pattern of an LSI circuit or the like drawn thereon is disposed between the chamber 3 and the air, facing the X-ray window 2. A wafer 5 for example is placed in the air, opposite to the X-ray mask 4. Provided above the chamber 3 is a gas supplying portion 6 for guiding gaseous helium inside the chamber 3. Provided below the bottom portion of the chamber 3 is a gas discharging portion 7 for discharging gaseous helium out of the chamber 3. The gas supplying portion 6 has a pressure control chamber 8 having a relatively large volume and a restrictor 6c, which is provided in a pipe 6b for connecting this pressure control chamber 8 to the chamber 3. The restrictor 6c has an orifice 6a having a smaller diameter a than that of the pipe 6b. Provided above the pressure control chamber 8 is a differential pressure gauge 9 which measures the difference between pressure in the pressure control chamber 8 and atmospheric pressure. A helium tank 14 is connected to the side of the pressure control chamber 8 via a flow-rate control valve 10, which may be an electromagnetic valve. The flow-rate control valve 10 is connected to a pressure controller 11. The pressure controller 11 is connected to the differential pressure gauge 9, and controls the flow-rate control valve 10 based on the output signal of the differential pressure gauge 9 to adjust the flow rate of the gaseous helium to be supplied to the pressure control chamber 8. The gas discharging portion 7, connected to the chamber 3, has a restrictor 7a having a larger diameter b than the diameter a of the orifice 6a (i.e., a<b), an oxygen analyzer 12, a shield valve 13, a lead pipe 7c extending downward of the restrictor 7a, and an opening (outlet port) 7b open to the air. The shield valve 13 and oxygen analyzer 12 are located upstream of the restrictor 7a. The operation of the thus structured X-ray exposure apparatus will be described below. The X-ray source XS is activated to emit X-rays in the vacuum chamber 1. The X-rays pass through the X-ray window 2, the X-ray low attenuation atmosphere of the chamber 3 and the mask 4 to transfer the LSI pattern of the mask 4 on the wafer 5. At this time, the differential pressure gauge 9 measures the difference between the pressure in the pressure control chamber 8 and atmospheric pressure and inputs the pressure difference signal to the pressure controller 11. The pressure controller 11 controls the opening of the flow-rate control valve 10 in accordance with the pressure difference signal. Accordingly, some amount Q of gaseous helium is supplied via the flow-rate control valve 10 to the pressure control chamber 8 from the helium tank 14. In this case, if the pressure in the pressure control chamber 8 is controlled to be P.sub.0, the pressure P.sub.1 in the chamber 3 will be expressed by the following equation. EQU P.sub.1 .apprxeq.P.sub.0 (a/b).sup.4 where a is the diameter of the orifice 6a and b is the diameter of the restrictor 7a. Thus, if P.sub.0 =1 mmHg, a=1 mm and b=10 mm, it is apparent from the above equation that the minute pressure difference of 1/10000 mmHg (10.sup.-4 mmHg) is attained between the pressure P.sub.1 in the chamber 3 and the atmospheric pressure. With a typical X-ray mask (membrane thickness of 1 .mu.m and a size of 25 mm on each side), if the pressure difference between the pressure P.sub.1 in the chamber 3 and the atmospheric pressure is set to about 1/1000 mmHg (10.sup.-3 mmHg), the amount of deformation of the mask will be suppressed to about 0.5 .mu.m. Even with the use of an X-ray mask that is easy to deform, the amount of deformation will be suppressed to about 1 to 2 .mu.m for that pressure difference. If the pressure P.sub.1 in the chamber 3 is set slightly higher than the atmospheric pressure by about 1/10000 mmHg, therefore, the amount of deformation of the X-ray mask 4 can be suppressed significantly, thus ensuring adequate exposure at a sufficiently high accuracy even in the case of close exposure with the gap of about 30 .mu.m set between the mask 4 and the wafer 5. The place in the X-ray exposure apparatus where one wants to measure the oxygen concentration is inside the chamber 3 where the purity of helium should be concerned. If the oxygen analyzer 12 is provided at the gas discharging portion 7 for gaseous helium as in the above-described embodiment, the oxygen concentration in the chamber 3 would always be monitored without directly measuring that oxygen concentration and no pump is needed to suck sample gas. In other words, as the chamber 3 has no portion open to the air except the gas discharging portion 7, it is assured that the oxygen concentration in the chamber 3 located upstream of the gas discharging portion 7 will be equal to or lower than the oxygen concentration at the gas discharging portion 7, if the oxygen concentration is measured at the outlet port of the gas discharging portion 7. The most probable cause for an increase in oxygen concentration in the chamber 3 is the diffusion of oxygen near the gas discharging portion 7, which is prevented by the lead pipe 7c located between the restrictor 7a and the opening 7b in the above embodiment. when no gaseous helium flows when the apparatus is activated or deactivated, the shield valve 13 can shield the inside of the chamber 3. Although the pressure control chamber 8 is provided in the above embodiment, it may be replaced with a pipe 16 as shown in FIG. 2. Although the pressure control chamber 8 needs a certain volume to provide uniform pressure, it may be replaced with a simple pipe 16. In this case, when the diameter of the pipe 16 is larger than the diameter a of the orifice 6a, it is necessary to form a restrictor corresponding to the orifice 6a as shown in FIG. 2. If the diameter a of the orifice 6a is equal to that of the pipe 16, the pipe 16 alone will accomplish the same function as the orifice 6a in the embodiment in FIG. 1. This eliminates the need for separately providing the orifice 6a to the pipe 16, thus simplifying the structure. Another embodiment will now be described referring to FIG. 3. While the X-ray source XS, vacuum chamber 1 and X-ray window (beryllium foil) 2 are not shown in FIG. 3 for the sake of convenience, the structures and arrangement of those members are the same as those of the first embodiment. In the diagram, as per the previous embodiment, the chamber 3 is filled with highly pure gaseous helium to provide an X-ray low attenuation atmosphere. An X-ray mask 4 having a pattern of an LSI circuit or the like drawn thereon is disposed between the chamber 3 and the air, facing the X-ray window 2. A wafer 5 for example is placed in the air, opposite to the X-ray mask 4. Provided above the chamber 3 is a gas supplying portion 6 for guiding gaseous helium inside the chamber 3. Provided at the side of the bottom portion of the chamber 3 is a gas discharging portion 7 for discharging gaseous helium out of the chamber 3. The gas supplying portion 6 has a pressure control chamber 8 having a relatively large volume and a pipe 6b, which connects this pressure control chamber 8 to the chamber 3. Provided above the pressure control chamber 8 is a differential pressure gauge 9 which measures the difference between pressure in the pressure control chamber 8 and atmospheric pressure. A helium tank 14 is connected to the side of the pressure control chamber 8 via a flow-rate control valve 10, which may be an electromagnetic valve. The flow-rate control valve 10 is connected to a pressure controller 11. The pressure controller 11 is connected to the differential pressure gauge 9, and controls the flow-rate control valve 10 based on the output signal of the differential pressure gauge 9 to adjust the flow rate of the gaseous helium to be supplied to the pressure control chamber 8. The gas discharging portion 7, connected to the side of the bottom portion of the chamber 3, has a pipe 7c extending upward from the bottom portion. This pipe 7a has an upward opening (outlet port) 7b formed at the same height as the height h from the bottom of the chamber 3 to nearly the center of the mask 4. A shield valve 13 is provided where the pipe 7c is connected to the chamber 3. An oxygen analyzer 12 is attached to the pipe 7 near the opening 7b. In the X-ray exposure apparatus having the above structure, if the atmospheric pressure is P.sub.0, the pressure in the chamber 3 is P.sub.1 and the difference between the height of the mask 4 and that of the opening 7b is .DELTA.h (cm), the pressure difference that occurs due to the height difference will be given by the following equation. EQU P.sub.1 -P.sub.0 =(.gamma..sub.air -.gamma.He).multidot..DELTA.h where .gamma..sub.air air is the specific weight of air and .gamma.He is the specific weight of helium. For example, with the height difference .DELTA.h=10 cm and EQU .gamma..sub.air =1.293.times.10.sup.-6 (Kg/cm.sup.3) EQU .gamma.He=0.179.times.10.sup.-6 (Kg/cm.sup.3), then a pressure difference of ##EQU1## would be produced. In view of the amount of deformation of the mask 4, it is desirable that the opening 7b be located at the same height as the height (h) from the bottom of the chamber 3 to nearly the center of the mask 4. The allowance of the height difference .DELTA.h should be properly set in accordance with the type of the mask based on the above equations. In the thus constituted X-ray exposure apparatus, a predetermined amount of gaseous helium is supplied to the chamber 3 via the valve 10, pressure control chamber 8 and pipe 6b from the helium tank 14. At this time, the differential pressure gauge 9 measures the difference between the pressure in the pressure control chamber 8 and atmospheric pressure and inputs the pressure difference signal to the pressure controller 11. The pressure controller 11 controls the opening of the flow-rate control valve 10 in accordance with the pressure difference signal. Accordingly, gaseous helium whose quantity corresponds to the pressure difference is supplied via the flow-rate control valve 10 to the pressure control chamber 8 from the helium tank 14. The gaseous helium supplied to the chamber 3 is properly discharged through the gas discharging portion 7. In this case, since the height from the bottom of the chamber 3 to the opening 7b is about the same as the height from the bottom of the chamber 3 to the center of the mask, the pressure in the chamber 3 becomes approximately equal to the atmospheric pressure as apparent from the above-given equation. Accordingly, the X-ray mask 4 will hardly deform, thus ensuring adequate exposure at a sufficiently high accuracy even in the case of close exposure with the gap of about 30 .mu.m set between the mask 4 and the wafer 5. FIG. 4 shows an embodiment which has a restrictor 6c with an orifice 6a further provided to the pipe 6b of the embodiment shown in FIG. 3. The provision of the restrictor 6c having the orifice 6a of a small diameter as in this embodiment will allow the pressure in the chamber 3 to vary only slightly with respect to a change in gas pressure in the pressure control chamber 8. Therefore, a significant pressure change will not occur on the mask 4, thereby preventing the mask 4 from being deformed significantly or being damaged. When the pressure control chamber 8 and the restrictor 6c are provided as in the above embodiment, it is sufficient that the difference between the pressure in the pressure control chamber and the atmospheric pressure be controlled at a relatively coarse accuracy on the order of about 1 mmHg as mentioned earlier. Therefore, the differential pressure gauge 9 should not necessarily be a high-precision differential pressure gauge. In other words, the pressure in the chamber 3 can be controlled at a high accuracy greater by a factor of several tens over the prior art as long as the flow-rate control valve 10 and the pressure controller 11 have about the same precision as those of the prior art. FIG. 5 shows an embodiment which has both the differential pressure gauge 9 and pressure controller 11 removed from the embodiment in FIG. 4. The flow-rate control valve 10 is manually controlled in this embodiment. Even if the pressure in that chamber 8 is manually adjusted, the restrictor 6c will prevent a large pressure change in the pressure control chamber 8 from being directly transmitted to the chamber 3, and the pressure change would be suppressed by about a factor of 1000 of the actual pressure change in the pressure control chamber 8. This embodiment will therefore sufficiently serve for practical usage, without automatic pressure control involving the differential pressure gauge 9 and pressure controller 11. Since the automatic pressure control mechanism is not employed in this embodiment, the structure of the exposure apparatus will be simplified considerably. A different embodiment will now be described referring to FIG. 6. The same reference numerals as used for the embodiment in FIG. 4 will be given to the identical components in this embodiment to avoid repeating their description. In this embodiment, a pipe 7e extends upward from the side of the bottom portion of the chamber 3 and then extends horizontally. The shield valve 13 and oxygen analyzer 12 are provided at the horizontal portion of this pipe 7e. An outer pipe 7f is provided over a lead pipe 7c that extends downward from the distal end of the pipe 7e. The outer pipe 7f is fitted over the lead pipe 7c in such a manner that the distance between the outlet port of the outer pipe 7f and the bottom of the chamber 3, or the height h from the bottom of the chamber 3 to the outlet port of the outer pipe 7f, equals the height h from the bottom of the chamber 3 to the center position of the mask 4, with the downward opening 7b of the lead pipe 7c located slightly inward from the outlet port of the outer pipe 7f. A helium retainer 7d for surely preventing air diffusion or penetration from the gas discharging portion is therefore formed between the outer pipe 7f and the lead pipe 7c, i.e., around the lead pipe 7c. With the above structure, gaseous helium coming from the opening 7b rises since it has lighter specific weight than air, and stays in the helium retainer 7d. The gaseous helium retained around the opening 7b will prevent air (oxygen) from entering the chamber 3 through the opening 7b due to diffusion. As the pipe 7e is bent upward so that a part of the pipe passage of the gas discharging portion 7 is located higher than the opening 7b in consideration of gaseous helium having a smaller density than air, air having a larger specific weight than the gaseous helium, if some enters through the opening 7b, would stay somewhere inside the pipe 7e and would not enter the chamber 3. As mentioned earlier, the height difference of 1 cm between the outlet port of the gas discharging portion 7 and the center of the X-ray mask 4 generates a pressure difference of about 1/1000 mmHg. Since the height from the bottom of the chamber 3 to the center of the X-ray mask 4 is set nearly equal to the height from the bottom of the chamber 3 to the outlet port of the gas discharging portion 7 or the outlet port of the outer pipe 7f, the difference between the atmospheric pressure and the pressure in the chamber 3 will be kept very low. This prevent the mask 4 from being deformed and damaged. An embodiment shown in FIG. 7 is the embodiment in FIG. 6 to which a pipe 15 for connecting the helium tank 14 to the helium retainer 7d is added. In other words, the helium tank 14 and the outer pipe 7f are coupled together by the pipe 15, leading gaseous helium in the helium tank 14 to the helium retainer 7d. This structure allows a large amount of gaseous helium to always stay around the opening 7b to purge around the opening 7b with the gaseous helium. It is therefore possible to more surely prevent air (oxygen) from entering the chamber 3 due to diffusion. An embodiment shown in FIG. 8 has a shield valve 17 provided at the gas supplying portion 6 in the embodiment in FIG. 6. More specifically, the shield valve 17 is attached, adjacent to the orifice 6a, to the restrictor 6 provided in the pipe 6b of the gas supplying portion 6, and a bypass passage 18 including the shield valve 17 is provided in the pipe 6b. With this structure, to fill inside the chamber 3 with gaseous helium at the time the exposure apparatus is activated, the shield valve 17 is opened to supply a large amount of gaseous helium via the bypass passage 18 to the chamber 3 from the helium tank 14, filling inside the chamber 3 with the gaseous helium in a short period of time. A further embodiment will be described below with reference to FIG. 9. In this embodiment, the differential pressure gauge 9 is directly coupled to the chamber 3 so as to directly measure the pressure in the chamber 3. In this case, the differential pressure gauge 9 in use should be a relatively high-precision type which can measure the difference between the pressure in the chamber 3 and the atmospheric pressure on the order of a predetermined pressure difference of 10.sup.-3 mmHg. Further, the helium tank 14 is coupled via the flow-rate control valve 10 to the chamber 3 by the pipe 16 which constitutes the gas supplying portion 6. In this embodiment too, the pipe 7e of the gas discharging portion 7 extends upward from the side of the bottom portion of the chamber 3 and then extends horizontally. The distal end of the pipe 7e is bent downward to form the downward opening 7b. In this case, the height from the bottom of the chamber 3 to the center position of the mask 4 is set equal to the height from the bottom of the chamber 3 to the opening 7b of the gas discharging portion 7. Further, the shield valve 13 and oxygen analyzer 12 are attached to the horizontal portion of this pipe 7e. Although it has been just mentioned that the height h from the bottom of the chamber to the opening 7b of the gas discharging portion 7 is set equal to the height from the bottom of the chamber 3 to the center position of the mask 4 in this embodiment, strictly speaking, the relation among the pressures at the individual points shown in FIG. 10 is expressed by the following equations in consideration of the resistance component of the pipe. EQU P.sub.f =P.sub.a +.gamma..sub.air .multidot..DELTA. EQU p.sub.b =P.sub.a +.gamma.He.DELTA.h+.zeta.(.gamma.He/2).upsilon..sup.2 EQU P.sub.b -P.sub.f =-(.gamma..sub.air -.gamma.He).multidot..DELTA.h+.zeta.(.gamma.He/2 where .zeta.(.gamma.He/2).upsilon..sup.2 is the resistance component of the pipe. It is desirable that the height from the bottom of the chamber 3 to the opening 7b be made variable to cancel the resistance component of the pipe. FIG. 11 shows an embodiment which has this height changing function. As shown in FIG. 11, a pressure gauge 21 is attached to the side of the chamber 3 at the same height from the bottom of the chamber 3 as that therefrom to the center of the mask 4. This pressure gauge 21 may be a sensor which is attached to the side of the chamber 3 to detect the amount of deformation of the measuring mask that has the same characteristic as the mask 4. A flexible pipe 22 is connected to the distal end of the gas discharging portion 7, and the lead pipe 7c having the opening 7b is provided downward at the distal end of this flexible pipe 22. The distal end of the lead pipe 7c is coupled to a driving mechanism 23, and moves up and down by the action of the flexible pipe 22 in accordance with the movement of the shaft of the driving mechanism 23. The driving mechanism 23 is connected to a controller 24, which controls the mechanism 23 in accordance with the pressure detected by the pressure gauge 21 to automatically adjust the distance between the bottom of the chamber 3 and the opening 7b according to the detected pressure. As the height from the bottom of the chamber 3 to the opening 7b has only to be adjusted at the time of initializing the exposure apparatus, the height to the opening 7b may be fixed or unchangeable during operation. While the same reference numerals as used for the previous embodiment are given to the identical components in the embodiment of FIG. 11 to avoid repeating their description, those components have the same structures and perform the same functions as those of the previous embodiment. Although the foregoing description of the individual embodiments has been given with reference to the case where the X-ray mask is disposed vertically, this invention may be adapted for an exposure apparatus where the X-ray mask 4 is disposed horizontally, as shown in FIG. 12. More specifically, the X-ray mask 4 is attached horizontally to the bottom of the chamber 3, and the X-ray window 2 is provided in that top portion of the chamber 3 which faces this mask 4. The gas supplying portion 7, pressure control chamber 8, differential pressure gauge 9, flow-rate control valve 10, pressure controller 11 and helium tank 14 are provided at one side portion of the chamber 3 (left-hand side in the diagram) with the same connecting relation as the previous embodiments. The gas discharging portion 7 is arranged on the opposite side of the chamber 3 (right-hand side in the diagram). The distal end of the lead pipe 7c of the gas discharging portion 7 is bent downward, and the distance between the opening 7b at the distal end and a mounting table 25 or the height h from the mounting table 25 to the opening 7b is set equal to the distal between the surface of the mask 4 and the mounting table 25 or the height h from the mounting table 25 to the mask 4. In this embodiment too, since the height h to the opening 7b is set about the same as the height to the mask, the pressure in the chamber 3 becomes approximately equal to the atmospheric pressure as per the above-described embodiments. Accordingly, the X-ray mask 4 will hardly deform, so that adequate exposure will be accomplished at a sufficiently high accuracy even in the case of close exposure with the gap of about 30 .mu.m set between the mask 4 and the wafer 5. Although the gas supplying portion 6 and gas discharging portion 7 have each been explained as a single line in the foregoing description of the individual embodiments, at least one of the gas supplying portion 6 and gas discharging portion 7 may branch to a plurality of lines. In this case, the total area of the cross sections of the divided gas flow passages should be set in such a way that the total cross-sectional area of the gas discharging portion 7 is greater than that of the gas supplying portion 6 to meet the above-described conditions. The cross-sectional shapes of the gas flow passages of the gas supplying portion 6 and gas discharging portion 7 are not limited to a circle, but may take various other forms, such as a rectangle and an ellipsoid. Although the oxygen analyzer 12 is provided at the gas discharging portion 7, the oxygen analyzer 12 may be disposed inside the chamber 3 so that it can directly measure the oxygen concentration inside the chamber 3 of interest. Additional advantages and modifications will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details, and representative devices, shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents.
	summary
	summary
	abstract	A gauge is provided for measuring one or more characteristics of a construction material such as a road surface. The gauge includes a detector, a base that carries the detector, and a source housing carried by the base and defining a shield material circumferentially extending inwards. A source rod is positioned within the housing and carries a source that is translatable between a shielded position within the housing and a measuring position external of the housing. The source rod has a source shield on the top thereof and a shield material spaced-downwardly from the source such that the source is completely enclosed when contained within the base.
054616484	abstract	A supercritical water oxidation reactor includes a vessel with an interior urface, two cooling sections, a heat exchanger, an oxygenating section, a pump, and a trap. The interior surface of the vessel has a corrosion-resistant, artificial ceramic or diamond-like coating. The artificial diamond coating is thin and crystal-like in structure. The heat exchanger is located between the two cooling sections. The heat exchanger and the two cooling sections surround the exterior of the vessel. The oxygenating section comprises a porous cylindrical baffle positioned within the vessel. The porous baffle transfers oxygen, hydrogen peroxide, or other oxygenating substances to an aqueous hazardous waste introduced into the reactor.. In accordance with another aspect of the invention, the oxygenating section includes a shaft having a helical extension. The shaft has a corrosion-resistant, artificial diamond or diamond-like coating on its outer surface. The shaft rotates the helical extension to assist in removing solids from the aqueous hazardous waste.
047568712	description	SPECIFIC DESCRIPTION As can be seen from the drawing, after removal of spent nuclear fuel elements at 10 from a nuclear reactor core 11, the spent fuel is sprayed with a solution 22 or with an emulsion 13 or is immersed in either or is coated with a melt 14 containing gadolinium or some other substance with a high neutron cross section to form an absorbent coating which is dried at 15 and 16 or solidified by cooling as shown at 17. The coated spent fuel particles are then further coated with a water-repellent film 18 before they are subjected to storage at 19 for decay. After the desired storage term and before reprocessing at 20, the coating is removed at 21. Fresh nuclear fuel is shown at 22 can also be coated with a foil, this time containing the absorbing substance as shown at 23 and thus can be stored and handled at 24 without the danger of reaching criticality. Before the fresh fuel is introduced into the nuclear reactor, however, the coating is removed at 25. SPECIFIC EXAMPLES EXAMPLE 1 Spent metal sheathed nuclear fuel elements are removed from the nuclear reactor core into a tank containing cooling water in the form of a solution of gadolinium nitrate and gadolinium chloride although other water soluble gadolinium salts or water-soluble salts of other neutron-absorbing elements can also be used. The solution concentrations can be such that the neutron-absorbing substance is present in an amount between substantially 0.5% to the solubility limit of the water-soluble salt of the absorbing substance. Deposition of gadolinum on the metallic shell, effected galvanically by polarizing the metal shell negatively, i.e. as a cathode against an inert (e.g. stainless steel) anode and applying a low-voltage direct current, e.g. about 24 volts. Depending upon the activity of the nuclear fuel, coating was continued to build up thicknesses sufficient to reduce the neutron emission below any level which can sustain criticality. EXAMPLE 2 Spent nuclear fuel elements are coated by spraying or dipping in a single component or multicomponent lacquer, preferably an epoxy resin lacquer in which gadolinium oxide powder is dispersed in the lacquer. The lacquer is permitted to set or solidify. EXAMPLE 3 The fuel elements are coated by gadolinium acetyl acetonate by immersing them in a melt of this substance or in an aqueous solution thereof or by spraying them with this solution or pouring this solution over them. Upon drying of the coating, the fuel elements are found to be water repellent and incapable of assembling into a critical mass. EXAMPLE 4 Spent graphitic fuel elements are introduced into a melt of gadolinium acetyl acetonate and are stored in a solution of the melt at least until gases begin to be driven off. The gases appear to be air which is expelled by heat from the pores to promote penetration of the melt and the solution so that these graphite pores are penetrated and at least partially filled with flowable material. After drying, electron microscopy, autoradiography and neutron activation analysis in the nuclear reactor shows a penetration of the neutron-absorbing substance into the deepest levels of the graphite fuel elements. The surface of the fuel elements were then immersed in hot saturated aqueous gadolinium acetate solution and then cooled and dried. They are found to be filled with gadolinium acetate crystallites. When the graphitic fuel elements are heated to a temperature above about 750.degree. C. after this treatment, the gadolinium acetate is decomposed, leaving behind gadolinium oxide which is water insoluble and is immobile in the graphitic material even at high temperatures. Gadolinium oxide thus becomes an integrated component of the spent fuel element and resists loss therefrom even with major traumatic insults to the fuel elements during accidents of transport or storage by the effects of water and heat and even upon rupture of the fuel elements. EXAMPLE 5 Comparative tests with Example 4 were made with graphitic fuel elements treated with aqueous gadolinium acetate at room temperature and atmospheric pressure. The amount of the neutron-absorbing substance taken up by the fuel element increased with the residence time of the fuel element in the solution. EXAMPLE 6 Nuclear fuel elements are coated with a synthetic resin foil. The foil is a polyvinyl chloride and is applied in a melt which contains as neutron-absorbing substance europium oxide. The concentrations of this material in the foil range from 0.5% to 25% by weight. Foil wraps of polyester, (Mylar) containing europium oxide as well as gadolinium and calcium oxide also were used effectively. Before introduction of the fuel element into the reactor or reprocessing, the foil was removed as previously described.
	claims	1. An emergency core cooling system for a BWR plant not including external recirculation piping, comprising:an active emergency core cooling system consisting of only a first safety division and a second safety division, each of said first and second divisions including a high-pressure motor-driven core cooling system, and a low-pressure motor-driven core cooling system, which is commonly used as a residual heat removal system, a heat removal capacity of said residual heat removal system being not less than 100% of the necessary capacity to cool the core and the containment to meet a safety requirement at a design basis loss of coolant accident;a first emergency power supply equipment provided for the first safety division, the first emergency power supply equipment configured to supply electricity to the first safety division; anda second emergency power supply equipment provided for the second safety division, the second emergency power supply equipment configured to supply electricity to the second safety divisions;wherein said emergency core cooling system does not include additional safety divisions that contain an active emergency core cooling system, andwherein said emergency core cooling system meets the safety requirement at a design basis loss of coolant accident considering a single failure of any active components including said first or second emergency power supply equipment. 2. The emergency core cooling system according to claim 1, wherein the first power supply equipment is a first emergency diesel generator, and wherein the second power supply equipment is a second emergency diesel generator. 3. The emergency core cooling system according to claim 1, wherein the first power supply equipment is an emergency diesel generator, and wherein the second power supply equipment is an emergency gas turbine generator. 4. The emergency core cooling system according to claim 1, wherein the first power supply equipment is a first emergency gas turbine generator, and wherein the second power supply equipment is a second emergency gas turbine generator. 5. The emergency core cooling system according to claim 1, further comprising a passive cooling system comprising a, third safety division configured to physically separate said passive safety system from the active emergency core cooling system in the first and second safety divisions, said passive safety system including at least an isolation condenser to condense reactor steam and return condensate to the reactor in a high-pressure condition at a station blackout, wherein said active emergency core cooling system is configured to cool the core with said passive cooling system not operating. 6. A BWR plant comprising an emergency core cooling system comprising:an active emergency core cooling system consisting of only a first safety division and a second safety division, each of said first and second divisions including a high-pressure motor-driven core cooling system, and a low-pressure motor-driven core cooling system, which is commonly used as a residual heat removal system, a heat removal capacity of said residual heat removal system being not less than 100% of the necessary capacity to cool the core and the containment to meet a safety requirement at a design basis loss of coolant accident;a first emergency power supply equipment provided for the first safety division, the first emergency power supply equipment configured to supply electricity to the first safety division; anda second emergency power supply equipment provided for the second safety division, the second emergency pourer supply equipment configured to supply electricity to the second safety division;wherein said emergency core cooling system does not include additional safety divisions that contain an active emergency core cooling system, and said BWR plant does not include an external recirculation piping, andwherein said emergency core cooling system meets the safety requirement at a design basis loss of coolant accident considering a single failure of any active components including said first or second emergency power supply equipment. 7. The BWR plant according to claim 6, wherein the first power supply equipment is a first emergency diesel generator, and wherein the second power supply equipment is a second emergency diesel generator. 8. The BWR plant according to claim 6, wherein the first power supply equipment is an emergency diesel generator, and wherein the second power supply equipment is an emergency gas turbine generator. 9. The BWR plant according to claim 6, wherein the first power supply equipment is a first emergency gas turbine generator, and wherein the second power supply equipment is a second emergency gas turbine generator. 10. The BWR plant according to claim 6, further comprising a passive cooling system comprising a third safety division configured to physically separate said passive safety system form the active emergency core cooling system in the first and second safety divisions, said passive safety system including at least an isolation condenser to condense reactor steam and return condensate to the reactor in a high-pressure condition at a station blackout, wherein said active emergency core cooling system is configured to cool the core with said passive cooling system not operating.
	summary
	description	This application claims the benefit of U.S. Provisional Patent Application No. 61/663,102 entitled “Method and Apparatus for Electron Pattern Imaging” filed Jun. 22, 2012, which is hereby incorporated by reference in its entirety. 1. Field of the Invention The present invention is generally directed to a method and apparatus for electron pattern imaging and, more specifically, to the use of an image intensifier to amplify light in an electron backscatter diffraction (EBSD) pattern collection system in a scanning electron microscope (SEM) environment. 2. Description of Related Art In the field of material microanalysis, EBSD is used to study the microcrystalline characteristics of materials. The method uses the electron beam of an SEM incident on a sample and an electron imaging system to collect the electron diffraction pattern. With reference to FIG. 1, an example of a current arrangement for conducting EBSD is illustrated. The current arrangement includes a phosphor screen 1, lens (not shown), and a camera 3. The EBSD pattern is produced by illuminating electron beams 5 onto a sample 7 using the SEM 9. The EBSD pattern which falls on the phosphor screen 1 is converted to a corresponding light pattern which is focused by the lens system onto the camera 3. An EBSD map of the sample is produced by rastering the incident beam through an array of points on the sample 7 and collecting corresponding EBSD patterns at each point. EBSD sensitivity and collection speed is determined by at least one of the following: (a) incident electron beam accelerating voltage and current; (b) diffraction characteristics of the sample; (c) phosphor efficiency; (d) lens transmission; and (e) sensitivity and signal-to-noise ratio of the camera. However, this current arrangement suffers from various deficiencies as discussed hereinafter. For example, SEMs typically operate at up to 30 keV accelerating voltages. However, it is more desirable to operate at much lower voltages and beam currents to minimize the interaction volume within the sample, thereby optimizing spatial resolution and minimizing damage and contamination of the specimen due to the high energy electron interaction. Under low voltage and low current conditions, most SEMs do not provide sufficient beam current for optimal illumination of the phosphor screen. Often accelerating voltage on the sample needs to be increased at the expense of spatial resolution and specimen degradation in order to obtain acceptable EBSD patterns. Diffraction characteristics of the sample also create limitations in the current arrangement. The intensity of the EBSD pattern produced by the sample depends upon several factors: composition, degree of crystallinity, homogeneity, surface preparation, etc. These conditions diffuse and/or weaken the diffraction pattern incident on the phosphor screen. The efficiency of the phosphor also plays an important role in the production of the EBSD pattern. Electrons interact with the phosphor material to produce light. The interaction volume is a function of the electron energy, phosphor composition, and thickness. The ideal phosphor needs to be very thin to minimize electron interaction volume, but sufficiently absorbing to convert all of the pattern electrons to light. An additional criteria is that the wavelength of light produced by the phosphor match the sensitive range of the camera. In addition, lens clarity and F-number determine the amount of light transmitted from the phosphor to the camera. Geometric constraints limit the maximum size of lenses that can be used. Finally, the sensitivity and signal-to-noise ratio of the camera also play a role in determining the quality of EBSD pattern obtained by the current system. Within the given constraints of sample characteristics and current technology (SEM current, phosphor, and lens efficiency) exposure time to achieve an acceptable EBSD pattern is fundamentally limited by sensitivity and S:N of the camera. This determines the minimum limit of pattern detection and maximum pattern collection speed. Accordingly, a need exists for a method and apparatus for EBSD pattern collection that amplifies light passing from the phosphor screen to the camera without degrading the signal-to-noise ratio. An object of the present invention is to amplify light in an EBSD pattern collection system using an image intensifier in conjunction with a device for safely operating the image intensifier in an electron microscope environment. Accordingly, provided is a system for electron pattern imaging that includes: a device for converting electron patterns into visible light provided to receive an electron backscatter diffraction (EBSD) pattern from a sample and convert the EBSD pattern to a corresponding light pattern; a first optical system positioned downstream from the device for converting electron patterns into visible light for focusing the light pattern produced by the device for converting electron patterns into visible light; a camera positioned downstream from the first optical system for obtaining an image of the light pattern; an image intensifier positioned between the device for converting electron patterns into visible light and the camera for amplifying the light pattern produced by the device for converting electron patterns into visible light; and a device positioned within the system for protecting the image intensifier from harmful light. The device is selected from the group consisting of: a near infrared (NIR) cut-off filter positioned upstream from the image intensifier for preventing NIR light from reaching the image intensifier, a light sensitive sensor in communication with a controller for detecting when a predetermined level of harmful light has been reached, and a short wavelength photo-cathode material positioned within the system to prevent excitation by NIR light. The image intensifier may be positioned at any one of the following locations: (1) between the device for converting electron patterns into visible light and the first optical system(2) within the first optical system; or (3) between the first optical system and the camera. If the image intensifier is positioned between the first optical system and the camera, then a second optical system may be positioned between the image intensifier and the camera. The second optical system may include a focusing lens, a relay lens, a fiber optic relay, or any combination thereof. The system may further include an alignment mechanism for aligning at least one of the camera and the image intensifier with the device for converting electron patterns into visible light. Also provided is a method of electron pattern imaging. The method includes: positioning a device for converting electron patterns into visible light adjacent to a sample; positioning a first optical system downstream from the device for converting electron patterns into visible light; positioning a camera downstream from the first optical system; positioning an image intensifier between the device for converting electron patterns into visible light and the camera; positioning a device within the system for protecting the image intensifier from harmful light; receiving an electron backscatter diffraction (EBSD) pattern from a sample with the device for converting electron patterns into visible light such that the device for converting electron patterns into visible light converts the EBSD pattern to a corresponding light pattern; focusing the light pattern produced by the device for converting electron patterns into visible light with the first optical system; amplifying the light pattern produced by the device for converting electron patterns into visible light with the image intensifier; and obtaining an image of the light pattern with the camera. The device for protecting the image intensifier from harmful light is selected from the group consisting of: a near infrared (NIR) cut-off filter positioned upstream from the image intensifier for preventing NIR light from reaching the image intensifier, a light sensitive sensor in communication with a controller for detecting when a predetermined level of harmful light has been reached, and a short wavelength photo-cathode material positioned within the system to prevent excitation by NIR light. In addition, provided is a system for electron pattern imaging that includes: a device for converting electron patterns into visible light provided to receive an electron backscatter diffraction (EBSD) pattern from a sample and convert the EBSD pattern to a corresponding light pattern; a camera positioned downstream from the device for converting electron patterns into visible light for obtaining an image of the light pattern; an image intensifier positioned between the device for converting electron patterns into visible light and the camera for amplifying the light pattern produced by the device for converting electron patterns into visible light; and a device positioned within the system for protecting the image intensifier from near infrared (NIR) light. These and other features and characteristics of the present invention, as well as the methods of operation and functions of the related elements of structures and the combination of parts and economies of manufacture, will become more apparent upon consideration of the following description and the appended claims with reference to the accompanying drawings, all of which form a part of this specification, wherein like reference numerals designate corresponding parts in the various figures. As used in the specification and the claims, the singular form of “a”, “an”, and “the” include plural referents unless the context clearly dictates otherwise. For purposes of the description hereinafter, the terms “upper”, “lower”, “right”, “left”, “vertical”, “horizontal”, “top”, “bottom”, “lateral”, “longitudinal”, and derivatives thereof shall relate to the invention as it is oriented in the drawing figures. However, it is to be understood that the invention may assume various alternative variations, except where expressly specified to the contrary. It is also to be understood that the specific devices illustrated in the attached drawings, and described in the following specification, are simply exemplary embodiments of the invention. Hence, specific dimensions and other physical characteristics related to the embodiments disclosed herein are not to be considered as limiting. With reference to FIG. 2, a system for electron pattern imaging, denoted generally as reference numeral 10, includes: a device for converting electron patterns into visible light, such as a phosphor screen 12, provided to receive an electron backscatter diffraction (EBSD) pattern from a sample (not shown) and convert the EBSD pattern to a corresponding light pattern. The system 10 further includes a first optical system 14 positioned downstream from the phosphor screen 12 for focusing the light pattern produced by the phosphor screen 12. The first optical system 14 may be positioned within a chamber 16 having a vacuum formed therein to receive the light pattern through a vacuum window 18. While the phosphor screen 12 is described hereinabove and hereinafter as an example of a device for converting electron patterns into visible light, this is not to be construed as limiting the present invention as YAG crystals or any other suitable device may be utilized to convert electron patterns into visible light. The system 10 further includes a camera 20 positioned downstream from the first optical system 14 for obtaining an image of the light pattern. The camera 20 may include any suitable image sensor, such as a CCD 22. An image intensifier 24 is positioned between the phosphor screen 12 and the camera 20 for amplifying the light pattern produced by the phosphor screen 12. Desirably, the image intensifier 24 is positioned between the first optical system 14 and the camera 20 as shown in FIG. 2. However, this is not to be construed as limiting the present invention as the image intensifier 24 may also be positioned (1) between the phosphor screen and the first optical system or (2) within the first optical system. If the image intensifier 24 is positioned between the first optical system 14 and the camera 20, then a second optical system (not shown) may be positioned between the image intensifier 24 and the camera 20. The second optical system may include a focusing lens, a relay lens, a fiber optic relay, or any combination thereof. The image intensifier 24 may be a PROXIFIER® image intensifier manufactured by PROXITRONIC Detector Systems GmbH. With reference to FIG. 3, such an image intensifier 24 includes a photo-cathode 26 and a fluorescent screen 28. In operation, light impinges upon the photo-cathode 26 through an input window of the image intensifier 24. Due to the photoelectric effect, electrons are produced which escape from the photo-cathode 26 with very little energy. By a high potential electrical acceleration field between the photo-cathode 26 and the fluorescent screen 28 of 10 kV to 15 kV, the electrons are strongly accelerated and, at the same time, closely focused. They strike the fluorescent screen 28 with high kinetic energy and stimulate fluorescence. The use of such an image intensifier is for exemplary purposes only and is not to be construed as limiting the present invention as any suitable image intensifier may be utilized. The system 10 may further include an alignment mechanism (not shown) for aligning the camera 20 and/or the image intensifier 24 with the phosphor screen 12. The system 10 of the present invention further contemplates the use of a device or method to protect the image intensifier 24 from harmful light, thereby allowing the system 10 to safely and cost-effectively utilize such an image intensifier 24. With reference to FIG. 4A, the device may be a near infrared (NIR) cut-off filter positioned upstream from the image intensifier 24 for preventing NIR light from reaching the image intensifier 24. The near infrared (NIR) cut-off filter may be provided as a filter coating 40 provided on the vacuum window 18. Alternatively, the near infrared (NIR) cut-off filter may be a filter 40′ positioned behind the vacuum window 18 (shown in phantom in FIG. 4A). Still further, the near infrared (NIR) cut-off filter may be a filter 40″ positioned in front of the image intensifier 24. With reference to FIG. 4B, the device for protecting the image intensifier 24 from harmful light may also be embodied as a light sensitive sensor 42 in communication with a controller 44 for detecting when a predetermined level of harmful light has been reached. The controller 44 may be configured to shut down the image intensifier 24 when the predetermined level of harmful light is detected by the light sensitive sensor 42. Still further, and with reference to FIG. 4C, the device for protecting the image intensifier 24 from harmful light may be configured as a short wavelength photo-cathode material 46 positioned within the system 10 to prevent excitation by near infrared (NIR) light. Desirably, the photo-cathode material 46 is an internal part of the image intensifier 24 as shown in FIG. 4C. As such, the photo-cathode material 46 may be integral with the photo-cathode 26 of the image intensifier 24 or it may be a separate component. Additionally, other alternatives for protecting the image intensifier 24 are as follows. The controller may be configured to depower the image intensifier 24 when a level of harmful light is detected by the image intensifier 24 by monitoring an electronic signal from the image intensifier 24. Alternatively, the controller may be configured to enable power to the image intensifier 24 only when a level of chamber vacuum indicates a closed chamber (i.e., no external light). A final alternative is to configure the controller to enable power to the image intensifier 24 only when the software provided thereon detects a safe operating condition such as appropriate microscope conditions software running properly and other system safety checks. The system 10 described hereinabove enables an intensified CCD array to be used for EBSD. In order for such an intensified CCD array to be utilized in such a manner, the controller must be configured to allow for automatic intensified gain control based upon feedback from the CCD 22. In addition, the controller must also be configured to provide gain/contrast control based on analysis intensity and contrast in the EBSD images. Accordingly, the system 10 is capable of obtaining high quality EBSD patterns with high spatial resolution of the sample by using the intensified CCD array even at low electron accelerating beam voltages. The use of the image intensifier 24 between the phosphor screen 12 and the camera 20 also improves low level signal EBSD pattern detection. This overcomes the limitations of CCD light sensitivity and enables detection of weak EBSD patterns, such as those provided in poorly formed crystals, inhomogeneity, poor surface preparation, etc. By positioning the image intensifier 24 in the system as described hereinabove, the speed of EBSD pattern recognition is improved, thereby shortening the time required to achieve an EBSD pattern suitable for indexing. Finally, by simultaneously controlling incident beam voltage and the gain of the image intensifier 24 to achieve optimal image contrast in the CCD 22, the EBSD sensitivity is improved. Although the invention has been described in detail for the purpose of illustration based on what is currently considered to be the most practical and preferred embodiments, it is to be understood that such detail is solely for that purpose and that the invention is not limited to the disclosed embodiments, but, on the contrary, is intended to cover modifications and equivalent arrangements. Furthermore, it is to be understood that the present invention contemplates that, to the extent possible, one or more features of any embodiment can be combined with one or more features of any other embodiment.

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