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	abstract	A central dryer for electron beam curing is described which includes a first application unit for the application of a first coating to a web. The central dryer for electron beam curing also includes an irradiation unit in which a first electron beam generator and a second electron beam generator are arranged for the irradiation of the web. The electron beam generators arranged in the irradiation unit have connections for at least one pump system to generate an operating vacuum. In addition to this the central dryer for electron beam curing also has a web guiding system which feeds the web successively but not necessarily in direct succession, to the first application unit, the first electron beam generator, the second electron beam generator.
	summary
043026800	claims	1. A cover construction for a shielding container for the transportation and/or storage of radioactive fuel elements comprising: an outer transportation cover detachably, sealingly securable to an outer rim portion of the container; an inner shielding cover comprising an inner cover portion having overlapping sealing engagement with an inner rim portion of the container and an outer cover portion movable inwardly and outwardly relative to said inner portion; means defining a bayonet connection between said outer cover portion and the container; and screw and stop means connecting said inner and outer cover portions whereby rotation of said screw means effects limited inward and outward relative movement between said outer and inner cover portions. 2. The cover construction defined in claim 1 including sealing means interposed between the peripheral portions of the inner and outer cover portions to effect a seal therebetween on relative movement of said portions toward each other.
	abstract	The present invention relates to a method for obtaining a shielding element for minimizing the penumbra of a beam of hadrons outside a target area, the hadron beam being guided in a longitudinal direction by an irradiation unit, and the beam having a width (σ). The method includes: (i) defining a closed or open contour of said target area; (ii) providing a block having a longitudinal thickness capable of blocking the passage of said beam and having a lateral surface perpendicular to said longitudinal thickness; (iii) forming an aperture of a shape similar to said contour and crossing said longitudinal thickness of said block for letting through said beam, said aperture forming a longitudinal internal surface; and (iv) trimming said block so as to form a longitudinal external surface around said longitudinal internals surface, said longitudinal internal and longitudinal external surfaces delimiting a side wall.
06236698&	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS The following are descriptions on a reactor nuclear instrumentation system and a reactor power distribution monitor system including the reactor nuclear instrumentation system according to the present invention. Embodiments will be described hereunder with reference to the accompanying drawings. FIG. 1 is a block diagram schematically showing a configuration of a reactor power distribution monitor system according to one embodiment of the present invention. The reactor power distribution monitor system is applied to a boiling water reactor (BWR) and includes a reactor nuclear instrumentation system 30 comprising a detector and a signal processing device and a reactor power distribution computing device 31 for computing a power distribution of a reactor core. The reactor power distribution computing device 31 is a part of a process control computer and monitors a reactor core performance. In the boiling water reactor, a reactor pressure vessel 2 is housed in a reactor container 1, and a reactor core 3 is housed in the reactor pressure vessel 2. The reactor core 3 is cooled by a coolant combining a moderator. The reactor core 3 is mounted with many fuel assemblies 4 as shown in FIG. 2 and FIG. 3. In these many fuel assemblies 4, four fuels assemblies constitutes one set, and a control rod 5 having a shape of cross in its cross section is mounted between one set of fuel assemblies 4 so as to be freely take in and out from a lower portion. The reactor core 3 is constructed so that a plurality of sets of four fuel assemblies 4 are mounted therein. Further, the reactor core 3 is provided with a plurality of incore nuclear instrumented fuel assemblies 32 constituting a detector of the reactor nuclear instrumentation system. The incore nuclear instrumented fuel assemblies 32 are arranged at a position different from a position where the control rod 5 is arranged. Further, the incore nuclear instrumented fuel assembly 32 is arranged at a corner water gap G formed between four fuel assemblies 4, as shown in FIG. 2 and FIG. 3. Further, the incore nuclear instrumented fuel assembly 32 includes a thin and long tube-shaped nuclear instrumentation tube 33. The nuclear instrumentation tube 33 is dispersively provided with a plurality of neutron detectors 34 which function as a fixed type (stationary or immovable) neutron detecting means and a .gamma.-ray heat detectors 35 which function as a fixed type .gamma.-ray detecting means in an axial direction thereof. In the nuclear instrumentation tube 33, a plurality of fixed type neutron detectors 34 are dispersively arranged as an LPRM detector at equal intervals in a core axial direction, and in this manner, a neutron detector assembly 37 is constructed. In a boiling water reactor, usually, four fixed type neutron detectors 34 are dispersively arranged at equal intervals in the core axial direction. Further, each of the neutron detectors 34 is electrically connected to a signal processing device 40 by means of a signal cable 38 penetrating through a penetration portion 39, and in this manner, a power range neutron flux measurement system 41 is constructed. Further, in the nuclear instrumentation tube 33, a plurality of fixed type .gamma.-ray heat detectors 35 are dispersively arranged in the core axial direction as a gamma thermometer in an arrangement of assembly (called merely gamma thermometer herein) and measures a gamma ray heat. The .gamma.-ray heat detectors 35 has the same numbers as the fixed type neutron detectors 34 arranged in the core axial direction or more, for example, 8 in the core axial direction, and in this manner, a .gamma.-ray heat detector assembly 44 is constructed as a gamma thermometer 44. Each of the .gamma.-ray heat detectors 35 of the gamma thermometer 44 is electrically connected to a gamma thermometer signal processing device 48 by means of a signal cable 45 penetrating through a penetration portion 46, and in this manner, a gamma thermometer power distribution measurement system 50 is constructed. The reactor nuclear instrumentation system 30 is composed of the power distribution neutron measurement system 41 and the gamma thermometer power distribution measurement system 50. A detector group of the reactor nuclear instrumentation system 30 is housed in the incore nuclear instrumentation assembly 32. The incore nuclear instrumentation assembly 32 measures a neutron flux and a .gamma.-ray heat at a predetermined measuring point in the reactor core 3. In the reactor nuclear instrumentation system 30, the movable neutron detector and the .gamma.-ray detector are unnecessary, so that a mechanical drive device included in the conventional reactor nuclear instrumentation system can be omitted. Therefore, it is possible to simplify a structure of the reactor nuclear instrumentation system 30, and a driving part is dispensed in the reactor nuclear instrumentation system 30, so that maintenance free can be achieved, and also, an exposure work of a worker can be dispensed or greatly reduced. The reactor pressure vessel 2 or a primary pipe system (not shown) is provided with a reactor core condition (present) data measuring device 52 which measures core process data such as a coolant core flow rate (or an approximate re-circulation flow rate), a core pressure and a coolant inlet temperature, a control rod position in a control rod drive device and the like. Although the reactor core operating status (present) data measuring device 52 is simply illustrated as one measuring device in FIG. 1, actually, it is reactor core operating status data measuring means which is composed of a plurality of measuring equipments for measuring or monitoring various reactor core process data. The reactor core operating status data measuring device 52 is connected to a condition data processing device 55 via a signal cable 54 penetrating through a penetration portion 53, and in this manner, a process data measurement system 56 is constructed. The condition data processing device 55 of the process data measurement system 56 is not a exclusively independent device, but may be constructed as a process control computer or a part thereof. Thus, the process data measurement system 56 is included in the process control computer constituting the reactor core power distribution computing device 31. Further, the process data measurement system 56 may be constructed as a part of the reactor nuclear instrumentation system 30 in the light of a concept of a detector and a signal processing device. Moreover, the process data measurement system 56, the power range neutron flux measurement system 41 and the gamma thermometer power distribution measurement 50 are electrically connected to the reactor power distribution computing device 31. Then, a signal processed by respective signal processing devices 40, 48 and 55 is inputted to a power distribution computing module 58 of the reactor power distribution computing device 31. The reactor power distribution computing device 31 is composed of: a power distribution computing module 58 which computes a neutron flux distribution, a power distribution, a margin of a thermal operation limit value or the like in a reactor core 3; a power distribution learning module 59 which is inputted with and corrects the computed result from the power distribution computing module and obtains a reactor core power distribution reflecting an actually measured process data; and an input-output device 60 including a display device. The following signals are inputted to the reactor power distribution computing device 31. That is, the input signals includes a detection signal (.gamma.-ray heat measurement signal) S.sub.1 from the .gamma.-ray heat detector 35 functioning as a gamma thermometer, a neutron flux detection signal S.sub.2 from the neutron detector 34, and a reactor core present data detection signal S.sub.3 from the reactor core present data measuring device 52. The power distribution computing module 58 of the reactor power distribution computing device 31 processes the inputted reactor core present data detection signal S.sub.3 according to a three-dimensional nuclear thermal-hydraulics computing code with the use of a physical model stored in the process control computer, and then, computes a neutron flux distribution, a power distribution, a margin of a thermal operation limit value or the like in a reactor core 3. The physical model is a spacer model taking an influence on an node power by the fuel spacer into consideration. A computation result computed by using the physical model is inputted to the power distribution adaption (learning) module 59 which functions as a power distribution correction module. The power distribution learning module 59 corrects a power distribution computation result based on the physical model with reference to the .gamma.-ray heat measurement signal S.sub.1, and then, returns it to the power distribution computing module 58. Subsequently, the module 58 makes an evaluation with respect to a reactor power distribution reflecting actually measured data and having high reliability and a thermal limit value. By the way, as shown in FIG. 1 to FIG. 3, the incore nuclear instrumentation assembly 32 constitutes the reactor nuclear instrumentation system 30. The incore nuclear instrumentation assembly 32 is integrally arranged in the nuclear instrumentation tube 33 in a manner of combining a local power range monitor system (LPRM) 37 which is a neutron detector assembly functioning as a fixed type fission chamber (neutron detecting means), and the gamma thermometer 44 which is a .gamma.-ray heat detector assembly functioning as a fixed type gamma ray detecting means. Further, the nuclear instrumentation tube 33 is arranged in a state of vertically extending in the core 3 of the reactor. The LPRM 37 includes N (number, integer) (N.gtoreq.4), for example, four fixed type neutron detectors 34 which are dispersively arranged at equal intervals in an axial direction, and the gamma thermometer 44 includes 8 or 9 gamma (.gamma.) ray heat detectors 35 which are dispersively arranged in an axial direction. The neutron detectors 34 of the LPRM 37 and the gamma ray heat detectors 35 of the gamma thermometer 44 are housed in the nuclear instrumentation tube 33 while a coolant being guided so as to flow upwardly in the nuclear instrumentation tube 33. In FIG. 2 and FIG. 3, there is shown an example of the gamma thermometer 44 which is constructed in a manner that eight .gamma.-ray heat detectors 35 are arranged in a fuel effective portion H of the core axial direction. An arrangement interval of each .gamma.-ray heat detector 35 in the core axial direction is determined taking an arrangement interval of each neutron detector 34 of the LPRM 37 in the core axial direction into consideration. More specifically, if an axial distance between neutron detectors 34 of the LPRM 37 is set as L, in the gamma thermometer 44, four of the eight .gamma.-ray heat detectors 35 are arranged at the same axial position as the fixed type neutron detector 34, three of them are arranged at an intermediate position of the neutron detector 34 at an interval L/2, and the lowest .gamma.-ray heat detector 35 is arranged at a distance L/4 to L/2 below the lowest neutron detector 34 and in the fuel effective portion of 15 cm or more from the lower end of the fuel effective portion, and axial centers of these detectors are aligned with each other. In the case of locating the .gamma.-ray heat detectors 35 above the uppermost neutron detector 34, the uppermost .gamma.-ray heat detector 35 is arranged so as to be situated at a distance L/4 to L/2 above the lowest neutron detector 34 and in the fuel effective portion of 15 cm or more from the lower end of the fuel effective portion. The lowest .gamma.-ray heat detector 35 is arranged at a position near the lower end of the fuel effective portion as much as possible in a fuel effective length. In the case where the fuel effective length (371 cm at present) is divided into 24 nodes in a core axial direction, preferably, the .gamma.-ray heat detector 35 is arranged so that its center is aligned with the axial center of the second core axial direction node from the bottom. When such an arrangement is made, in the .gamma.-ray heat detector 35 of the gamma thermometer 44, it is possible to detect a .gamma.-ray heat on a lower end side of the reactor core, and to measure the .gamma.-ray heat on the lower end over a wider range of the fuel effective length in the core axial direction. This is because of preventing the following matter. That is, a power of the node on the lowest end is originally low due to a neutron leakage, and a sensitivity by the .gamma.-ray heat detector 35 is low, and further, a contributing range of gamma ray to the .gamma.-ray heat detector 35 is 15 cm or more as described later. For this reason, unless the .gamma.-ray heat detector 35 is separated 15 cm or more from the lower end of the fuel effective length, other .gamma.-ray heat detector 35 arranged at a core axial direction position measures a heating effect of .gamma.-ray from the vertical direction in the axial direction. On the contrary, the lowest .gamma.-ray heat detector 35 detects only .gamma.-ray heat contribution from above. As a result, a correlation equation of power measurement is different. In an axial directional design of a recent fuel assembly 4, there are many cases where a natural uranium blanket is used as the node on the lowest end. For this reason, even if the natural uranium blanket portion of a low power (output) is measured, the output signal of the gamma thermometer 44 is very low. Therefore, there is no means of interpolating and extrapolating a power distribution at a position below the lowest neutron detector 34. The gamma thermometer 44 has a long rod-shaped structure as shown in FIG. 4 and FIG. 5. The gamma thermometer 44 is a thin and long rod-shaped assembly having a diameter of e.g., about 8 mm .phi., and has a length of substantially covering a fuel effective length in a core axial direction. In the gamma thermometer 44, a cover tube 62 formed of stainless steel is used as a metallic jacket, and a metallic long rod-shaped core tube 63 is housed in the cover tube 62. Further, the cover tube 62 and the core tube 63 are fastened, and then, are fixed to each other by means of shrinkage fit, cooling fit or the like. A sleeve or annular space portion 64 constituting an adiabatic portion is formed between the cover tube 62 and the core tube 63. A plurality of e.g., 8 or 9 annular space portions 64 are dispersively arranged at intervals in an axial direction. The annular space portion 64 is formed by cutting an outer surface of the core tube 63 along a circumferential direction. Then, a gas having a low heat conductivity, for example, an argon Ar gas is encapsulated in the annular space portion 64. The annular space portion 64 may be formed on the cover tube 62 side which is a jacket tube. The .gamma.-ray heat detector 35 is provided at a position where the annular space portion 64 is formed, and thus, a sensor section of the gamma thermometer 44 is constructed. The core tube 63 has an internal hole 65 which extends to an axial direction along the center thereof. In the internal hole 65, a mineral insulated (MI) cable sensor assembly 65 is fixed by means of brazing, caulking or the like. The cable sensor assembly 66 is provided with a built-in heater 67 which functions as a correction rod-shaped exothermic body of the gamma thermometer 44 at the center thereof, and a differential type thermocouple 68 which functions a plurality of temperature sensors. The built-in heater 67 and the thermocouple 68 are hardened by an electric insulating layer or a metal/metal alloy filler 69, and then, are integrally housed in a metallic cladding tube 70. The metallic cladding tube 70 is closely abutted against others at both outer and inner surfaces thereof. The built-in heater 67 comprises a sheath heater and is integrally constructed in a manner that a heater wire 72 is coated with a metallic cladding tube 74 via an electric insulating layer 73. Further, the thermocouple 68 is integrally constructed in a manner that a thermocouple single wire 75 is coated with a metallic cladding tube 77 via an electric insulating layer 76. The differential type thermocouple 68 located in the internal hold 65 of the core tube 63 is arranged so as to correspond to the annular space portion 64, and thus, the .gamma.-ray heat detector 35 is constructed. As shown in FIG. 5, each thermocouple 68 is set in a manner that a high temperature side contact 78a is situated on the sensor section formed in the annular space portion 64, that is, on the center of the adiabatic portion in the axial direction, and a low temperature side contact 78b is situated at a downward position slightly separating from the adiabatic portion (the low temperature side contact 78b may be situated at an upward position slightly separating from the adiabatic portion). The thermocouple 68 is coaxially inserted around the built-in heater 67 by the same number as the .gamma.-ray heat detector 35. The gamma thermometer 44 is an incore power distribution detector (.gamma.-ray heat detector) assembly, and the incore power distribution measuring principle is shown in FIG. 6A and FIG. 6B. In a reactor such as a boiling water reactor or the like, a .gamma.-ray is generated proportional to a fission yield of a nuclear fuel mounted in the reactor core 3 housed in the reactor pressure vessel 2. The generated .gamma.-ray flux heats a structural element of the gamma thermometer 44, for example, the core tube 63. The heat energy is proportional to a .gamma.-ray flux, and then, the .gamma.-ray flux is proportional to a fission yield. In the annular space portion 64 of each .gamma.-ray heat detector 35 which is a sensor section of the gamma thermometer 44, a performance of eliminating a heat by a diametrical coolant is worse due to a heat resistance of the annular space portion 64. For this reason, there is generated a heat flux as shown by an arrow A, which makes a detour in an axial direction, and as a result, a temperature difference is caused. So, the high temperature side contact 78a and the low temperature side contact 78b of the differential type thermocouple 68 are arranged as shown in FIG. 5, and it is possible to detect the temperature difference by a voltage signal. The temperature difference is proportional to the .gamma.-ray heat, and therefore, it is possible to obtain a .gamma.-ray heat which is proportional to a local fission yield from the voltage signal of the differential type thermocouple 68. This is the measuring principle of the gamma thermometer 44. On the other hand, in the fuel assembly 4, as shown in FIG. 2, a fuel bundle (not shown) bundling up many fuel rods (not shown) is housed in a rectangular and cylindrical channel box 80. The fuel bundle is constructed in a manner that many fuel rods are bundled by a fuel spacer 81 (see SP position of FIG. 7) so as to provide a square lattice arrangement. The fuel spacer 81 holds a clearance between fuel rods. For example, seven fuel spacers 81 is dispersively arranged in the channel box 80 along an axial direction of the fuel bundle. The fuel assembly 4 is mounted in the reactor core 3 in a state that its upper and lower ends are tightly fixed by an upper tie-plate and a lower tie-plate, respectively. Each fuel rod housed in the fuel assembly 4 is fixed in a manner that a fuel sintering pellet is filled in a fuel cladding tube made of a zirconium alloy, and upper and lower ends of the fuel cladding tube are deposited by an end plug. For example, an uranium oxide fuel or a uranium/plutonium mixed oxide (MOX) fuel is used as the fuel sintering pellet. In a boiling water reactor, a great many of fuel assemblies 4 are mounted in the reactor core 3, and each fuel assembly 4 is formed with a coolant passage outside and inside the channel box 80. Many fuel assemblies 4 are mounted in the reactor core 3 in a state of standing together in the large number, and a computation of power distribution of the reactor core is carried out by means of the reactor power distribution computing device 31 according to a three-dimensional nuclear thermal-hydraulics simulation computation (three-dimensional nuclear hydrothermal computing code) stored in the process control computer. The three-dimensional nuclear thermal-hydraulics computing code has a spacer model. The reactor power distribution computing device 31 is also called as a reactor core power distribution computing device or as a reactor core performance monitor and is one of functions stored in the process control computer of the reactor. In the reactor power distribution computing device 31, the following information is inputted to the power distribution computing module 58. That is, the information includes a control rod pattern obtained from the reactor core present data measuring device 52, a core flow rate, a reactor doom pressure (internal pressure of reactor pressure vessel), and various parameters such as a reactor heat power and a detection signal of a core inlet coolant temperature obtained from the reactor core condition (present) data. In concrete, the reactor core present data signal S.sub.3 from the condition data processing device 55, the neutron flux signal S.sub.2 from the signal processing device 40 and the .gamma.-ray heat signal S.sub.1 from the gamma thermometer signal processing device 48 are respectively inputted to the power distribution computing module 58 as a measured data. Meanwhile, the power distribution computing module 58 of the reactor power distribution computing device 31 has a built-in three-dimensional nuclear thermal-hydraulics computing code which is a physical model. Then, an input signal is computed and processed with the use of the three-dimensional nuclear thermal-hydraulics computing code, and thus, an incore power distribution is obtained by computation (calculation). In such three-dimensional nuclear thermal-hydraulics computing code, an influence on an node power by the fuel spacer 81 is previously evaluated and is stored in the process control computer as a spacer model. In the fuel assembly 4, the fuel rods are bundled up, and then, 7 to 8 fuel spacers 81 are dispersively arranged in an axial direction so as to keep predetermined interval between fuel rods. The arrangement effect of the fuel spacers 81 has not been considered in the conventional three-dimensional nuclear thermal-hydraulics computing code. The fuel spacer 81 is mainly made of a zirconium alloy having a low neutron absorption. It is found that a cooling water which is a moderator is locally reduced due to an existence of the fuel spacer 81, and for this reason, a thermal neutron flux is decreased. Also, the fuel spacer 81 has an effect of absorbing a neutron although the effect is slight, and therefore, this effect should not be disregarded. Incidentally, in the case of computing a power distribution of the fuel assembly 4, frequently, the computation is carried out after the conventional fuel assembly is divided into 24 nodes in the axial direction. Although it is general that the number of nodes thus divided is 24, the fuel assembly may be divided into any of a range from 12 nodes to 26 nodes in accordance with a size of a reactor core. A node power is lowered due to an existence of the fuel spacer as described above, and it is found that there is an error of about 0.05 (i.e., 5% of node average power) at the maximum in a state standardized so that an average power of each node in the axial direction of the reactor core 3 becomes 1.0, depending upon an axial direction position and a diametrical direction in the reactor core 3. However, in the conventional movable neutron flux measuring device (TIP), all of power distribution of 24 nodes in the core axial direction is read, and then, the whole node power is computed as a reference measurement signal according to the three-dimensional nuclear thermal-hydraulics computing code of the reactor power distribution computing device 31. Then, in the power distribution learning module which corrects the computed result, a node power including the fuel spacer is corrected, and as a result, the fuel spacer effect in the axial direction is accurately read (captured). A power distribution of an axial direction of the reactor core 3 is computed according to the three-dimensional nuclear hydrothermal computing code on the basis of all nodes, for example, an axial measurement data less than 24 nodes and not the whole core axial direction node, and then, the computed result is learnt and corrected. In this case, however, if a gamma ray heat of the .gamma.-ray heat detector portion corresponding to each position of the nuclear instrumentation tube 33 is computed on the basis of the power distribution result computed according to the three-dimensional nuclear thermal-hydraulics computing code having no spacer model, a learning error is caused in the node having the spacer, and then, the influence is given to other nodes. Next, to give an example of a physical model of a three-dimensional nuclear thermal-hydraulics simulation computing code which is employed in a boiling water reactor, a reactor power computation will be explained with reference to a flowchart shown in FIG. 11. In this case, a three-dimensional nuclear thermal-hydraulics computing code is generally called as correction group will be explained below. [A] Input Data Reading: Data required for computation, that is, the whole reactor core coolant flow rate, a control rod pattern, the whole reactor core generated power level, a power distribution (primary approximation) of the previous computation, a combustion distribution, LPRM and GT actually measured values, are inputted to the three-dimensional nuclear thermal-hydraulics computing code. The coolant flow rate is a reactor core operating status (present) data, and for this reason, some of process data are obtained from the operating status data measuring device 52, and then, are computed by means of the process data measurement system 56. The LPRM and GT actually measured values are inputted from each neutron detector 34 of the LPRM 37 and the .gamma.-ray heat detector 35 of the gamma thermometer (GT) 44. The power distribution (primary approximation) of the previous computation and a combustion distribution are previous data stored in the reactor power distribution computing device. [B] Supposition of the Initial Value of Incore Power Distribution P.sub.ijk : Usually, an incore (in-channel) power distribution or the like is supposed as the initial value required for a repeating computation called as Void Iteration, and then, the supposed value is set as a temporary incore power distribution P.sub.ijk for the following computation to advance the following computation. In this case, subscripts i and j are indicative of fuel assembly 4 position in the reactor core 3 and the subscript k is indicative of a reactor core axial position. [C] Computation of Incore Void Distribution: In order to compute an incore void axial distribution VF.sub.ijk of each fuel assembly 4 in the reactor core 3 and a reactor core average void axial distribution VFB.sub.k of a bypass range, required for an effective multiplication factor (constant) and an incore power distribution computation which will be described in the next item [D], the incore void axial distribution VF.sub.ijk of each fuel assembly 4 and the reactor core average void axial distribution VFB.sub.k are computed according to the following procedures. (1) Computation of Each Fuel Assembly In-channel Flow Rate W.sub.ij : A coolant flowing into the reactor core 3 is divided into an in-channel (incore) flow flowing in each fuel assembly 4 and a bypass range flow in each fuel assembly 4 at a the reactor core bottom portion, and then, these in-channel flow and bypass range flow again join together at a reactor core top portion outlet. Therefore, a distribution computation need to be carried out so that each pressure loss of the in-channel flow rate W.sub.ij and the bypass range flow rate BPF of each fuel assembly 4 becomes equal when these flows pass through respective channels in the reactor core 3. Moreover, there are a kind of fuel assembly 4 (e.g., 8.times.8 fuel or 9.times.9 fuel) and a kind of an orifice (e.g., peripheral orifice and center orifice) as a factor of giving a great influence to the distribution of the incore flow rate. For this reason, a coefficient of pressure loss for each axial portion of a hydrothermal characteristic representative fuel assembly channel and fuel assembly 4 is inputted every kind of the fuel assembly in the reactor core 3 and every kind of the orifice. The coefficient of pressure loss uses the result previously computed according to a reactor core hydrothermal analysis code every kind of the fuel assembly in the reactor core 3 and every kind of the orifice. The computation of the in-channel flow rate distribution is carried out according to a repeating computation because a pressure loss of each fuel assembly 4 depends upon a power distribution and a void distribution. In this case, the void distribution in fuel channel (channel box) gives a great influence to the computation of pressure loss. A pressure loss of the fuel assembly is classified into four, that is, a friction pressure loss, a local pressure loss, a position pressure loss and an acceleration pressure loss. In order to compute the pressure loss, a known two-phase flow pressure loss equation is used such that an equation used in a usual single-phase flow is multiplied by a two-phase flow friction resistance magnification. (2) Computation of Each Fuel Assembly Incore Enthalpy Axial Distribution H.sub.ijk and Bypass Range Core Average Enthalpy Axial Distribution HB.sub.k : In this enthalpy axial distribution computation, each fuel assembly incore (in-channel) enthalpy axial distribution H.sub.ijk and the bypass range core average enthalpy axial distribution HB.sub.k are computed by using the temporary incore power distribution P.sub.ijk used as the initial value, the in-channel flow rate W.sub.ij of each fuel assembly 4 computed in the following item [D], and the inputted bypass range flow rate BPF. There is already a heat generated by a nuclear fission in the fuel rod as a factor of increasing an enthalpy of each fuel assembly in-channel node, and besides, there are gamma heat in the fuel rod, a neutron moderation of the coolant, gamma heat, and a heat transfer effect to the bypass range via a fuel channel. Considering these factors, an enthalpy of each fuel assembly In-channel node is computed. The enthalpy computation is carried out for each fuel assembly channel with the use of an inputted core inlet coolant enthalpy, an incore power distribution P.sub.ijk and the in-channel flow rate W.sub.ij successively over a range from the bottom to the top portion of the reactor core. Supposed that a bypass flow is sufficiently and uniformly mixed in the bottom portion of the reactor core, in the bypass range enthalpy axial distribution, an average distribution of the reactor core is used. The factors of increasing an enthalpy of each bypass range node include a heat generated by a nuclear fission in the fuel rod, and besides, gamma heat in the fuel rod, a coolant in the bypass range, contribution by a neutron moderation-absorption of the control rod and gamma heat, and a heat transfer effect to the bypass range from a fuel channel, and a heat transfer effect from the bypass range to the outside of the reactor core. Considering these factors, an enthalpy of the bypass range core average enthalpy axial distribution HB.sub.k is computed. The computation is carried out successively over a range from the bottom portion to the top portion of the reactor core with the use of the following mathematical equation as the inputted reactor core inlet coolant enthalpy, the bypass flow rate BPF and the reactor core axial power distribution P.sub.k. [Mathematical expression 1] ##EQU1## (3) Computation of Each Fuel Assembly In-channel Void Rate Axial Distribution VF.sub.ijk and Reactor Core Average Void Axial Distribution VFB.sub.k of Bypass Range: Based on each fuel assembly in-channel (incore) enthalpy axial distribution H.sub.ijk and the bypass range core average enthalpy axial distribution HB.sub.k computed in the above item [C], the fuel assembly in-channel void axial distribution VF.sub.ijk and the reactor core average void axial distribution VFB.sub.k of a bypass range are computed. In order to compute a void content (amount or rate) from a coolant enthalpy, there is the following equation (1) as a method of computing a void content of a sub-cool range. [Mathematical expression 2] EQU Quality: X.sub.ijk =(H.sub.ijk -h.sub.sat)/(h.sub.g -h.sub.sat) (1) In place of the above equation (1), the following equation (3) is computed on the basis of a liquid enthalpy HL.sub.ijk. [Mathematical expression 3] EQU Flow Quality: XF.sub.ijk =(H.sub.ijk -HL.sub.ijk)/(h.sub.g -HL.sub.ijk) (2) Then, a flow quality XF.sub.ijk is computed with the use of an equation in which a drift flux model is applied to an expression of relation of the flow quality XF.sub.ijk and the void content. In this case, h.sub.sat represents an enthalpy of saturated water, and h.sub.g represents an enthalpy of saturated steam. The reactor core average void rate axial distribution VFB.sub.k of a bypass range and an outlet void content VFB.sub.ex are computed from the bypass range core average enthalpy axial distribution HB.sub.k, and the computing method is the same as the In-channel case. [D] Computation of Effective Multiplication Factor and Incore Power Distribution: This item [D] is a so-called nuclear computation section, and a nuclear constant of each node of the reactor core 3 is computed with the use of the incore in-channel void rate axial distribution VF.sub.ijk and the bypass range void axial distribution VFB.sub.k computed in the above item [C], and then, an effective multiplication factor k.sub.eff and an incore power distribution P.sub.ijk are computed. A power P.sub.ijk of each node is computed according to a nuclear fission by a neutron in each node. Therefore, in order to compute the incore power distribution P.sub.ijk, a neutron flux distribution .phi..sub.ijk in the reactor core 3 must be computed. Now, assuming that the neutron flux .phi. is divided into three groups, that is, a fast neutron flux .phi..sub.1, an intermediate neutron flux .phi..sub.2 and a thermal neutron flux .phi..sub.3 according to an energy of neutron flux .phi., these three groups, that is, neutron flux .phi..sub.1, .phi..sub.2 and .phi..sub.3 are obtained by solving the following diffusion equations (3), (4) and (5). [Mathematical expression 4] ##EQU2## -D.sub.2.gradient..sup.2.phi..sub.2 +.SIGMA..sub.2.phi..sub.2 =.SIGMA..sub.sl1.phi..sub.1 (4) EQU -D.sub.3.gradient..sup.2.phi..sub.3 +.SIGMA..sub.3.phi..sub.3 =.SIGMA..sub.sl2.phi..sub.2 (5) where, PA1 where, PA1 k.sup..infin. : infinite multiplication factor of each point of reactor core PA1 M.sup.2 : neutron migration area PA1 A.sup..infin. : correction term for diffusion of fast neutron and thermal neutron PA1 where, PA1 where, PA1 where, PA1 CPR=CP/ABP . . . (13) PA1 where, PA1 where, PA1 where, PA1 where, PA1 where, PA1 where, PA1 where, PA1 where, D.sub.g : diffusion coefficient of energy g-group PA2 .SIGMA..sub.g : eliminated cross section PA2 .SIGMA..sub.slg : moderating cross section PA2 .SIGMA..sub.fg : fission cross section PA2 .nu..sub.g :number of neutrons generated per fission PA2 .DELTA.X: length (about 15 cm) of X direction and Y direction of each node PA2 .DELTA.Z: length (about 15 cm) of Z direction, that is, axial direction of each node PA2 Fw: ratio of In-channel active coolant channel area to the whole coolant channel area PA2 .SIGMA..sub.lijk : eliminated cross section of fast neutron flux group PA2 .nu..sub.ijk : number of average neutron generated per fission of the whole neutron group PA2 CP: critical power PA2 ABP: actual power of fuel assembly PA2 IFT: type of design of fuel cross section PA2 Exp: node burn-up PA2 UH: node history relative water density PA2 U: node instantaneous relative water density PA2 P: node power PA2 f (IFT, Exp, UH): PA2 .DELTA.f1 (IFT, Exp, UH): PA2 .DELTA.f2 (IFT, Exp, UH): PA2 .DELTA.f3 (IFT, Exp, UH): effect by Xe, Sm PA2 S.sub.sp : spacer occupied area ratio of in-channel PA2 .rho..sub.sp : equivalent water density of fuel spacer PA2 V.sub.s : weight coefficient taking an axial volume rate of the node having the fuel spacer 81 into consideration PA2 X.sub.g : nuclear constant X of g-group (one of D.sub.g, .SIGMA..sub.g, .nu..sub.g .SIGMA..sub.fg, and .SIGMA..sub.slg) PA2 X.sub.g : node average nuclear constant in the case of considering effect by a g-group spacer PA2 f.sub.xg (IFT, EXP, UH): PA2 .DELTA.f.sub.1xg (IFT, EXP, U, UH): PA2 .DELTA.f.sub.2xg (IFT, EXP, UH, P): PA2 .DELTA.f.sub.3xg (IFT, EXP, U, UH, P): effect by Xe, Sn PA2 V.sub.sxg : weight coefficient taking an axial volume rate in the node having the spacer of g-group nuclear constant X into consideration PA2 C.sub.0xg, C.sub.1xg, C.sub.2xg : adjustment factor of g-group nuclear constant determined every fuel type PA2 Wc.sub.k,m : computation value of .gamma.-ray heating value (read value) of the GT sensor located on the axial center of k node situated at the nuclear instrumentation tube position m PA2 P.sub.k,m,n : axial k node average power of the fuel assembly n in four fuel assemblies around the nuclear instrumentation tube position m PA2 .DELTA.P.sub.k-1 :P.sub.k-1 -P.sub.k PA2 .DELTA.P.sub.k+1 :P.sub.k+1 -P.sub.k PA2 c: correlation function from node power to .gamma.-ray heating value PA2 m: nuclear instrumentation tube position PA2 n: four assemblies located around the nuclear instrumentation tube PA2 W(z): .gamma.-ray heating value at a z position in an axial direction PA2 P(z): node core power density PA2 R(z): correlation equation from power density to .gamma.-ray heating value PA2 .lambda.: .gamma.-ray transport mean free path of core axial direction (obtained by gamma ray Monte Calro computation as shown in FIG. 22) PA2 Wc.sub.k,m : computation value of .gamma.-ray heating value (read value) of the GT sensor located on the axial center of k node situated at the nuclear instrumentation tube position m PA2 P.sub.k,m,n : axial k node average power of the fuel assembly n in four fuel assemblies around the nuclear instrumentation tube position m PA2 .DELTA.P.sub.k-1 : P.sub.k-1 -P.sub.k PA2 .DELTA.P.sub.k+1 : P.sub.k+1 -P.sub.k PA2 c: correlation function from node power to .gamma.-ray heating value PA2 m: nuclear instrumentation tube position PA2 n: number of four assemblies located around nuclear instrumentation tube In this case, g is a suffix of 1, 2 and 3 PA3 (infinite multiplication factor having Xe to base power)-1.0 PA3 effect by difference between node history relative water density UH and instantaneous relative water density U PA3 Doppler effect correction by difference from temperature in base output PA3 (active coolant) passage PA3 equilibrium Xe to base power, value of nuclear constant X.sub.g PA3 effect by difference between history relative water density and instantaneous relative water density PA3 Doppler effect correction by difference from temperature in base power In order to solve the above diffusion equation, assuming that a buckling (see the following mathematical expression 5) of neutron flux of each energy group is equal, the above equations of energy three-group such as fast, intermediate and slow neutron fluxes are summarized to the diffusion equation of the fast neutron group, and then, only fast neutron flux distribution .phi..sub.lijk is computed. [Mathematical expression 5] ##EQU3## The incore power distribution P.sub.ijk is computed by adding an effect by the intermediate neutron flux .phi..sub.2 and the thermal neutron flux .phi..sub.s to a nuclear fission by the fast neutron flux distribution .phi..sub.lijk thus obtained. (1) Computation of Effective Multiplication Factor k.sub.eff and Incore Fast Neutron Flux Distribution .phi..sub.lijk : Assuming that a buckling B.sup.2.sub.g of neutron flux of each energy group is equal, the fast neutron flux .phi..sub.1 is distributed in the reactor core according to the following equations (6) and (7) which modifies the diffusion equations [Mathematical expression 6] EQU .gradient..sup.2.phi..sub.1 +B.sup.2.phi..sub.1 =0 (6) [Mathematical expression 7] ##EQU4## In order to solve a numerical value of the above equation, the above equation is transformed into the following difference equation (8). [Mathematical expression 8] ##EQU5## [Mathematical expression 9] ##EQU6## In the above equation (9), for simplification, the fast neutron flux .phi..sub.lijk of node (i, j, k) is represented as .phi..sub.ijk. The above difference equation is prepared for each node of the reactor core 3, and then, is transformed into a simultaneous equation, and thus, the equation is solved so as to obtain a fast neutron flux distribution .phi..sub.ijk of each node of the reactor core. The fast neutron flux distribution .phi..sub.ijk is obtained by carrying out a repetition computation and making a numerical solution. The repetition computation is carried out at the same time with repetition computation of the fast neutron flux and the effective multiplication factor K.sub.eff, and is called as Source Iteration. The solution of the difference equation is as described above. Before obtaining the solution of the difference equation, there is a need of computing nuclear constants k.sup..infin..sub.ijk, M.sup..infin..sub.ijk, and A.sup..infin..sub.ijk of each node included in the difference equation and a boundary conditions contacting with an outer surface of the reactor core. In particular, the nuclear constant greatly varies due to a void content in the node and the boundary, and for this reason, the nuclear constant is newly computed every Void Iteration. A channel void content of each fuel assembly is computed with the use of the In-channel void content (In-channel void rate axial distribution VF.sub.ijk and the bypass range void (void rate axial distribution) VFB.sub.k computed in the above item [C](3). A moderator relative history density U.sub.ijk defined in the following equation (10) is computed, and then, nuclear constants k.sup..infin..sub.ijk, M.sup.2.sub.ijk, and A.sup..infin..sub.ijk, of each node are computed by a fitting equation using these as parameters. [Mathematical expression 10] ##EQU7## In the case of computing nuclear constants k.sup..infin..sub.ijk, M.sup.2.sub.ijk, and A.sup..infin..sub.ijk of each node, in addition to the moderator relative history density U.sub.ijk, a moderator relative history density UH.sub.ijk, a burn-up E.sub.ijk, a presence of control rod C.sub.ijk and the like are used as parameters, and then, these effects are taken into consideration as the necessity arises. A great influence is given to the nuclear constant depending upon condition that the moderator is burnt up to the burn-up E.sub.ijk how void history, and for this reason, the moderator relative history density UH.sub.ijk is defined by the following equation (11) using introduced parameters. [Mathematical expression 11] ##EQU8## In this case, the power P.sub.ijk supposed in the above item [B] is used in the computation of the nuclear constant k.sup..infin..sub.ijk so as to make a Doppler correction, and also, an average xenon of the node is computed so as to make a xenon correction. A coefficient required for the computation of the above fitting equation is inputted as a library data with the use of the result computed every kind of fuel assembly and every parameter according to the fuel assembly nuclear hydrothermal computing code. (2) Computation of Reactor Core Power Distribution P.sub.ijk : A power distribution P.sub.ijk is computed with the use of the incore fast neutron flux distribution .phi..sub.ijk computed in the above item [D](1). By using the conditions supposed in the above item [D](1), a power P.sub.ijk of each node (i, j, k) is expressed by the following equation (12). [Mathematical expression 12] ##EQU9## [External Character 1] .SIGMA..sub.lijk and .nu..sub.ijk are computed by a fitting equation using the moderate relative density U.sub.ijk and by a fitting equation using the burn-up E.sub.ijk as parameter, respectively. A coefficient used in these fitting equations is computed in the same manner as the nuclear constant computed in the above item [D](1). Therefore, first, .SIGMA..sub.lijk and .nu..sub.ijk required for computing the incore power distribution P.sub.ijk are computed by the fitting equation, and then, the incore power distribution P.sub.ijk is computed by the above equation (12) with the use of k.sup..infin..sub.ijk and .phi..sub.ijk computed in the above item [D](1). [E] Void Repetition Computation Convergence Criterion and Power Distribution Learning: This item [E] is a so-called Void iteration convergence criterion. More specifically, a comparison is made between the incore power distribution P.sub.ijk supposed in the above item [B] and the power distribution P.sub.ijk computed in the above item [D](2). The comparison is carried out with respect to all nodes in the reactor core, and if the comparative result is coincident, the Void Iteration converges. If the comparative result is not coincident, the computation sequence returns to the above item [B], and then, the power distribution P.sub.ijk is again supposed and corrected, and thus, the computation stated in the items [B] to [E] is repeated until the comparative result is coincident. In the convergence criterion, a comparison of effective multiplication factor of the whole reactor core is carried out at the same time with the comparison of power distribution. By the way, in the case of learning a power distribution, in the Void Iteration, further, a comparison is carried out between a .gamma.-ray heat (actually measured heating value) from the gamma thermometer (GT) 44 and a .gamma.-ray heat (computed heating value) from the computed power distribution P.sub.ijk, and then, the difference between heating value is computed as a ratio. The difference of ratio is interpolated and extrapolated with respect to the axial node having no GT detector 35, and then, is computed as a difference data (actually measured value)/(computed value) between a reactor core coordinate position having each GT 44 and 24 axial modes for a .gamma.-ray heat computation of GT 44 of a nuclear instrumentation coordinate to which the actually measured value is applicable on the basis of a symmetry of the reactor core although there is no GT 44. The difference data is BCF.sub.ijk shown in FIG. 10. A computation value of the node power distribution around the GT 44 is corrected so as to be adapted to the BCF.sub.ijk, and Void Iteration is repeated so that the .gamma.-ray computed heating value of GT 44 is coincident with the actually measured value. In the case where the Void Iteration converges, and the .gamma.-ray computed heating value of GT 44 is coincident with the actually measured value, and further, the computed power distribution P.sub.ijk is coincident with the previous repeated power distribution, the computation sequence proceeds to a computation of a thermal margin value in the reactor core 3 which will be described in the following item [F]. If coincidence is not obtained, the computation sequence returns to the item [B], and then, the power distribution P.sub.ijk is again supposed and corrected, and thus, the computation stated in the items [B] to [E] is repeated until the comparative result is coincident. [F] Computation of Thermal Margin Value In the item [F], a thermal margin value of each node of the reactor core 3 is computed with the use of the numerical solution converged in the Void Iteration. However, the solution obtained from the Void Iteration is an average value of each node. In the case of computing a thermal margin value, there is a need of carrying out a computation relative to the maximum generating power fuel rod. Then, the power P.sub.ijk of each node computed in the above item [D](2) is divided by the number of fuel rods of node, and then, is multiplied by a local power keeping coefficient, and thus, a power of the maximum power generating fuel rod of each node is first computed. The following computation of thermal margin value is carried out with respect to the fuel rod generating the maximum power. The void content and burn-up of the node gives a great influence to the local power keeping coefficient required for the computation of the thermal margin value, in addition to the presence of a control rod inserted adjacent to or in the vicinity of the node. In the present computing code, these three variables are used as a parameter, and the local power keeping coefficient is computed by a fitting equation for each node. A coefficient required for the fitting equation is computed with the use of the result previously computed every fuel assembly 4 and every each parameter according to the fuel assembly nuclear characteristic computing code, and thus, is inputted to the present computing code as a library data. (1) Computation of Maximum Linear Heating Generation Ratio LHGR.sub.ijk : The maximum linear heat generation ratio LHGR.sub.ijk of each node (i, j, k) is obtained by dividing it by an axial unit node length .DELTA.Z because a power of the maximum power generation fuel rod of each node has been already computed. The maximum linear heat generation ratio of the whole node of the reactor core is set as the whole reactor core maximum linear heat generation ratio MLHGR. (2) Computation of Minimum Critical Power Ratio MCPR: A critical power ratio CPR is defined by the following equation (13). [Mathematical expression 13] A critical power CP is a power which is anticipated that the fuel assembly 4 of a computing target generates transition boiling, and is obtained from a GEXL correlation equation based on an experiment simulating a shape of an actual fuel rod. A critical power ratio CPR is an index indicative of a thermal freedom until the fuel assembly 4 of a computing target generates transition boiling. In the present (condition) computing code, the critical power ratio CPR is computed for each fuel assembly 4, and the minimum of the computed ratios is set as a minimum critical power ration MCPR. [G] Output of Computed Result: In this item [G], the computed result is outputted as the necessity arises. By carrying out the repetition computation as described above, an axial power distribution of each fuel assembly is obtained. This is a typical method for obtaining the axial power distribution of each fuel assembly. In fact, a secondary correction model is introduced into the method, and it is general that a design is made in order to improve an accuracy of the power distribution and the effective multiplication factor. However, the explanation is not a subject matter of the present invention and it is hence omitted. A reactor power distribution computing method is applicable to the method mentioned hereinbefore, and has the following features as described below. More specifically, a fuel spacer effect should be taken into consideration in parameters of nuclear constants k.sup..infin..sub.ijk, M.sup.2.sub.ijk and A.sup..infin..sub.ijk of core axial node having the fuel spacer 85. For example, in a state that the control rod is not inserted, k.sup..infin..sub.ijk is obtained by the following equation (14). [Mathematical expression 14] ##EQU10## Thus, the infinite multiplication factor is obtained. However, in this case, conventionally, the effect by the fuel spacer 81 has not been taken into consideration with respect to values of the node history relative water density UH and the instantaneous relative water density U. On the other hand, the moderator relative density U.sub.ijk is defined by the following equation (10). [Mathematical expression 15] ##EQU11## In each fuel spacer 81, a spacer portion is additionally computed from the In-channel passage void VF.sub.ijk in the case of disregarding the fuel spacer, and then, a new moderator relative density U.sub.ijk,sp is computed by the following equation (10A). [Mathematical expression 16] ##EQU12## The moderator (water) relative density U.sub.ijk,sp defined in the above equation (10A) is stored as a history relative water density UH for each spacer position of each fuel assembly 4, and an infinite multiplication factor of the node having the fuel spacer 81 is computed in the following manner. For example, in a state that the control rod 5 is not inserted into the reactor core 3, an average infinite multiplication factor k.sup..infin..sub.ijk of the node having the fuel spacer shown in the above equation (14) is expressed as below by a load average of k.sup..infin..sub.jk,UN in the case of disregarding the spacer into consideration and the infinite multiplication factor k.sup..infin..sub.ijk,UN,SP (however, U and UH is replaced with USP and UHSP taking the fuel spacer into consideration) taking a fuel spacer defined in the same manner into consideration, and adjustment factors (C.sub.0 +C.sub.1 U+C.sub.2 U.sup.2). In this case, C.sub.0, C.sub.1 and C.sub.2 are each a constant different from a fuel type. Namely, the above equation (14) is rewritten into the following equation (14A). [Mathematical expression 17] EQU [(1-V.sub.s)k.sup..infin..sub.ijk,UN +V.sub.s k.sup..infin..sub.ijk,UN,sp ][C.sub.0 +C.sub.1 U+C.sub.2 U.sup.2 ] (14A) In the same manner, infinite multiplication factors M.sup.2.sub.ijk and A.sup..infin..sub.ijk of the node having the fuel spacer are respectively defined by the following load average of the moderator relative density U.sub.ijk in the case of disregarding the fuel spacer and the moderator relative density U.sub.ijk,sp in the case of taking the fuel spacer into consideration. [Mathematical expression 18] EQU (1-V.sub.s)M.sup.2.sub.ijk,UN +V.sub.s M.sup.2.sub.ijk,UN,sp (15) EQU (1-V.sub.s)A.sup..infin..sub.ijk,UN +V.sub.s A.sup..infin..sub.ijk,UN,sp (16) The state that the control rod 5 is inserted into the reactor core 3 is a method for preparing a data library used conventionally, and the method is readily carried out by using a ratio to a state that the control rod 5 is not inserted. By defining in the manner as described above, it is possible to correct a library data such as the infinite multiplication factor of the node having the fuel spacer 81, migration area or the like, and a neutron flux can be accurately computed in the node having the fuel spacer 81. According to this definition, a node power is obtained by the same equation as the above equation (12) with respect to the node having the fuel spacer 81. [Mathematical expression 19] ##EQU13## In this case, in order to take the spacer effect into consideration, a symbol, to which a mark * is given as an additional character, is indicative that the parameter in the case of having a spacer is obtained by being subjected to a load average process to compute a node average parameter. The computing method is the same as the equations (14A), (15) and (16). According to the above explanation, the correction first group diffusion equation is obtained on the assumption that buckling of neutron flux .phi..sub.1, .phi..sub.2 and .phi..sub.3 of each energy group is equal. Even in the case of considering an influence caused by the fact that the thermal neutron distribution of the core axial node diverges from a basic mode by a spectrum mismatch effect between the node and the node adjacent thereto, the aforesaid concept for obtaining the diffusion equation may be adaptable. Further, without using the correction first group diffusion equation, in the case of using a three-group diffusion equation, each of the constants D.sub.g, .SIGMA..sub.g. .nu..sub.g .SIGMA..sub.fg, and .SIGMA..sub.slg can be generally expressed as shown below by a load average of a value in the case of taking the fuel spacer into consideration and a value which does not take it into consideration, and by adjustment factors (14B) and (14C). In this case, there is shown an example of a nuclear constant in the case where the control rod is not inserted. [Mathematical expression 20] ##EQU14## [Mathematical expression 21] ##EQU15## Therefore, this spacer model is generally applicable in addition to a correction first group code. Next, the following is a description on an operation by the neutron flux distribution monitor system and reactor core power distribution computing method. According to the above first embodiment, in the reactor power distribution computing device 31, various core present data such as control rod pattern obtained from the present data measuring device 52 of the reactor core 3, core flow rate, reactor doom pressure, core inlet coolant temperature are collected to the present (condition) data processing device (including the process control computer) 55, and then, a reactor thermal power (output) or the like is computed. The reactor core present (operating status) data measuring device 52 is actually composed of a plurality of monitor equipments and is the general term of a device for collecting process data of various operation parameter of the reactor and is expressed as one measuring equipment for simplification. Further. The process data processed by the present data processing device 55 is transferred to the three-dimensional nuclear thermal-hydraulics computing code of the reactor core power distribution computing device 31 (used as part of the process control computer or as an exclusive computer independently located) as various required data. The three-dimensional nuclear thermal-hydraulics computing code computes an incore power distribution with the use of these process data and the reactor core nuclear instrumentation data S.sub.1 and S.sub.2. In the reactor core power distribution computation, a correction spacer nuclear constant parameter is held with respect to a node having the fuel spacer 81, in addition to the nuclear constant parameter which does not take the fuel spacer 81 into consideration as conventionally. Further, in the history relative water density of the fuel spacer portion, the equation (10A) is integrated as the equation (11) so that the moderator relative history density UH.sub.ijk,sp is held. For example, as shown in the above equations (14A), (15) and (16), in the node having the fuel spacer 81, the effect by the fuel spacer 81 is taken into consideration, and then, in accordance with the necessity, a parameter reflecting as a function of a burn-up, history relative water density and instantaneous relative water density is subjected to a load average process so as to obtain a node average value. Whereby a solution of correction first group difference equation is obtained, and in first group node average neutron flux computing process, an effect of lowering first group neutron flux by the fuel spacer effect is reflected. Further, in order to obtain a node average power from first group neutron flux .phi..sub.1, contribution by second and third group neutron fluxes .phi..sub.2 and .phi..sub.3 need to be reflected. In this case, also, in the node having the fuel spacer 81, the equation (12A) is used in place of the equation (12), and therefore, in the effect by the spacer 81 to the first group neutron flux .phi..sub.1, a diffusion coefficient is large in the first group. Because of this reason, the contribution by the spacer effect is not so large. The effect obtained by the fact that the third group neutron flux .phi..sub.s, of the energy group lowers is taken into consideration. Furthermore, in the equation (12A), .SIGMA.*.sub.iijk includes the spacer effect by the fuel spacer 81 by load average, but may be used as .SIGMA..sub.ijk which does not take the spacer effect into consideration for simplification. As a result, it is possible to accurately reflect node power lowering on the spacer portion which has not been taken into consideration in the conventional power distribution computing method, with respect to the node having the fuel spacer 81. Moreover, in this first embodiment, it is possible to learn and correct the axial power distribution with the use of the LPRM detectors which are less then 24, four or more measurement data, and the three-dimensional nuclear thermal-hydraulics computing code evaluating an influence on the node power by the fuel spacer. The actual thermocouple output signal S.sub.1 of the gamma thermometer 44 is converted from a voltage into a gamma ray heating value by means of the gamma thermometer signal processing device 48, and then, is inputted to the reactor power distribution computing device 31. In the reactor power distribution computing device 31, a correction in difference between a simulation computation value of the .gamma.-ray heating value obtained from the reactor core power distribution computed by the three-dimensional nuclear thermal-hydraulics computing module 58 and the actually measured .gamma.-ray heating value is computed as a ratio by means of the power distribution correcting (learning) module 59. Further, in the power distribution learning (adaption) module 59, a numerical difference ratio limited to the axial direction, that is, a ratio of the computed .gamma.-ray heating value and the actually measured .gamma.-ray heating value is set as a correction data of the .gamma.-ray heating value difference correction of the total core axial node by interpolating and extrapolating the ratio data to each axial node by a straight line or a quadratic curve. Moreover, a learning correction of the power distribution learning module 59 of the reactor power distribution computing device 31 will be described below with reference to a flowchart of FIG. 10. The power distribution computing module 58 of the reactor power distribution computing device 31 is built in the process computer and computes an incore power distribution according to the aforesaid method with the use of the three-dimensional nuclear thermal-hydraulics computing code which takes an influence on the node power by the fuel spacer into consideration. A .gamma.-ray heat computation value Wc.sub.k,m on the .gamma.-ray heat detector 35 position is computed by the following equation (17) on the basis of a power of a node corresponding to a height position of the .gamma.-ray heat detector of four fuel assemblies 4 located around the .gamma.-ray heat detector 35 and a power of a vertical node adjacent to the node. [Mathematical expression 22] ##EQU16## An actually measured value of the .gamma.-ray heating value Wm.sub.k,m at .gamma.-ray heating value measurement points k and m is inputted to the power distribution learning module 59, and then, the above .gamma.-ray heating value Wc.sub.k,m is compared with a .gamma.-ray heating value measurement value Wm.sub.k,m, and thus, a correction coefficient BCF.sub.k,m is computed by the following equation (18). [Mathematical expression 23] EQU BCF.sub.k,m =Wm.sub.k,m /Wc.sub.k,m (18) A .gamma.-ray heating value correction coefficient BCF.sub.k,m is an index indicative of a difference (error) between the actually measured value of the .gamma.-ray heating value and the .gamma.-ray heat computation value according to the physical model. The .gamma.-ray heating value is proportional to a power of fuels around the .gamma.-ray heat detector 35, and for this reason, a power distribution computation value P.sub.k,m,n is corrected by the following equation (19) with the use of the correction coefficient BCF.sub.k,m, whereby it becomes possible to obtain a power distribution which eliminates an error by the physical model and has high reliability. In this case, Pa.sub.k,m,n shows a node power at the corrected .gamma.-ray heating value measurement points k and m. [Mathematical expression 24] EQU Pa.sub.k,m,n =BCF.sub.k,m.times.P.sub.k,m,n (19) By the way, the .gamma.-ray heat detector 35 of the gamma thermometer 44 is located continuously in only core axial direction, and at the node position having no .gamma.-ray heat detector 35, it is impossible to compute the correction coefficient BCF.sub.k,m. For this reason, in other core axial nodes, the correction coefficient BCF.sub.k,m obtained at measurement points k and m is computed by interpolating and extrapolating it to a straight line or a quadratic curve, and then, the total node power Pa.sub.k,m,n of axial direction is computed with the use of the correction coefficient BCF.sub.k,m. Moreover, the fuel assembly having the gamma thermometer 44 in the core diametrical direction position is learned and corrected with the use of a signal from the gamma thermometer 44 on the identical position on the basis of symmetry of the reactor core. In this case, a subscript k shows a core axial node, and subscripts m and n merely show a coordinate of reactor core based on the relationship between the fuel assembly 4 and nuclear instrumentation tube. The coordinate makes it possible to replace with the coordinate i and j of reactor core of the fuel assembly 4 used in the above equations (1) to (16). Furthermore, there is the following method as a method for correcting the power distribution computed by the physical model which refers to the actually measured value of the .gamma.-ray heating value. The neutron flux .phi..sub.k,m,n obtained via the void repetition computation of the equation (8) is corrected with the use of the correction coefficient BCF.sub.k,m, and then, a correction neutron flux .phi.a.sub.k,m,n is computed by the following equation (20). [Mathematical expression 25] EQU .phi.a.sub.k,m,n =BCF.sub.k,m.times.GF.sub.k,m.multidot..phi..sub.k,m,n (20) In this case, GF.sub.k,m is previously computed according to a lattice computation in an equation of transformation from the .gamma.-ray heating value into a neutron flux. In the case of substituting the correction neutron flux .phi..sub.k,m,n for the equation (8), it is a matter of course that the above equation (8) is not satisfied. Then, in the following equation (8A) and the equation (8), (B.sup.2.sub.i.j.k,+.DELTA.B.sup.2.sub.i,j,k is obtained by correcting buckling B.sup.2.sub.i,j,k of neutron flux of each energy group and satisfying the equation (8). In this case, .DELTA.B.sup.2.sub.i,j,k is indicative of a correction of buckling B.sup.2.sub.i,j,k of neutron flux when the correction neutron flux .phi.a.sub.i,j,k satisfies the equation (8). [Mathematical expression 26] ##EQU17## In the axial position having no .gamma.-ray heat detector 35, it is possible to obtain the correction neutron flux .phi.a.sub.k,m,n of the total reactor core axial node, that is, the incore neutron flux distribution the correction neutron flux .phi..sub.i,j,k computed with the use of the correction coefficient BCF.sub.k,m obtained by interpolating and extrapolating it with a straight line or a quadratic curve in the axial direction. Thus, in the core diametrical direction, it is possible to compute a correction of buckling .DELTA.B.sup.2.sub.i,j,k with respect to all of core axial nodes based on symmetry of the reactor core. The power distribution computed in the above manner is displayed by means of a display device of the input/output device 60 as shown in FIG. 1. As described above, in the power distribution computing module 58 including the three-dimensional nuclear thermal-hydraulics computing code as the physical model, in order that the power of each core axial node of four assemblies 4 around the nuclear instrumentation tube 33 is adapt to the node power correction previously computed, the correction is distributed to the peripheral nodes on the basis of the computed result at the same proportional distribution, and then, the corresponding power adjustment factors or first group neutron flux adjustment factors of each of nodes are anticipated, and thus, returned to the power distribution computing module 58 including the three-dimensional nuclear hydrothermal computing code. The three-dimensional nuclear thermal-hydraulics repetition computation and repetition computation of learning correction are carried out, and finally, if the difference between the previous (n-1) node power Pn-1 of each core axial node and the node power Pn of the present (n) computed is smaller than a fixed value, the repetition computation converges. Then, a computation of operation limit value is carried out, and the computation is completed. The flow is shown in FIG. 10. In FIG. 7, a curved line a (shown by a mark .tangle-solidup.) is a power distribution computed result of the fuel assembly 4 in the case of taking a local distortion of neutron flux due to the fuel spacer portion into consideration in the three-dimensional nuclear thermal-hydraulics simulation computing code. A curved line b (shown by a mark .box-solid.) is a power distribution computed result of the fuel assembly 4 in the case where a local distortion of neutron flux due to the fuel spacer 81 is not taken into consideration. In FIG. 7, these curved lines a and b shows a corrected result so that a position where the GT signal is obtained is coincident with the GT signal. A portion A of FIG. 7 is enlarged in FIG. 8, and is shown by a broken line b. On the other hand, in FIG. 8, there is shown a true value by a solid curved line c based on measurement point data of the gamma thermometer (GT). A broken line d of FIG. 8 shows a node power curve before correction. The node power curve d before correction has been computed according to the three-dimensional nuclear thermal-hydraulics computing code which does not take the fuel spacer into consideration. Then, when the node power curve d before correction is corrected on the basis of the GT measurement point data, in the case where there is the output signal of the gamma thermometer (GT) 44 in the node having the fuel spacer SP, the power is corrected to a minus side larger than the case where the neutron flux distortion at the GT position of the node power curve d before correction is not taken into consideration, and then, a power of up and down (vertical) nodes adjacent to the node is corrected to a minus side with the same proportional distribution. The node power correction is made with the proportional distribution between nodes around the nuclear instrumentation tube, a learning correction .DELTA.S is large. In the case where there is a power peak between GT positions, the node having no fuel spacer SP has no GT measurement data at the peal position although the node power is high, and for this reason, the peak position become an maximum correction .DELTA.Ma. Therefore, the node power after correction is shown by the curved line d, and then, there is a problem that the node power peak value is evaluated smaller. On the other hand, as shown in a portion B of FIG. 7, in the case where there is a measurement data of the gamma thermometer (GT) 44 in the core axial node having no fuel spacer SP and there is no measurement data of the gamma thermometer (GT) 44 in the core axial node having the fuel spacer SP, the learning correction is small, and the power of the node having the fuel spacer SP is excessively evaluated because a local distortion of neutron flux is not taken into consideration. On the contrary, in the case where the three-dimensional nuclear thermal-hydraulics computing code evaluates an influence on the node power by the fuel spacer SP, like the present invention, as shown by the curved line a of FIG. 7, the power distribution by the fuel spacer SP has a concave and convex portion in the axial direction from first. Therefore, no excessive correction is made even if the correction is interpolated and extrapolated in the axial direction. That is, as shown in FIG. 7, the .gamma.-ray heating value of the GT detector 35 at the nuclear instrumentation tube position is coincident with a node power around the GT detector 35 converted by a correlation relation (in this case, its details are omitted) between a .gamma.-ray heating value and a node power. As seen from the core power distribution learned and computed by the three-dimensional nuclear thermal-hydraulics computing code and from the effect described in FIG. 7 to FIG. 9, in the power distribution computing method of taking the fuel spacer 81 into consideration, it is possible to make a learning correction with high precision on the basis of the measurement data of each GT detector 35 of the less axial GT 44, and it is found that the reactor power distribution can be effectively obtained with high precision. Next, the following is a description on a reactor nuclear instrumentation system according to a second embodiment of the present invention. In the reactor nuclear instrumentation system according to this second embodiment, the reactor core 3 is provided with a plurality of incore nuclear instrumentation assemblies 32 which constitute a reactor power detecting device in a core diametrical direction. As shown in FIG. 12 and FIG. 13, the incore nuclear instrumentation assembly 32 includes a nuclear instrumentation tube 33 mounted between four fuel assemblies 4. The nuclear instrumentation tube 33 is provided integrally with a neutron detector assembly 37 which functions as a fixed type LPRM, and a fixed type gamma thermometer 44. In the neutron detector assembly (LPRM) 37, N (number, integer) fixed type (stationary or immovable) neutron detectors 34 are dispersively arranged at equal intervals in a core axial direction. The fixed type neutron detector 35, for example, four are arranged. On the other hand, in the gamma thermometer 44, a plurality of fixed type .gamma.-ray heat detectors 34, and N .gamma.-ray heat detectors 34, for example, four are arranged in the same axial direction as the fixed type neutron detector 35. FIG. 12 and FIG. 13 show an arrangement such that the .gamma.-ray heat detectors 35 of the gamma thermometer 44 are dispersively arranged in the same axial direction as the fixed type neutron detector 34 of the N fixed type LPRM 37. In the incore nuclear instrumentation assembly 32, gain adjustment of the neutron flux detector 34 of the fixed type LPRM 37 is directly compared and corrected with the use of a measurement value of the .gamma.-ray heat detectors 35 of the gamma thermometer 44 which are arranged in the same axial direction so as to correspond to a reactor nuclear instrumentation system 30. The reactor nuclear instrumentation system 30 and the reactor power distribution computing device 31 has the same configuration as the reactor power distribution monitor system shown in FIG. 1, and therefore, the explanation of the overlapping portion is omitted. In the incore nuclear instrumentation assembly which is a detector of the reactor nuclear instrumentation system shown in FIG. 12 and FIG. 13, the number (N: integer) of .gamma.-ray heat detectors 34 of the gamma thermometer 44 in the core axial direction is the same number (N) as the fixed type neutron detector 34 located in the core axial direction, and is arranged in the same core axial direction. As shown in FIG. 1, the LPRM detection signal S.sub.2 of the fixed type neutron detector (LPRM detector) 34 is processed by means of the power range detector signal processing device 40. The signal processing device 40 includes a pre-amplifier, a high pressure source, a pulse height discrimination circuit, a gain adjuster circuit or the like, and their details are omitted. In the power range detector signal processing device 40, a plurality of LPRM detection signals S.sub.2 is made average so as to prepare a power range average output signal (APRM signal). If the APRM signal level exceeds a predetermined value, a trip signal for scramming the reactor is supplied to a logic circuit of a safety guard system. The safety guard system makes a decision whether a state when the plurality of APRM output the trip signal is an operating state required for scramming, according to a predetermined logical decision, and then, scrams the reactor. The power range neutron detector (LPRM detector) 34 of the reactor core 3 is an ionization chamber type detector, and a fission material (uranium) is applied to an inner surface of an outer wall of the neutron detector 34. A high voltage is applied between the outer wall and the central electrode. In the neutron detector 34, an inert gas Ar is encapsulated as an ionization gas. In the ionization chamber type neutron detector 34, succeeding electron circuits such as an amplifier, a pulse height discriminator filter vary in its characteristic with time, and a so-called drift phenomenon happens. Further, detection sensitivity varies depending upon a change of uranium U-235 applied onto the inner surface of the outer wall of the neutron detector 34. Taking these factors into consideration, the LPRM signal used as the APRM signal need to be used as the APRM signal after being properly corrected. On the other hand, a measurement signal (mV signal) S.sub.1 of a differential type thermocouple 68 constituting the .gamma.-ray heat detectors 35 of the gamma thermometer 44 is converted from an analog signal into a digital signal by means of an A/D converter processor (not shown), and thereafter, is amplified in digital form. For this reason, the output signal S.sub.1 of the differential type thermocouple 68 has no change in a heat field, and then, almost no drift phenomenon happens. In the reactor nuclear instrumentation system, the .gamma.-ray heat detectors 35 of the gamma thermometer 44, that is, a gamma thermometer (GT) detecting portion is located on the same position as the neutron detector 34 of the fixed type LPRM 37. Thus, there is no need of obtaining a signal of the LPRM detector position by moving it like the movable type neutron detector (TIP). In the reactor nuclear instrumentation system of this second embodiment, a calibration of the signal output of the fixed type neutron detector 34 is directly carried out with the use of the .gamma.-ray heating value computed from the detection signal from the .gamma.-ray heat detector 35 situated at the same level in the axial direction, whereby it becomes possible to make a calibration at a high speed and with high reliability without using the power distribution computing device which is mounted with the three-dimensional nuclear hydrothermal simulation computing code. In the incore nuclear instrumentation assembly 32, the GT detecting portion signal S.sub.1 located at the same position as the LPRM detector 34 is directly and electronically retrieved, and thereby, it is possible to measure a .gamma.-ray heating value by electronic data reading and conversion to .gamma.-ray heating value. Therefore, it is possible to carry out a calibration of the LPRM detector 34 for a very short time (e.g., 5 to 10 minutes), and at one time per day or one time per time. In this case, when the signal level S.sub.2 of the LPRM detector is calibrated as a .gamma.-ray heating value of the gamma thermometer (GT) 44, the LPRM input signal of the APRM is proportional to a local power, and it is possible to provide an excellent local average power which does not depend upon the result computed by the physical model of the three-dimensional nuclear thermal-hydraulics simulation computing code included in the process control computer. The .gamma.-ray heating value of the gamma thermometer (GT) 44 is substantially proportional to a node average power around the GT, and a local power distribution of the fuel assembly corner fuel rod on the nuclear instrumentation 33 side does not so depend upon the computed result as compared with the case of the LPRM detector 34. In the case of calibrating the LPRM signal at short intervals, a change in an uranium isotope of the LPRM detector 34 and in a local power distribution of a cross section of the fuel assembly 4 is supposed as zero. Further, the LPRM detection signal level is calibrated to a value approximate to the node average power around the LPRM detector 34. Thus, even if the signal level generated by the LPRM detector 34 contains a weight of the local power of the fuel assembly cross section, when the node power varies, the LPRM signal is proportional to the change, and then, changes. Therefore, as compared with the case of calibrating the LPRM signal via a predictive LPRM signal based on the computed result by the three-dimensional nuclear hydrothermal simulation model as the conventional reactor nuclear instrumentation system, reliability is high, and it is possible to readily calibrate the LPRM detector signal S.sub.2 even when the reactor power distribution computing device 31 is temporarily fault or is during maintenance. In particular, the power range detector processing device 40 is a device constituting the safety guard system, and also, the gamma thermometer signal processing device 48 is composed of a digital circuit which has a simple software and includes a microprocessor having a reliability more than a process computer. Therefore, the present invention has a higher reliability as compared with the case of depending upon the computed result of the reactor power distribution computing device 31 which has a physical model and repeats convergence computation with the use of many stored data. In the reactor nuclear instrumentation system, the GT 44 is fixed and arranged in the reactor core 3. A part of the GT detectors has the same number as the LPRM detector 34 in the core axial direction and is arranged in the identical core axial direction. Thus, in order to calibrate the LPRM detector 34, the GT detector is a GT measurement system which is composed of the minimum numbers. The three-dimensional nuclear thermal-hydraulics simulation computing code has an efficient accuracy. Even if the GT detector portion is a little in the core axial direction and learning (adaption) data point is a little, if the spacer model is sufficiently taken into consideration in the three-dimensional nuclear thermal-hydraulics simulation, it is possible to dispense the number of the GT detector portion, that is, the number of the located .gamma.-ray heat detectors 35 as the reactor nuclear instrumentation system. By the reactor nuclear instrumentation system, no mechanical movement as the TIP need to be carried out in order to scan all GT detector 35 of the reactor core 3, so that a .gamma.-ray heating value approximate to the node power in the vicinity of the LPRM detector position can be computed for a very short time (about 5 to 10 minutes). In the case of carrying out a power level adjustment (gain adjustment) of the fixed type neutron detector 34 by a reading value of the .gamma.-ray heat detector 35 having no drift, the adjustment is made without using the core axial power distribution computed result by the process control computer. Thus, it is possible to calibrate a change in sensitivity due to drift of each power range local power detector (LPRM detector) constituting a part of the safety guard system with high reliability. Therefore, it is possible to dispense the movable neutron flux measuring device or a .gamma.-ray flux measuring device which has been conventionally required for calibrating the LPRM detector. The gamma thermometer 44 is a fixed type, and the number of .gamma.-ray heat detectors 35 is the same as the number of LPRM detectors 34, and thus, the GT measurement system can be composed of the minimum components. In the reactor nuclear instrumentation system of this second embodiment, the .gamma.-ray heat detectors 35 has the same number as the fixed type neutron detector 34 in the axial direction and is arranged in the identical axial direction. In the case of carrying out a power level adjustment (gain adjustment) of the fixed type neutron detector 34 by a reading value of the .gamma.-ray heat detector 35 having no drift, when the LPRM detector 34 is shifted from the node center axial directional position, the adjustment is made without using axial interpolation and extrapolation by a straight line or a quadratic line of the core axial power distribution computed result by the process control computer. Thus, it is possible to calibrate a change in sensitivity due to drift of each power range local power detector (LPRM detector) constituting a part of the safety guard system with high reliability. Usually, an ionization chamber type detector is used as the power range neutron detector (LPRM detector) 34 of the reactor core 3, and a fission material (uranium) is applied to an inner surface of an outer wall of the neutron detector 34. A high voltage is applied between the outer wall and the central electrode. In the neutron detector 34, an inert gas Ar is encapsulated as an ionization gas. In the ionization chamber type neutron detector 34, succeeding electron circuits such as an amplifier, a pulse height discriminator filter vary in its characteristic with time, and a so-called drift phenomenon happens. Further, detection sensitivity varies depending upon a change of uranium U-235 applied onto the inner surface of the outer wall of the neutron detector 34. In order to correct a change in neutron detection sensitivity, the change has been conventionally calibrated by means of the TIP device. However, the TIP device requires a mechanical drive mechanism for moving the movable type neutron detector to a core axial direction. On the contrary, in the present second embodiment, the TIP device is unnecessary, and the fixed GT detector 35 is arranged at the same position as the LPRM detector 34. Thus, it is possible to directly obtain a signal at the axial direction position of the LPRM detector 34 without interpolation and extrapolation. The output signal of the LPRM detector 34 is mainly a detection signal in accordance with a thermal neutron flux in a corner gap of the fuel assembly 4. Thus, the signal level relates not only to the average node power of four fuel assemblies 4 around the node, but also to local power peaking of the corner fuel rod of the fuel assembly on the corner gap side of the nuclear instrumentation tube 33 which strongly contributes to a thermal neutron flux level of the corner gap. The local power peaking varies with the combustion of fuel, and in the case of calibrating the LPRM detector 34 by means of the conventional TIP device, the calibration is carried out at a rate of one time for one month, and correction is made taking a deterioration (change) of the neutron detector 34 into consideration. The fixed type gamma thermometer (GT) 44 is fixed in the reactor core, no mechanical movement as the TIP need to be carried out in order to scan all GT detector 35 of the reactor core 3 (requiring about one hour to two hours), and thereby, it is possible to measure a .gamma.-ray heating value by electronic data reading and conversion to .gamma.-ray heating value. Therefore, it is possible to carry out a calibration of the LPRM detector 34 for a very short time (e.g., 5 to 10 minutes), and at one time per day or one time per time. In the case of calibrating the LPRM detector 34, when converting the signal level of the LPRM detector 34 into a .gamma.-ray heating value of the GT 44, an input signal of the safety guard system of the reactor is proportional to a local power, and it is possible to provide an excellent local average power which does not depend upon the result computed by the physical model of the three-dimensional nuclear thermal-hydraulics simulation computing code included in the process control computer. In this case, the .gamma.-ray heating value of the fixed type gamma thermometer (GT) 44 is substantially proportional to a node average power around the GT, and a local power distribution of the fuel assembly corner fuel rod on the nuclear instrumentation 33 side does not so depend upon the computed result as compared with the case of the LPRM detector 34. Thus, it is possible to carry out calibration in proportion to the local power of the fuel assembly 4 around the GT 44 with high precision. FIG. 14 and FIG. 15 show a reactor nuclear instrumentation system and a reactor power distribution monitor system including the same system according to a third embodiment of the present invention. The reactor power distribution monitor system of this third embodiment has the same configuration as the reactor power distribution monitor system shown in FIG. 1 and includes a reactor nuclear instrumentation system 30 and a reactor power distribution computing device 31. The reactor power distribution monitor system of this embodiment relates to an improvement in the incore nuclear instrumentation assembly 32 which functions as a reactor power detecting device constituting the reactor nuclear instrumentation system 30. The incore nuclear instrumentation assembly 32 is composed of a plurality of fixed type neutron detector assemblies (LPRM) 37 and fixed type gamma thermometers 44 which are mounted in the reactor core 3 and is housed in a nuclear instrumentation tube 33. The fixed type neutron detector assembly (LPRM) 37 is constructed in a manner that N (number, integer) (N.gtoreq.4) fixed type neutron detectors 34 are dispersively arranged at a predetermined distance L in a core axial direction. On the other hand, the fixed type gamma thermometer 44 is constructed in a manner that (2N-1) fixed type .gamma.-ray heat detectors 35 are arranged in the core axial direction. N (number, integer) detectors of the .gamma.-ray heat detectors 35 are arranged in the same core axial direction as the fixed type neutron detector 34, and the remainder (N-1) are arranged with a distance L/2 at the intermediate position in the core axial direction of the fixed type neutron detector 34. The reactor nuclear instrumentation system shown in FIG. 14 and FIG. 15 shows the case where a locating number of the fixed type neutron detector 35 constituting the fixed type LPRM 37 is four. For example, in a boiling water reactor mainly used nowadays, an effective length of a core axial direction is about 146 inches (3708 mm). The reactor core is divided into 8 equal parts, and the LPRM detector 34 and the .gamma.-ray heat detector 35 are arranged. In this case, a distance L/2 of the core axial direction is about 18 inches (457 mm). In the above manner, the .gamma.-ray heat detector 35 which is a GT detector portion, is arranged at equal intervals L/2, and the .gamma.-ray heat detectors 35 covers a lower end to an upper end of the effective length of the core axial direction, whereby it becomes possible to secure a computing precision of the three-dimensional nuclear thermal-hydraulics simulation computing code by making a learning correction over the whole core axial direction. Therefore, it is possible to finely make a learning correction in the core axial direction as compared with the case of the above second embodiment, so that a computing precision of the three-dimensional nuclear thermal-hydraulics simulation computing code can be secured. In the reactor nuclear instrumentation system of this third embodiment, the fixed type .gamma.-ray heat detector 35 has the same number as the fixed type neutron detector 34, and is arranged in the same axial direction as the fixed type neutron detector 34. Further, (N-1) fixed type .gamma.-ray heat detector 35 is arranged at the intermediate position of the N fixed type neutron detector 34, and thereby, many .gamma.-ray heat detectors 35 are arranged in the core axial direction so as to obtain the GT detector signal. Therefore, it is possible to improve an axial power distribution measurement precision as compared with the reactor nuclear instrumentation system shown in the second embodiment. FIG. 16 and FIG. 17 show a reactor power distribution monitor system according to a fourth embodiment of the present invention. The reactor power distribution monitor system of this fourth embodiment has the same configuration as the reactor power distribution monitor system shown in FIG. 1 and includes a reactor nuclear instrumentation system 30 and a reactor power distribution computing device 31. The reactor power distribution monitor system relates to an improvement in the incore nuclear instrumentation assembly 32 which functions as a reactor power detecting device constituting the reactor nuclear instrumentation system 30. In the incore nuclear instrumentation assembly 32, a fixed type neutron detector assemblies (LPRM) 37 and a fixed type gamma thermometers 44 are housed in a nuclear instrumentation tube 33, and are formed into a rod-like structure. The neutron detector assembly 32 is constructed in a manner that N (N.gtoreq.4) fixed type neutron detectors (LPRM detector) 34 are dispersively arranged at a predetermined distance L in a core axial direction. On the other hand, the fixed type gamma thermometer 44 is constructed in a manner that N detector of 2N fixed type .gamma.-ray heat detectors 35 are arranged in the core axial direction. N detectors of the .gamma.-ray heat detectors 35 are arranged in the same core axial direction as the fixed type neutron detector 34, and the remainder (N-1) are arranged with a distance L/2 at the intermediate position in the core axial direction of the fixed type neutron detector 34. The last one of detectors is arranged at a position separating from a distance L/2 to L/4 below the lowest position of the fixed type neutron detector 34. The lowest .gamma.-ray heat detectors 35 is arranged in the fuel effective length in the core axial direction. The reactor nuclear instrumentation system shown in FIG. 16 and FIG. 17 shows the case where a locating number of the fixed type neutron detector 35 constituting the fixed type LPRM 37 is four. In a boiling water type reactor (BWR) mainly used nowadays, the reactor core 3 is divided into 8 equal parts, and the LPRM detector 34 and the .gamma.-ray heat detector 35 are arranged. In this case, a distance L/2 of the core axial direction is about 18 inches (457 mm). According to the correction first group three-dimensional nuclear thermal-hydraulics simulation computing code in the process control computer using a diffusion equation, a computing precision of the nodes on the uppermost and lower ends in the core axial direction is liable to become worse due to an influence by neutron leakage. For this reason, in the core lower end side, computation must be carried out with high precision even if there is the possibility of an error of computation. However, in the BWR, a void is not so generated in the core lower end side, and the reactor power is easy to become high in its characteristic. Therefore, in the core lower portion, the fixed type .gamma.-ray heat detector 35 is actually interpolated rather than by extrapolating the difference between the GT measurement signal and the detection signal, and therefore, excellent learning is carried out with high precision. Thus, as described in this fourth embodiment, it is preferable that in the core lower end side, the fixed type .gamma.-ray heat detector 35 is arranged further below the lowest neutron detector 34 in a range from L/4 to L/2. The position where the fixed type .gamma.-ray heat detector 35 is located on the lowest end, that is, a distance below the fixed type neutron detector 34, is natural uranium blanket range per up and down (vertical) nodes (occasionally, 2 nodes on the upper end side) in a design of an axial direction of the latest BWR fuel assembly. Thus, the reactor power on upper and lower ends is low, and there is no need of measuring the upper and lower ends in its precision. For this reason, it is preferable that the axial center of the fixed type .gamma.-ray heat detector (GT detector) 35 is set above 1 node (about 15 cm) or more from the lower end of the fuel effective length. Moreover, it is found that the GT detector 35 responses an average power in a vertical range of 15 cm in the core axial direction. In the case where the locating number N of the fixed type .gamma.-ray heat detectors 35 is N=4, it is preferable that the lowest GT detector 35 is located at the intermediate point L/4 (about 9 inches) between the lowest neutron detector 34 and the lower end of the fuel effective length. If the GT detector 35 is located within 15 cm from the lower end of the fuel effective length in the axial direction, a correlation equation between the GT detector and the node power around the GT as GT reading value must be prepared, or the following matter must be accepted. That is, a slightly measurement error is caused between the computation result of the GT reading value and the GT measurement signal from the lowest GT detector 35. In the reactor nuclear instrumentation system of this fourth embodiment, in addition to an axial arrangement of the .gamma.-ray heat detector 35 of the reactor nuclear instrumentation system of the third embodiment, the .gamma.-ray heat detector 35 is arranged below the lowest neutron detector 34. Thus, the fixed type .gamma.-ray heat detectors 35 is arranged so as to substantially equally cover the fuel effective length, so that the extrapolation can be reduced. Therefore, it is possible to compute a node power in the vicinity of the lower end higher than the upper end of the fuel effective length from the measurement result of the reactor core power distribution with high precision. FIG. 18 and FIG. 19 show a reactor power distribution monitor system according to a fifth embodiment of the present invention. The reactor power distribution monitor system of this fifth embodiment has the same configuration as the reactor power distribution monitor system shown in FIG. 1, and includes a reactor nuclear instrumentation system 30 and a reactor power distribution computing device 31. The reactor power distribution monitor system relates to an improvement in the incore nuclear instrumentation assembly 32 which functions as a reactor power detecting device constituting the reactor nuclear instrumentation system 30. In the incore nuclear instrumentation assembly 32 is formed into a long rod-shaped structure, and a fixed type neutron detector assemblies (LPRM) 37 and a fixed type gamma thermometers 44 are integrally housed in a nuclear instrumentation tube 33. The fixed type neutron detector assembly (LPRM) 37 is constructed in a manner that N (N.gtoreq.4) fixed type neutron detectors (LPRM detector) 34 are dispersively arranged at a predetermined distance L in a core axial direction of the nuclear instrumentation tube 33. On the other hand, the fixed type gamma thermometer 44 is constructed in a manner that (2N+1) fixed type .gamma.-ray heat detectors 35 are arranged in the core axial direction. N detectors of the (2N+1) .gamma.-ray heat detectors 35 are arranged in the same core axial direction as the fixed type neutron detector 34, and the remainder (N-1) are arranged at the axial intermediate position. Further, the remainder, that is, two detectors are arranged below and above the lowest and uppermost fixed type neutron detectors 34. The lowest fixed type .gamma.-ray heat detector 35 is arranged below a distance L/4 to L/2 from the lowest neutron detector 34 in the fuel effective length of the core axial direction, and on the other hand, the uppermost fixed type .gamma.-ray heat detector 35 is arranged above a predetermined distance, that is, a distance L/4 from the uppermost neutron detector 34 in the fuel effective length of the core axial direction. The reactor nuclear instrumentation system shown in FIG. 18 and FIG. 19 shows the case where the locating number N of the fixed type neutron detector 34 is four. In this case, a distance L/2 of the reactor core axial direction is about 18 inches (457 mm), for example. In the reactor nuclear instrumentation system, further, one fixed type .gamma.-ray heat detectors 35 is added to the case where the locating number 2N of the fixed type .gamma.-ray heat detectors 35 in the reactor nuclear instrumentation system of the fourth embodiment. The added fixed type .gamma.-ray heat detector 35 is arranged at a position further above the lowest fixed type neutron detector 34 in the axial direction within the fuel effective length. As described above, the fixed type .gamma.-ray heat detectors 35 is arranged at a position further above the lowest fixed type neutron detector 34 in the axial direction within the fuel effective length, and thus, it is possible to reduce an extrapolation of an error between the core axial power distribution measurement value and the computed value by the simulation (process control computer) in the vicinity of the uppermost end of the fuel effective length, so that a precision of the axial power distribution on the upper end portion of the reactor core can be improved. In the conventional neutron detector which measures a thermal neutron flux, the upper end of the nuclear instrumentation tube 33 has a plunger structure which is inserted into a hole portion formed on the lower surface of the upper lattice plate. The plunger structure is different from most of parts of the nuclear instrumentation tube 33 of the reactor core. For this reason, in the .gamma.-ray heat detector 35, a .gamma.-ray transmission is great although having thermal neutron distortion and is hard to be affected by a structure of the nuclear instrumentation tube. Therefore, the .gamma.-ray heat detectors 35 is arranged at a position further above the lowest fixed type neutron detector 34 in the axial direction within the fuel effective length, and thereby, it is possible to preferably measure the core axial power distribution in detail and to improve a learning precision of the three-dimensional nuclear thermal-hydraulics simulator in the process control computer. Furthermore, it is preferable that the position for locating the uppermost .gamma.-ray heat detectors 35 is below 15 cm or more from the upper end of the fuel effective length on the basis of the same reason as mentioned above. In the reactor nuclear instrumentation system of this embodiment, in addition to the axial arrangement of the .gamma.-ray heat detectors 35 of the reactor nuclear instrumentation system of the third embodiment, the .gamma.-ray heat detectors 35 is arranged above the uppermost fixed type neutron detector 34 and below the lowest fixed type neutron detector 34. Thus, the fixed type .gamma.-ray heat detectors 35 is arranged so as to substantially equally cover the fuel effective length, so that the extrapolation can be reduced. Therefore, it is possible to compute a node power in the vicinity of the lower end higher than the upper end of the fuel effective length from the measurement result of the reactor core power distribution with high precision. FIG. 20 and FIG. 21 show a reactor power distribution monitor system according to a sixth embodiment of the present invention. The reactor power distribution monitor system of this sixth embodiment has the same configuration as the reactor power distribution monitor system shown in FIG. 1, and includes a reactor nuclear instrumentation system 30 and a reactor power distribution computing device 31. The reactor power distribution monitor system relates to an improvement in the incore nuclear instrumentation assembly 32 which functions as a reactor power detecting device constituting the reactor nuclear instrumentation system 30. In the incore nuclear instrumentation assembly 32 is formed into a rod-shaped structure, and a fixed type neutron detector assemblies (LPRM) 37 and a fixed type gamma thermometers 44 are integrally housed in a nuclear instrumentation tube 33. In the incore nuclear instrumentation assembly 32 constituting the reactor nuclear instrumentation system 30, the fixed type .gamma.-ray heat detector (GT detector) 35, which is arranged above the lowest fixed type neutron detector 34 at a distance L/4, is added to the incore nuclear instrumentation assembly of the third to fifth embodiments. In this case, the distance L is an interval in the axial direction of the fixed type neutron detector 34. FIG. 20 and FIG. 21 show the case where the locating number N of the fixed type neutron detector 34 is four. In a boiling water type reactor (BWR), it is general that the core has an effective length of 144 or 146 inches in a core axial direction. In the case, of dividing the core axial direction length into 24 node equal parts, a node, in which the maximum linear heat generation ratio is easy to be generated during an operation in the axial power distribution of the fuel assembly 4, is 4 nodes to 6 nodes from the bottom. In particular, in the first half of reactor operating cycle, a lower power peak operation is allowed within a range of operating limit value of the maximum linear heat generation ratio as much as possible, and in the end of operating cycle, the core axial or upper peak power distribution is made. In a core reaction effective operating method (BSO operation), the maximum linear heat generation ration is easy to be generated in the 4 nodes to 6 nodes from the bottom at the initial period to the intermediate period of operating cycle. In order to precisely evaluate a degree of freedom with respect to the maximum linear heat generation ration, the .gamma.-ray heat detectors 35 is arranged in the vicinity of the node, and thereby, it is possible to make a learning correction on the basis of the measurement value at the vicinity of the maximum peak portion of the core axial power distribution, and to improve measurement precision. In the reactor nuclear instrumentation system of this sixth embodiment, a plurality of fixed type .gamma.-ray heat detectors 35 in the incore nuclear instrumentation assembly 32 are additionally arranged above the lowest fixed type neutron detector 34 at a distance L/4, in addition to the locating position of the fixed type .gamma.-ray heat detectors 35 included in the reactor nuclear instrumentation system shown in the third to fifth embodiments. The position where the added fixed type .gamma.-ray heat detector 35 is arranged is a position where the maximum peaking is easy to be generated in the core axial direction in the latest high burnup (combustion) 8.times.8 fuel or high burnup 9.times.9 fuel core. Therefore, it is possible to precisely monitor a power distribution at a core position where the maximum linear heat generation ratio is easy to be generated and to improve a measurement precision. In particular, in the fixed type gamma thermometer 44, in the case where the locating number of the gamma ray heat detector in the core axial direction is limited in a mechanical design, it is possible to improve a precision in the limited number, thus being optimal. Next, the following is a description on a reactor power distribution monitor system according to a seventh embodiment of the present invention. The reactor power distribution monitor system of this seventh embodiment relates to improvement of an arrangement of a fixed type neutron detector and a fixed type .gamma.-ray heat detector in the incore nuclear instrumentation assembly 32 constituting the reactor nuclear instrumentation system. The reactor power distribution monitor system has the same configuration as the reactor power distribution monitor system shown in FIG. 1 and its details are omitted herein. The reactor power distribution monitor system has the same configuration as that of each former embodiment, that is, the fixed type neutron detector assembly (LPRM) 37 and the fixed type gamma thermometer 44 are integrally housed in the incore nuclear instrumentation assembly 32 constituting the reactor nuclear instrumentation system. The core axial position of the fixed type neutron detector 35 of the neutron detector assembly (LPRM) 37 is coincident with the center of the nodes divided in the fuel axial direction, which is used in the reactor power distribution computing device 31, and further, the core axial position of the .gamma.-ray heat detector 35 is also coincident with the center of node. In the reactor nuclear instrumentation system of this seventh embodiment, in the case where the power distribution computing device 31 computes a response of the .gamma.-ray heat detector, a consideration is taken such that a range of gamma ray is longer a thermal neutron. Further, by taking not only the axial node having the .gamma.-ray heat detector 35 but also contribution by a .gamma.-ray heating value of up and down nodes adjacent to each other into consideration, it is possible to improve a precision of power distribution by the minimum computation. A .gamma.-ray generated by a fission reaction has a range longer than a thermal neutron, and for this reason, the .gamma.-ray heat detector 35 need to be arranged in a fuel effective length of the core axial direction and at a position separating from the fuel effective end with 15 cm in the core axial direction. Referring now to FIG. 22, FIG. 22 is a chart showing a distribution of detected sensitivity in the case where the origin takes a surface .gamma.-ray source (.gamma.-ray surface source) which distributes in the fuel cross section, and a .gamma.-ray heat detector is located in an axial direction (X-axis) of the nuclear instrumentation tube 33. As seen from the detected sensitivity distribution result, the .gamma.-ray heating value is reduced in accordance with an axial distance from the .gamma.-ray surface source. However, the .gamma.-ray heat contributes to the fixed type gamma thermometer (GT) 44 by 6 inches (15 cm) or more in the axial distance, and for this reason, in order to improve a computation precision of the GT 44 reading value, there is a need of taking a power distribution within 23 cm in axial distance into consideration. Therefore, in the case where there is a difference in power between the node having the .gamma.-ray heat detector 35 and the adjacent axial nodes, it is found that an influence is given to a reading value of the .gamma.-ray heat detector 35. Taking a change in the core axial power distribution into consideration, a .gamma.-ray heating value is integrated in the core axial direction as the following equation (21), and then, is obtained therefrom. [Mathematical expression 27] ##EQU18## In this case, a computation of an actual .gamma.-ray heating value to the fixed type gamma thermometer (GT) 44 is carried out by a fuel assembly nuclear characteristic computing code and a .gamma.-ray transport computation. The .gamma.-ray heating value is computed by supposing a uniform .gamma.-ray source distribution in the core axial direction, that is, a uniform axial power distribution, and based on this, when the above equation (21) is again defined, the .gamma.-ray heating value is expressed by the following equation (22). [Mathematical expression 28] ##EQU19## As seen from FIG. 22, the .gamma.-ray heating value may disregard the contribution from about 23 cm or more in the core axial direction. In a general BWR, the reactor core 3 is divided into, for example, 24 nodes, and one node is substantially 6 inches (15 cm). Thus, in order to detect a .gamma.-ray heating value, it is sufficient to considering nodes directly adjacent to each other and adjacent nodes far from one adjacent node. If the fixed type .gamma.-ray heat detector 35 is situated at the center of the axial node of the reactor core 3, as shown in FIG. 23, the axial node having the .gamma.-ray heat detector 35 of the GT sensor portion and upper and lower (vertical) nodes adjacent to each other merely be taken into consideration, and an integration range is made to have the same length, and thus, a relative equation becomes simple. The above equation (17) is an example of this case. On the contrary, in the case where the fixed type .gamma.-ray heat detector (GT detector) 35 which is a GT sensor portion of the fixed type gamma thermometer (GT) 44, is not situated at the center of the axial node of the reactor core 3, as shown in FIG. 24, an influence of .gamma.-ray heating value W other than adjacent nodes is given to the .gamma.-ray heat detector 35. Therefore, as shown in FIG. 25, a sensor reading value on the center of the axial node is temporarily computed, and then, is interpolated to an actual sensor position reading value. By making a comparison, when the .gamma.-ray heat detector 35 is situated at the center of the axial node, it is possible to readily calculate the .gamma.-ray heating value. According to this seventh embodiment, the core axial position of the .gamma.-ray heat detector 35 (GT sensor portion) of the fixed type gamma thermometer 44 is coincident with the center of the axial node, and thereby, it is possible to make a simple polynomial of the node having the .gamma.-ray detector 35 and a node mean power of upper and lower nodes adjacent to each other. [Mathematical expression 29] ##EQU20## In the case where the core axial position of the fuel spacer is coincident with the center of the axial node, a concave portion of the core axial neutron flux by the fuel spacer affects only the computed result of the node average neutron flux and the node average power distribution. Therefore, it is very convenient because there is no need of preparing an influence to adjacent nodes as a correlation equation used in the three-dimensional BWR simulation computing code. In the reactor power distribution monitor system of this seventh embodiment, the fixed type neutron detector 34 and the .gamma.-ray heat detector 35 of the incore nuclear instrumentation assembly 32 constituting the reactor nuclear instrumentation system are situated at the center of the node divided in the fuel axial direction. In the case where the fixed type neutron detector 34 is not situated at the center of node, a correction is made by interpolating the axial distribution of the read calculation value of the fixed type neutron detector of the core axial adjacent node, thus being very troublesome. Moreover, the .gamma.-ray heat detector 35 is a .gamma.-ray source contributing to the detector position, that is, the power distribution advantageously contributes within a range of 15 cm. Thus, even if the .gamma.-ray heat detector 35 is situated on the center of the axial node with a height of 15 cm, the .gamma.-ray heat detector 35 receives the influence of power distribution of the up and down (vertical) adjacent nodes. The influence of power distribution from the adjacent nodes is attenuated in series by a function near to an exponential of the locating position z from the .gamma.-ray heat detector 35. Therefore, in the case where the .gamma.-ray heat detector 35 is not situated at the center of the axial node, there is a need of computing a reading value by an axial non-symmetrical weight distribution of the axial power distribution in the node having the .gamma.-ray heat detector 35 and the adjacent nodes. Conversely, in the case of converting the reading value of the .gamma.-ray heat detector 35 into a peripheral power distribution, interpolation or extrapolation is made in the axial direction so as to make the computation easy, and thus, the read value need to be computed. In the reactor nuclear instrumentation system of this seventh embodiment, the .gamma.-ray heat detector 35 is coincident with the center of the axial node, and thereby, the same weight of the adjacent node is used in the correlation equation with respect to locating positions of all .gamma.-ray heat detectors 35, so that the computation can be made simple, and also, a precision can be improved. Further, the axial position of the fuel spacer is coincident with the center of the axial node, and hence, an axial distortion of the neutron flux in the three-dimensional nuclear hydrothermal simulation computing model becomes maximum at the node center. By only converting the node into a substantially average node data, it is possible to consider the axial effect of the fuel spacer, so that a computation precision can be improved in the axial distortion of the neutron flux of the three-dimensional nuclear thermal-hydraulics simulation computing model. It is finally to be noted that the present invention is not limited to the described embodiments and many other changes and modifications may be made without departing from the scopes of the appended claims.
	abstract	An apparatus for curing a printing plate made of or having photo-curable material, a method of curing such a printing plate, and a printing plate cured by the method. One embodiment of the method includes curing a printing plate made of or having photo-curable material thereon. The method includes producing light energy on part of the printing plate using a light exposure unit capable of generating at least a first illumination intensity and a second illumination intensity, such that curing can produce printing features on the plate that can be switched to have either flat tops or round tops according to the illumination intensity output by the light exposure unit.
054616565	description	DETAILED DESCRIPTION OF THE INVENTION Referring to FIGS. 1 to 3, there are four basic components in the preferred embodiment of the ECR-X source 10 of the present invention, namely, a spherical chamber 11 filled by a heavy gas at a low pressure, a cylindrical grid 12 that surrounds the chamber 11 and forms a resonant microwave cavity 13, a microwave power source 18, and a magnetic mirror formed by magnets 32 and 33, which is effective within sphere 11, as will be explained. The microwave source and magnets combine to form an electron cyclotron resonance (ECR) plasma within the sphere 11, as will also be explained. The actual source of the X-ray radiation is the small dielectric sphere 11 filled with a heavy non-reactive gas. The sphere 11 is about the size of a tennis ball, as indicated by the scale shown in FIGS. 2 and 3. This dielectric sphere is preferably made of quartz, but it may be glass or ceramic or other material that will not react with the heavy gas. The gas is preferably xenon, but it also can be krypton or argon, or any other inert gas. Further, the gas can be a heavy non-inert gas, so long as the gas is non-active in relation to the surrounding sphere, or a mixture of gases. A heavy gas is desired so that it will line the inside surface of chamber 11, absorb a substantial portion of the electrons from a hot electron ring, to be discussed, and emit X-ray energy. Further, the penetration of the atoms of the heavy gas into the dielectric wall of sphere 11 is negligible so that the gas pressure in chamber 11 remains constant during practically unlimited time. In addition, the start-up of a plasma is very easy in heavy gases as compared to light gases due to the smaller mobility of the heavy ions. Also the electron density reached in the heavy gas plasma is higher than in light gases at the same microwave power, because of the reduced losses due to the low mobility of the heavy ions. The sphere 11 is filled with the heavy gas in a well-known manner, for example, by evacuating the sphere on a commercially available vacuum pump, at an elevated temperature, to out gas any impurities in the quartz or other material of the sphere, in order to ensure prolonged life of the sphere as an X-ray source. Once evacuated and processed to remove impurities, the sphere is filled with the heavy gas, and the tubulation used for out-gassing and filling is sealed. In the preferred embodiment, the X-ray source apparatus includes a non-vacuumated cylindrical resonant cavity 13 which resonates in the TE.sub.111 mode. Cavities that resonate in other modes that have the electrical field component perpendicular to the static magnetic field axis also may be used. Cavity 13 is preferably formed of a grid 12, preferably made of beryllium or aluminum. The cavity 13 is about 1 liter in volume and it encompasses the closed spherical dielectric chamber 11. Microwave power is passed into cavity 13through the waveguide 14, which is connected through the flanges 15 and 16 with the waveguide 17 that supplies microwave energy from an oscillator 18, which may be a magnetron fed from an electrical supply 19. The microwave energy from oscillator 18 is preferably at a frequency of 2.45 GHz. The coupling between the magnetron 18 and the waveguide is conventionally optimized by means of a movable plunger 20 which may be mounted on a slide or threads for adjusting its position. The magnetron is protected against any reflected microwave energy by a conventional circulator device 21. FIGS. 2 and 3 show the construction of the X-ray radiating block in some detail. As shown in these figures, the resonant cavity is composed of a cylindrical lateral wall 12 and two flanges 22 and 23. The cylindrical wall 12 is made of a grid of light metal, such as beryllium or aluminum. The grid 12 contains the microwave energy employed, at 2.45 GHz, but is transparent to the X-rays emitted from sphere 11. If necessary, the base material of the grid 12, such as beryllium or aluminum, is preferably coated with a thin coating of high electrical conductivity material, such as copper, to minimize its heating. The coating can be about 10 micrometers thick. Moreover, air cooling (ambient or forced) of the grid 12 and resonant cavity 13 is advantageously employed to provide the quartz chamber cooling. Grid 12 is attached to a rectangular waveguide section 14 which is connected with the microwave supply 18 through the flange 15. The cylindrical wall 12 is fixed between the two flanges 22 and 23, which can be made of duraluminum or copper, and two dielectric cushions 24 and 25 which serve to support the spherical plasma chamber 11 within the resonant cavity. These dielectric cushions are preferably porous teflon, and they may be mechanically attached to flanges 22 and 23. Alternatively, the cushions may be a foam material or any other cushioning dielectric non-magnetic material that can be appropriately shaped and secured to support sphere 11 in place without absorbing microwave energy. The cylindrical wall 12 and two flanges 22 and 23, once being assembled in the apparatus of FIGS. 1 to 3, form a resonant cavity for the microwave energy at 2.45 GHz. This cavity is fed microwave energy from source 18 through the circulator 21, and waveguide sections 17 and 14, waveguide 14 being tightly connected to the flanges 22 and 23 to minimize any losses. Wave guides 14 and 17 are standard units of copper or aluminum. As shown in FIG. 2, flanges 26 and 27 are attached by bolts to flanges 22 and 23. A braided copper seal can be inserted between flanges 22 and 26, and between flanges 23 and 27, to be compressed when the bolts attaching the flanges are drawn together, to avoid any microwave energy leakage. Flanges 26 and 27 include cylindrical tubes 28 and 29, which are internally threaded. As shown in FIG. 2, the apparatus also includes two disk-shaped permanent magnets 32 and 33, which are preferably SmCo.sub.5, fixed in casings 34 and 35. These casings and tubes 28 and 29 have a common thread, so that the casings can be screw threaded with the tubes, and thereby adjust the position of magnets 32 and 33 by rotating the casings. The only element of the apparatus that must be occasionally replaced is the sphere 11, which has a nominal useful life of about 1000 hours of operation as the plasma chamber. The cost of this sphere can be quite low, and the replacement operation does not require any special skill. In this regard, the sphere 11 is removably mounted in the cavity 13. To replace the spherical chamber 11, one need only separate one of the flanges 26 or 27, replace the old chamber 11 by a new one, preferably leaving dielectric cushions 24 and 25 in place (unless their replacement is appropriate), and then re-attach the flange. This operation can be done in a short time. To start the ECR-X source, one starts the microwave oscillator 18. This transmits microwave energy at a frequency of 2.45 GHz through the waveguide to the resonant cavity 13. Since the magnets 32 and 33 are in place, the presence of the microwave energy in the chamber 11 causes the ECR plasma to form and the X-ray emission to begin. The X-ray emission is a bremsstrahlung with the peak energy between 100 and 200 keV depending on the microwave power applied, e.g., from 200 to 500 watts. To stop the X-ray emission, one simply turns off the microwave oscillator. This is adequate for many sterilization or preservation methods as disclosed herein. The effectiveness of any X-ray source to sterilize and preserve food and other materials is a function of the amount of X-ray energy and the time of exposure. According to the present invention, a method of sterilizing an article, a product or materials, in a rapidly moving processing line or the like can be achieved by placing the X-ray sources shown in FIGS. 1 to 3 one after another along the line so that each article or product and all the material advancing along the line receives a cumulative X-ray dose sufficient to cause sterilization. Also, the ECR chambers of successive sources can be offset from one another, so as to irradiate the article or product from different angles. The food products and other items to be sterilized or preserved using the foregoing X-ray source can be unwrapped or wrapped. It can, for example, include products wrapped in plastic foil, wine in bottles, sterile gauze in a carton, and other covered items, provided that the wrapping is transparent to the X-ray. An understanding of the operation of the ECR plasma in the sphere 11, and additional insight into the mode of operation of the X-ray source, can be gleaned with reference to FIGS. 4 to 7 and the following discussion. The disk-shaped magnets 32 and 33 in FIG. 2, are shown in FIG. 4(a) with an illustration of the magnetic field lines they produce. Also, the axial and radial directions are depicted. As shown by the closeness of the field lines, the magnetic field increases as one moves from the center of the magnetic field toward either of the magnets, as shown by the graph in FIG. 4(b). Also, it decreases as one moves from the center of the field in a radial direction, as shown by the graph in FIG. 4(c). In FIG. 5, the arrows within the plasma chamber 11 show the direction of the forces on the plasma particles, which are produced by the magnetic field, to create a well-known magnetic mirror. For microwave energy at 2.45 GHz, electron cyclotron resonance (ECR) occurs for a magnetic field strength of 865 gauss. The magnets 32 and 33, and their spacing, are chosen, and adjusted in position, so that the magnetic field strength along the axis between the magnets, as shown in FIG. 4(a), is higher than 865 gauss. This also is true for the field strength at the center of the magnetic field moving in a radial direction. However, as shown in FIG. 4(c), as one nears the spherical chamber wall the magnetic field decreases. Thus, near the internal surface of the sphere 11, along the centerline between magnets 32 and 33, a point will be reached where ECR resonance occurs, as shown in FIG. 6. This produces an equatorial ring 42 of hot electrons, which has been observed as an associated effect in some past ECR work. See W. D. Dougar-Jabon, K. S. Golovanivsky and V. D. Schepilov, "Accumulation of Multicharged Ions in Plasma with Electrostatic Well Induced by ECR", Physica Scripta (Sweden), Vol. 18, 506-507, 1978. The hot electron ring 42 also is shown in FIGS. 2, 3 and 7. In the operation of sphere 11, electrons in the hot electron ring spiral off from the ring due to the so-called flute instability, which is a well known physical phenomenon. These spiraling electrons collide with a dense layer of atoms of the gas within sphere 11, and with the internal surface of the sphere. As a consequence, X-rays are emitted in a divergent array, from the heavy gas particles and the material of the sphere 11, along an equatorial region about sphere 11, as shown in FIGS. 1 and 2. With the magnets 32 and 33 properly positioned, the hot electron ring or an ECR zone 42 will be circular and symmetrical within the sphere 11, and the X-ray emission will consequently be homogeneous along the equatorial region. By a proper adjustment of the magnets, the ECR zone can be properly placed, in proximity to the sphere wall, to achieve the optimum X-ray emission. Any X-rays passing back through waveguide 14 will diminish in strength with distance, and the divergent nature of the X-rays will cause them to be absorbed by the copper waveguide 14. Also, the waveguide 17 can include a turn to ensure that no X-rays will reach the region of oscillator 18. Alternatively, a waveguide turn can occur just outside one of the cylinders 28 and 29, without disrupting grid 12, so that grid 12 can extend the entire 360.degree. around the ECR plasma in sphere 11. In this fashion, the X-ray radiation can be emitted radially over 360.degree.. While chamber 11 has been described as a sphere, it can also be ellipsoid in shape. Other shapes, such as a cube, can also be used for chamber 11, so long as the chamber has enough physical strength to withstand the external pressure, but a sphere is preferable both because of its strength and symmetrical shape. Some of the parameters that are believed to be particularly advantageous are that the microwave resonant cavity 13 has a volume of about 1 liter, height of about 7 cm. and a diameter of 13 cm. The microwave energy is desirably at 300 to 500 watts at the stated frequency of 2.45 GHz. The ECR chamber 11 has a volume preferably about 100 cm.sup.3, and it is about 6 cm. in diameter, made of quartz glass, with a wall thickness of 1.0 to 1.5 mm. Also, the gas within the ECR chamber 11 is preferably xenon, at an internal pressure of 10.sup.-5 to 10.sup.-4 Torr, preferably 3 to 8.times.10.sup.-5 Torr, and more preferably 6 to 7.times.10.sup.-5 Torr. The X-ray peak energy is around 180 keV, with an integral intensity of around 1 watt. Further, the grid 12 is preferably duraluminum, with a thin coating of copper that is about 10 micrometers thick. The dielectric cushions 24 and 25 are preferably porous teflon, and the magnets 32 and 33 are SmCo.sub.5 disks, each having a height of about 2.5 cm., a diameter of about 5.5 cm., and a magnetic induction on the surface of about 0.4 Tesla. The magnetic field in the region of the hot electron ring is 865 gauss. The hot electron energy in the ECR plasma in sphere 11 is generally 150 to 250 keV, and the hot electron density is 2.times.10.sup.9 per cm.sup.3. The ring plasma volume is 3 cm.sup.3 ; the plasma lifetime is 5 microseconds; the hot electron current is about 0.2 milliamps; and the energy delivered to the wall by the hot electrons is about 40 watts. The source weighs about 20 kgs. and absorbs about 1 kW of electrical power. If X-ray emission is desired in only a window along the equatorial line of emission, a cylinder of lead can be employed, surrounding the grid 12, to cover the height over which the X-ray emission extends, leaving an opening over the area of the desired window of emission. The window may have a desired shape, e.g., a point source hole, a slit, a rectangle, an annular configuration, etc. Lead having a thickness of 5 mm will suffice. One set of operable dimensions for the grid 12 is to have solid members that are about 1 mm in width or diameter separated by spaces that are about 5 mm. FIGS. 8 to 11 illustrate various methods of radiating an article (or a product), a stream of articles, or a flow of a material with X-rays, for example, for sterilization or preservation. FIG. 8 shows a series of conveyors, each of which includes a belt 55 and a pair of rollers 57. Tunnels 58 are supported such that the conveyor belt 55 passes through or just below the tunnel, each tunnel having a series of lead or lead rubber curtains 59 of at least 3 mm, more preferably at least 5 mm, thickness. Also, an ECR plasma sphere 11, and the associated elements in FIGS. 1 to 3, are contained in each of the tunnels. An article or material to be sterilized or preserved will pass along the successive conveyors. As it meets a tunnel a first curtain will be pushed aside by the article, or automatically withdrawn as the article is sensed, by a photodetector or microswitch or the like. Successive curtains are preferably spaced so that the first curtain will close before the second curtain is opened. Once the article passes the last input curtain 59, it will be radiated as it continues along the conveyor, and then pass through the successive output curtains. By opening and closing the curtains in succession, no X-ray radiation will escape out of the ends of the tunnel. Alternatively, a single tunnel can include a series of ECR X-ray plasma sources and, as shown in FIG. 9, successive ones of the sources can be offset in order to radiate different portions of the article. By passing articles to be sterilized along the conveyors in FIG. 8, through one or more of the tunnels in FIGS. 8 and 9, the bacteria on the articles can be killed and the articles thereby sterilized. This method of irradiating an article with X-rays can be used for such articles as fruit or vegetables, chicken that may contain salmonella, shell fish or other foods, even wine in bottles, or wrapped meats. Also, it can be used for seeds that are to be used in foods, such as caraway seeds, and seeds that are to be planted. Further, it can be used for medical instruments and supplies, whether loose or packaged. The present invention also concerns a method for irradiating articles or materials such as whole blood and cellular blood components, for example, to inactivate lymphocytes and some viruses at dosages that are not harmful to the viability of other desired blood cell components. And it can be used for irradiating waste products which may contain blood, viruses, bacteria or other organisms that present a risk of spreading disease or infection if disposed without treatment, such as food processing by-products (e.g., raw chicken scraps), used or disposable medical devices, instruments, supplies and blood contaminated materials, and medical diagnostic test by-products (e.g., blood samples, fluids and cultures). The method of radiating fruits and other foods with X-rays according to the present invention eliminates the bacteria that causes the food to rot which acts to preserve the foods. Thus, the food can safely be shipped long distances, even over slow (and less expensive) travel routes. Further, fruits and other food can be allowed to ripen on the vines or in the orchards, so as to obtain a mature taste and appearance, as well as food value, and then sterilized locally, using the X-ray source and methods of the present invention, before being shipped to distant markets. Moreover, since the X-ray source and methods of the present invention can operate on loose or packaged products, the X-ray source can be located in a processing or packaging plant either upstream or downstream of the processing or packaging stations. The sterilization caused by radiation at doses established by the U.S. Food and Drug Administration may not remove all bacteria, and therefore foods may still require refrigeration, but it materially reduces the level of bacteria present and will prolong shelf life. For treatment of products where human or animal consumption of irradiated products is not a concern, the sterilization dosage may be sufficiently high to kill all undesired bacteria, viruses or other organisms. FIG. 10 shows a water supply pipe 62, which contains a series of centrally located ECR plasma sources 11, each with its associated elements shown in FIGS. 1 to 3. By passing the water along pipe 62, past the successive ECR plasma channels 11, the bacteria in a water supply can be killed and the water supply thereby sterilized. A similar process could be used for any fluid material flow, whether a gas, liquid or solid (solid, gel or particulate matter) flow. It is noted that the method of sterilizing water using the non radioactive X-ray source immersed in the flow according to the present invention does not present the potential hazard of contaminating the entire water supply, which hazard would exist if a radioactive X-ray source were to be used for such purposes. FIG. 11 shows a representation of a small scale sterilization station which includes a total of 24 ECR-X sources 11 arranged in a 4.times.6 matrix and a conveyor 100 for advancing articles along a path by and between the ECR-X sources. In this embodiment, each pair of a block of six ECR-X sources are coupled by a conventional microwave waveguide to a single microwave magnetron power supply, for example, a 12 kw c.w. commercial supply (not shown). Thus, only two such magnetrons are needed. The conveyor 100 is shown as having a tortuous path which is selected to maximize the dose of X-ray radiated from all directions on the articles being treated. Other paths, as well as more or lees ECR-X sources, could be used. The walls surrounding the track, and top and bottom of the station may be sealed to trap the X-rays by an appropriate material, e.g., lead lined walls and the aforementioned curtains for passing articles into and out of the station. One suitable conveyor 100 is a single chain having hooks from which the articles may be suspended spaced apart, singly or in bulk containers or bundles, as appropriate. Preferably, the articles travel in a common equatorial plane of the ECR-X sources. Where necessary, the ECR-X sources may be arranged with equatorial planes that are offset horizontally, angularly, vertically or some combination thereof, to provide an adequate cumulative dose to the articles being treated. Advantageously, the microwave magnetron sources may be driven by the same motor that operates the conveyor track. For example, a 100 kw gasoline-powered motor typical for operating a conveyor track can be used to produce 25 kw to operate the two magnetrons of this embodiment and to drive the chain conveyor. Such a facility can be constructed in any location where gasoline (or similar fuel source) can be obtained. If the conveyor path has a length of 25 meters and is advanced at a speed of 1 cm/sec, then the X-ray dose provided is on the order of 2.5 kgy and the yield is on the order of 1 kgy per ten seconds. The faster the conveyor speed the lower the cumulative dose. Thus, the operating parameters can be selected to provide the following yields for the identified products: pork 5 kg per 10 seconds; fresh fruit 2.5 kg per 10 seconds; white potatoes 25 kg per 10 seconds; and poultry 1 kg per 10 seconds. Advantageously, the entire station, composed of one or several blocks of 24 sources, can be installed on the bed of a large truck or trailer with a 100 kW motor and driven from location to location, for example, for processing grains, fruits and vegetables during harvesting in any location and environment. Either the truck bed may be sealed against X-ray leakage, or the bed may be parked inside a structure that can be sealed to trap X-rays. It should be understood that the ECR-X source and methods of the present invention also could be used to irradiate articles, products and materials in a batch mode, wherein the X-rays are generated within a X-ray shielded structure and the structure has a platform for supporting the articles to be irradiated and a door (or similar opening) for inserting and removing articles such that the door has appropriate seals against X-ray leakage and interlocks for preventing X-ray generation when the door is open. Door seals could include, e.g., braided wire contact seals, X-ray absorbing dielectric seals, and/or waveguide chokes. The interlock system used may be a dual interlock of the type used in conventional domestic microwave ovens. Another aspect of the invention is directed to a source and a method for irradiating body tissue with X-rays at a dosage level and for a time sufficient to medical or dental diagnostic or therapeutic purposes. Such methods include generating an ECR plasma to produce X-rays in a given direction, for example, to expose a film for X-ray evaluation of tissue, bone and other structures, including mammography and computer aided tomography (CAT scans). Such methods also include generating an ECR plasma to produce X-rays for medical therapeutics, for example, cancer therapy, diathermy, and activating X-ray responsive drugs. In this regard, the X-ray dosages to be used are those generally used in medical and dental diagnostic and therapeutic practices. Advantageously, the small and light weight of the ECR-X source, together with a lead shield that covers all of the cavity except a suitably shaped window, provide easy maneuverability to locate the source proximate to the patient and easy portability of the apparatus, for example, for a mobile medical clinic. In addition, the small size and simplicity of operation permits providing emergency service vehicles such as ambulances, fire rescue vehicles and the like with portable X-ray machines, which may be hand held, for obtaining X-ray images of injured patients prior to moving them. One skilled in the art will appreciate that the present invention can be practiced by other than the described embodiments which are presented for purposes of illustration and not of limitation.
044926685	claims	1. An improved nuclear fuel assembly having a plurality of elongated rods with an end plug disposed at least at one longitudinal end of each of said rods and having at least one tie plate with a plurality of circumferentially enclosed cylindrical tie plate sleeves with one of said tie plate sleeves having a circumferentially enclosed internal cavity for receipt therethrough of one of said end plugs of one of said elongated rods, wherein the improvement comprises: (a) said one tie plate sleeve having a circumferential outer surface and an elongated out-of-cavity flat on an end portion of said outer surface; and (b) said one end plug having a stud portion disposable in said internal cavity of said one tie plate sleeve and having a cylindrical body portion disposed between said one elongated rod and said stud portion, said cylindrical body portion including a longitudinally extending, internal flat portion on the stud-portion-facing end of said cylindrical body portion, with said internal flat portion seatable against and surroundably engageable with said out-of-cavity flat to prevent rotation of said one elongated rod about its longitudinal axis when said stud portion of said one end plug is disposed in said internal cavity of said one tie plate sleeve. 2. The improved nuclear fuel assembly of claim 1, wherein said one tie plate sleeve also includes an additional out-of-cavity flat generally identical with, and diametrically oppositely disposed from, said out-of-cavity flat, and wherein said cylindrical body portion of said one end plug further includes an additional internal flat portion generally identical with, and diametrically oppositely disposed from, said internal flat portion. 3. The improved nuclear fuel assembly of claim 1, wherein said internal flat portion has a radially tapered outer surface.
	description	The present invention is directed generally at neutron absorbing material, and more particularly at a method of using neutron absorbing materials to control and reduce the extent of damage and contamination resulting from an uncontrolled, unintended, or purposefully harmful release and distribution of radioactive materials. Radioactivity involves spontaneous disintegration of unstable atomic nuclei by the emission of subatomic particles including alpha particles (helium ions), beta particles (electrons), and/or neutrons along with the emission of electromagnetic radiation including X-rays or gamma rays. Due to the charged nature of alpha and beta particles, their penetrating ability into matter is limited. However, electromagnetic radiation and especially neutrons can penetrate deeply into matter and thick physical barriers are necessary to provide adequate shielding. Neutron emission and subsequent bombardment is especially damaging to living system and has the unique quality that it can interact with the nucleus of many atoms and destabilize them to cause/promote additional radioactivity. The release and subsequent interactions of neutrons with another atoms nucleus is the basis of nuclear fission. For example, when a neutron hits a uranium 235 atom, this atom in turn releases about 2.5 neutrons, on average, from the split nuclei. The neutrons released in this manner quickly cause the fission of two more atoms, thereby releasing four or more additional neutrons and thus a self-sustaining series of nuclear fissions, or a chain reaction is initiated resulting results in a sustainable release of nuclear energy. Analogously, when plutonium 239 itself absorbs a neutron, fission can occur, and on the average about 2.8 neutrons are released. Additionally, many atoms can be made radioactive after being hit with a neutron even though they are not capable of sustaining nuclear fission. Instead of releasing neutrons, these atoms release alpha particles, beta particles, and gamma radiation, all of which are highly damaging to living biological systems. Other classes of atoms, called neutron absorbers, can absorb neutrons and yet remain stable; i.e. not radioactive. Thermal neutrons (0 to 500 keV) are absorbed very effectively by specific elements including boron (or B-10 isotope), gadolinium, dysprosium, samarium, cadmium, or europium. Very high energy neutrons (1 to 15 MeV) are absorbed by elements such as zirconium, hafnium, tantalum, indium, hydrogen, or silver. The power of the atom has been harnessed in several ways including the development of nuclear weapons and in nuclear power plants. Modern nuclear power plants are inherently safe, however, accidents have occurred in the past. In 1979, a nuclear accident occurred in the United States at the Three Mile Island PWR near Harrisburg, Pa. and a small amount of radioactivity escaped from the containment vessel. In 1986, one of four nuclear reactors at Chernobyl exploded and burned for an extended period of time and radioactive material spread over the Ukraine, Scandinavia, and northern Europe. Since the terrorist attack on Sep. 11, 2001 in the United States, it has become increasingly apparent that the world is an inherently dangerous place. Terrorists, state enemies, criminals, etc. may have at their disposal a wide variety of mass destruction weapons including biological, chemical, and nuclear based weapons. This invention disclosure involves drastically reducing/minimizing the effects of nuclear based/derived weapons by the application of neutron absorber materials/particulates in various forms. Several key scenarios will be listed to give examples of the key embodiments of this invention. Other applications which are not listed but which are variations of this are also covered by this invention. It has been known since the 1950's that neutron absorbers may be used to moderate the reaction rate in nuclear reactors. Additionally, in the 1990's, it was known that neutron absorbers may have uses in waste packages for the long term storage of nuclear waste. However, there has been no previous knowledge for using neutron absorber materials in the manner identified herein. In a first embodiment, the present invention provides a method of controlling or containing radioactive contamination, the method comprising providing a neutron absorbing material, applying the neutron absorbing material to an open region of radioactive contamination, and absorbing emitted neutrons, thereby preventing expansion of radioactive contamination. In another embodiment, the present invention provides a method of controlling or containing a nuclear reactor breach comprising providing a neutron absorbing material, applying said neutron absorbing material to said nuclear reactor in a region of a reactor core, and controlling a fission reaction of said nuclear reactor by absorbing emitted neutrons. According to yet another embodiment, the present invention provides a method of protecting against radioactive exposure comprising supplying a neutron absorbing material, incorporating said neutron absorbing material into a construction material, and building a structure from said construction material incorporating said neutron absorbing material. The present invention provides methods of using selected neutron absorbing materials, and combinations of neutron absorbing materials to control and/or reduce the impact that an undesirable release of radioactive material can have on both humans and on the environment. According to the present invention, neutron absorbing materials may be employed in a curative or control capacity to limit the further damage and contamination after a release of radioactive material. Additionally, the present invention contemplates the use of neutron absorbing materials in a precautionary manner in which neutron absorbing materials are pre-positioned to prevent or control radioactive contamination. Consistent with the preset invention neutron absorbing materials are materials which contains from 1 to 100 atomic percent of elements or combinations of elements which have microscopic cross sections for thermal neutrons greater than 100 Barns. Exemplary materials have high macroscopic cross sections arising from the distribution of high microscopic cross section elements contained within the neutron absorbing materials. These materials may contain lanthanides having a cross section greater than 100 Barns, for example gadolinium, europium, dysprosium, erbium, or samarium which may be in combination with other neutron absorbing elements such as hafnium, zirconium, tantalum, silver, indium, and hydrogen, which are especially effective for absorbing higher energy neutrons. When these materials are employed to absorb neutrons, the neutrons are prevented both from inflicting health problems as well as from perpetuating nuclear reaction or radioactive contamination. According to a first exemplary embodiment, the present invention recognizes that, while nuclear power plants are inherently safe during general operating based on the many engineering safety controls, there is an increasing concern that nuclear power plants may be vulnerable to terrorist attack. An attack, for instance by an airplane hitting a nuclear power plant facility or using conventional explosives, could result in a reactor breech or deliberate destruction of the reactor cooling system resulting in overheating and loss of reaction containment resulting in a breech. In either case, radioactive elements could be released in to the atmosphere resulting in widespread contamination. Such a release of radioactive material could have continuing harmful effects for an extended period of time, as evidenced by the Chernobyl disaster where the reactor continued to release radioactive contaminants for many days until a physical containment barrier could by constructed by dropping liquid concrete from the air. Consistent with the first exemplary embodiment, the invention provides a method of rapidly containing nuclear contamination and reducing or eliminating any ongoing contamination. In the event of a loss of reactor integrity, whether as a result of a mechanical failure, a terrorist attack, etc., the neutron absorbing powder may be used to blanket or cover the reactor to, thereby, shutdown the nuclear reaction, i.e., the nuclear fission. By absorbing the neutrons that are released during the fission reaction, the reaction may be slowed or stopped by reducing the neutrons available to perpetuate the nuclear chain reaction. Consistent with the objective of blanketing, or otherwise covering and/or forming an encapsulating layer around the contaminating nuclear reaction, the neutron absorbing material may be provided in a variety of configurations. Most simply, the neutron absorbing material may be provided in a powder or granular configuration that will allow the neutron absorbing material to not only conform to varying geometries, but also to invade crevices, cracks, etc. Such powder or granular material may be airlifted and dropped onto a desired location and/or sprayed, as well as other suitable methods of transport and application. Additionally, the neutron absorbing material may be provided as a slurry or suspension, such as in a fluid or gel continuous phase that can be pumped under pressure to a desired location, as well as being airlifted or sprayed. Additional containment of a breached reactor, or similar open source or radioactive contamination, may be achieved by providing the neutron absorbing powder mixed into cement. The cement can then be air-dropped, sprayed, or pumped on the reactor. Such a method of delivery will have the effect of transporting the neutron absorbing material to the reactor site as a suspension, such as disclosed above, However, when the concrete sets it will further provide a rugged, sealed physical barrier, i.e. sarcophagus. However, the cured concrete will still contain the neutron absorbing material, therein providing containment of radioactive contamination that exceeds what would be experienced by concrete alone. According to a second exemplary embodiment, a method is provided for either controlling or containing general sources of radioactive contamination. For example, it is thought that it would be difficult for small terrorist groups to construct full scale nuclear fission or fusion weapons. Unfortunately, it is believed that it is much more within the capacity of such groups to procure radioactive materials that may be combined with conventional explosives. The result is a so called “dirty bomb” that, when detonated, will spread the radioactive material over a wide area producing potentially serious radioactive contamination. In order to minimize the damage and health hazards at the point of attack or contamination, neutron absorbing materials can be used in a similar manner as the first embodiment. Specifically, neutron absorbing material as a powder, granule, slurry, suspension, paste, etc. may be applied to the contamination site. Desirably, neutron absorbing material may be employed at the contamination site before elements which can become activated and radioactive by neutron bombardment experience significant radiation exposure. For example, the cobalt in stainless steels when bombarded by neutrons can become radioactive, after which it may emit penetrating gamma radiation for tens if not thousands of years. Thus, to minimize health hazards after a dirty bomb attack, the contaminated area/site could be coated with neutron absorber materials which will absorb any neutrons prior to these neutrons contacting living flesh or materials which can be activated. Consistent with the desire to provide rapid control or containment of a radioactive contaminated site, neutron absorbing material may be provided as a powder, granule, slurry, suspension, etc., as discussed above. Furthermore, the neutron absorbing material may be provided in an adhesive or high viscosity matrix allowing application to vertical or inverted surface. In such forms, the neutron absorbing material may conveniently be provided in man-portable dispensers. Such dispensers may be similar to fire extinguishers in size and operation. Accordingly, the man portable dispensers may be provided to emergency response workers, increasing the response time and decreasing the damage inflicted. In a precautionary application according to further exemplary embodiments of the present invention, neutron absorbing materials may be incorporated into pre-positioned structures to provide protection against radiation exposure. For example, in the event of a nuclear explosion or widespread attack with dirty bombs, it may not be possible to move large numbers of people out of the contaminated area in a short time. Thus, for strategic planning, appropriate bomb shelters or safe areas may be set up in advance of any attack. Additionally, in some cases, civilians may be interested in providing additional protection to their homes or their personal bomb shelters. As mentioned previously, the most dangerous particle produced in a nuclear attack is highly penetrating neutrons. Thus, it may be especially important to protect civilians against the harmful effects of neutron bombardment. According to this aspect of the present invention, neutron absorbing materials may be integrated with construction items. According to one example, powder or granular neutron absorbing material may be mixed in concrete. Any structures, walls, foundations, etc. produced from the concrete will contain the neutron absorbing material, and will eliminate or reduce neutron penetration into the structure. Similarly, neutron absorbing material may be incorporated into brick or cinderblock during the manufacture thereof to achieve the same result. According to associated embodiments, neutron absorbing powder, granules, etc. may be incorporated into drywall. When used in construction, the neutron absorbing drywall will serve to reduce the passage of neutrons through the walls of a structure. In yet another exemplary embodiment, fine neutron absorbing particles may be mixed into paint prior to application. The paint may then be applied to surfaces in a conventional manner. By combining neutron absorber powders with construction materials as described above, houses, dwellings, and/or shelters may be produced with increased levels of survivability. Referring to Table 1, exemplary neutron absorbing elements are listed along with the associated absorption cross-sections for 2200 m/s neutrons. Absorption cross-section (Thermal Neutron Cross Section) refers to the apparent or effective area presented by a target particle to an oncoming particle, or electromagnetic wave, and is a general measure of the probability of an interaction (absorption) between the target particle and the oncoming particle or electromagnetic wave. While all of the listed elements are especially effective for absorbing thermal neutrons (0 to 500 keV), the preferred embodiments the present invention utilize specific members of the lanthanide series having a cross section greater than 100 Barns, such as gadolinium, samarium, europium, erbium and dysprosium due to their high cross sections and relatively low cost. TABLE 1Information About Key Neutron Absorbing Elements/IsotopesElement/Thermal Neutron Cross SectionIsotopeAbundance(microscopic, Barns)BoronNatural761B-1019.9%3,840GadoliniumNatural48,800Gd-15320,000Gd-15514.8%61,000Gd-15715.65% 254,300Gd-16120,600CadmiumNatural2520SamariumNatural5922EuropiumNatural4530DysprosiumNatural994 Additionally, in some applications, it may be desirable to mix these elements with still other elements which are more effective for absorbing very high energy neutrons (1 to 15 MeV). Exemplary elements which are especially effective for absorbing high energy neutrons include hafnium, zirconium, tantalum, silver, indium, and hydrogen. These high energy neutron absorbing elements can be conveniently incorporated into a solid material as a hydride. Neutron absorbing materials consistent with the invention may contain low energy thermal neutron absorbing elements, high energy neutron absorbing elements or combinations of both. Additionally, since it does not matter how the neutron absorbing element is bound, it can also be used in a metal alloy or in a ceramic alloy. For example, gadolinium can be dissolved in specialized stainless steel alloys and then processed in to a powder form to produce a thermal neutron absorber. Alternately, gadolinium can be oxidized to form Gd2O3, and the oxide could then be used for thermal neutron absorption. The Gd2O3 may be an especially effective addition for paints which often incorporate oxide particles as pigments. It is noteworthy in providing neutron absorbing materials that the neutron absorbing effectiveness of a particular element is independent of how it may be chemically bonded, assuming there has been no density change. This is because neutron absorption occurs in the nucleus of the atom. Chemical bonding is a function of the electron interactions of the atoms, and so does not effect the neutron absorption. Thus, highly neutron absorbing elements can be mixed together in solids either dissolved in a host lattice or via precipitation out of the lattice through the formation of distinct phase(s). While the microscopic cross section is due to a particular element, the material containing neutron absorbing elements has a macroscopic cross section that is dependent upon the distribution of alloying elements based on a volumetric basis. While one needs to calculate the macroscopic cross section in each particular case, the magnitude and value will be consistent with the make-up of the individual elements, i.e., high macroscopic cross sections are easily obtained by alloying with high microscopic cross section elements. The present invention has been described above with reference to specific exemplary embodiments that are only intended to provide illustration and understanding. Accordingly, the invention should not be construes as being limited by the above description of exemplary embodiments, but rather only by the appended claims.
052788806	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS A pressurizer tank 22 as shown in FIG. 1, is a vertically elongated tank supported on a substructure 30 in a reactor containment building. A compartment 32 is provided to substantially enclose the pressurizer tank 22, and the compartment 32 is shown in the drawings rather than the full containment building. The compartment 32 can be defined, for example, by interior concrete walls 34 in the containment. As shown in FIGS. 2, 5 and 6, an upper support 40 for the pressurizer tank 22 is provided at a space from the bottom 44 of the tank 22, preferably adjacent the top 46 of the tank. The upper support 40 includes a circular box girder 50 surrounding the pressurizer tank 22, attached to the walls 34 of the compartment 32 by sway struts 54. The sway struts 54 couple any lateral forces on the girder 50 to the compartment walls 34, such that such lateral forces are not transmitted through the pressurizer tank 22 to the substructure 30 on which the pressurizer tank 22 is supported vertically. The sway struts 54 are shown in detail in FIGS. 2-4, and the girder 50 in FIGS. 5 and 6. The girder 50 can rest vertically upon valve support bracket plates 118 protruding radially from the outer wall of the pressurizer tank 22, as shown in FIG. 7. The sway struts 54 define substantially inextensible coupling members which are attached between the pressurizer tank 22 and the compartment walls at tie points defining a horizontal pivot axis, as discussed more fully hereinafter. The pressurizer tank 22 is susceptible to lateral forces due to the structures carried at the top and above the pressurizer tank. A relief/discharge valve support frame 60 is disposed over the pressurizer tank 22 for mounting such structures. The support frame 60 includes vertical legs 62 and horizontal beams 64, the legs and beams being welded together to form a structural frame. The legs 62 are rested on and welded at their lower ends to the upper surface of the box girder 50. The relief/discharge support frame 60 carries discharge piping 72 and valve module elements 74, through which the pressurizer tank 22 can be vented to an in-containment refueling water storage tank (not shown). The piping 72 and valve modules 74 define a weight which is spaced from the substructure 30 of the containment interior concrete which supports the pressurizer tank 22. It is not advantageous to attempt to support the piping 72 and valves 74 independently of the pressurizer tank 22, because the piping 72 is coupled to the tank 22, and the tank is subject to dimensional variations due to changes in pressure and pressure in the reactor coolant system. Accordingly, it is necessary to ensure that the mounting of the pressurizer tank is sufficiently strong to hold the tank 22 and its supported appendages against lateral displacement in the event of a seismic shock or other lateral force. The box girder 50 is restrained against lateral displacement relative to the compartment 32 by a plurality of opposed sway struts 54 extending from the girder 50 to the walls 34 of the structural compartment 32 in which the pressurizer tank 22 is located. In the example shown, the compartment 32 is approximately square in plan view and the sway struts 54 extend from the girder 50 between the corners 82 of the compartment 32 and a respective attachment to the girder 50 approximately midway between the corners 82, in opposed pairs. The sway struts 54 can be conventional supporting struts (such as Grinnell FIG. 211 or the like). Attachment at the corners 82 of the compartment 32 provides maximum stiffness and load bearing capacity. Referring to FIGS. 2-4, the sway struts 54 comprise horizontal load bearing shafts 92 terminating in pivot fittings 94. The distal member of each pivot fitting 94 can be welded or similarly fixed to the girder 50 and to a member of the compartment 32 such as a vertical beam at a corner 82 as shown. The pivot fittings 94 comprise backplates 102 shaped substantially as U-brackets opening toward the shaft 92 of the respective sway strut 54. Connecting pins 104 extend through the backplates 102 and through an end of each sway strut shaft 92, the pins 104 each defining a horizontal pivot axis in the connection between the box girder 50 and the compartment wall 34. At the box girder 50, two vertically spaced attachment plates 108 having side edges 112 aligned substantially parallel to the respective sway strut pivot axis are welded to the outer periphery of the box girder 50 and provide a point of attachment for the backplates 102 of the sway strut fittings 94. The sway strut shafts 92 can include self-aligning bushings such that each end of the shafts remains aligned with the pivot axis defined by the connecting pin 104 extending through the backplate 102. FIGS. 5 and 6 show the box girder 50 in greater detail. The attachment plates 108 as shown can comprise two horizontal plates, spaced vertically and welded on the surface of the box girder 50. It is also possible to use a channel member or a solid block to define a point of attachment for the sway strut backplates 102. As also shown in FIG. 6, a plurality of coupling plates 114 are also welded or similarly attached to the girder 50. The coupling plates 114 protrude axially downwardly from girder 50 and are oriented along radii of the tank 22, for example at intervals of 45.degree.. These coupling plates 114 are positioned and dimensioned to extend between pairs of complementary coupling plates 118 which are welded onto the outer surface of the pressurizer tank 22. This form of connection is shown in FIG. 7. The coupling plates 114, 118 attached respectively to the girder 50 and to the pressurizer tank 22 have alignable holes 122 for receiving locking pins 124, thereby engaging the box girder 50 with its attached supporting frame 60 vertically on the pressurizer tank 22. The support frame carrying 60 the relief/discharge valve and conduit system is welded at the top of the box girder 50 encircling the pressurizer tank 22. As shown in FIG. 7, the relief/discharge valves 74 and conduits 72 can comprise a multi-tiered arrangement, for example enabling staged depressurization of the depressurizer tank 22 through conduits 72 of different diameters, each having one or more valves 74 which open a flow path to a discharge, for example, in an in-containment refueling water supply. FIG. 9 generally illustrates the flowpath couplings according to such an arrangement, with a nuclear reactor vessel 162 coupled in a coolant circuit with a steam generator 164, and also a high pressure makeup tank 166 and an atmospheric pressure refueling water supply tank 168. Pressurizer 22 controls the pressure of the coolant circuit through check valve 172, and also is coupled via depressurization valves 74 to discharge into the refueling water supply 168. The respective valves 74 can be operated successively to obtain staged depressurization. At least a first level 132 of the relief/discharge support frame 60 is insulated in order to reduce thermal stresses due to differences in temperature between the relief/discharge support frame 60 and the piping 72 associated with the relief/discharge system. During automatic staged depressurization of the reactor coolant system, different valves 74 are opened in sequence, thereby subjecting selected subsets of the conduits 72 to thermal loading. The girder 50 is not insulated in order to preclude high thermal loading on the sway struts 54. As mentioned above, the girder 50 is preferably attached to the tank 22 vertically via mating coupling plates 114, 118 attached respectively to the pressurizer tank 22 and the girder 50, held together by pins 124. Valve supporting brackets are welded on the pressurizer tank surface, and a bracket matching such a valve support bracket can be welded to the bottom of girder 50 to mate with the valve supporting brackets. Either a slot in the girder bracket 114 is custom machined at the correct location, or an oversized bracket 114 with a slot already provided is custom fit and welded in place to properly fit the pressurizer valve support bracket in that case. Of course, it is readily possible to use two brackets on either of the girder 50 and the pressurizer 22 to mate with one bracket on the other, or to attach single plates to each of the girder 50 and the tank 22, e.g., using bolts. It is desirable to leave a small gap between the girder brackets 114 and the valve brackets 118, which helps to prevent weak axis bending loads on the brackets 112, 114. Strip shims 136 are located between the pressurizer outer surface and the girder 50 with an appropriate hot gap 138 (e.g., 1/32 inch or 0.8 mm). Strip shims 136 provide the necessary lateral load transfer between the girder 50 and the pressurizer tank 22 while minimizing the heat transfer path. Since the relief/discharge support frame 60 is welded to the girder 50 and the girder is fastened to the building structure by the sway struts 54, all relief/discharge system lateral loads are borne directly by the compartment walls 34 rather than passing wholly or partly through the pressurizer 22 and then to the compartment 32 or to the substructure 30. The girder 50 can accommodate upward and downward displacement relative to the compartment walls 34 as the tank 22 enlarges and contracts during plant heatup and cooldown. The pivotal couplings of the sway struts 54, and their self aligning bushings, permit vertical pivoting of the sway struts 54 around their connecting pins 104, which allows some vertical displacement of the girder 50 (and tank 22) relative to the compartment walls 34. Insofar as the girder 50 moves up or down, the sway bars 54 still can accommodate the relative displacement while bearing any lateral forces which may occur. The sway struts 54 as shown are arranged in pairs aligned substantially along tangents to the wall of tank 22, which is cylindrical. Although four pairs of sway struts 54 are preferred at 90.degree. intervals, it would be possible to use a larger or smaller number to restrain the pressurizer tank 22 and the upper support 40 thereon. For example, a 120.degree. arrangement using three pairs of struts 54 could be employed to obtain a triangular configuration in which the struts 54 would still be substantially tangential to the girder 50 and arranged in opposed sets to bear lateral loading in any direction. The pressurizer tank 22 has a top dome 142, attached to a tubular body 144 at a weld 146. Preferably the girder 50 is disposed below the weld 146 in order to facilitate inspection of the weld, e.g. about 9 inches (23 cm) below the weld 146. The support according to the invention provides an improved restraint for the pressurizer tank 22, and eliminates the need for seismic lugs, for example of the type shown by lugs 152 in FIG. 10 and identified as prior art. All lateral forces according to the invention are coupled to the compartment walls 34, preferably at the corners 82, for maximum stiffness and load bearing capacity. The invention having been disclosed, certain variations within the scope of the inventive concept will now be apparent to persons skilled in the art. The invention is intended to encompass a reasonable range of equivalents, and accordingly, reference should be made to the appended claims rather than the foregoing specification in order to assess the scope of exclusive rights in the invention claimed.
	description	This application claims the benefit of and is a continuation-in-part of U.S. Ser. No. 15/427,414 which was filed on Feb. 8, 2017, the disclosure of which is incorporated herein in its entirety. This disclosure includes a customizable x-ray shielding system for primary use with an X-ray producing gantry. As used herein, the term “gantry” includes the ‘donut’ shaped part of a CT scanner that houses the components necessary to produce and detect x-rays to create a CT image. Typically, the x-ray tube and detectors are positioned opposite each other and rotate around the gantry aperture. Among the art considered in preparing this patent application are these references: U.S. Pat. Nos. 5,099,135; 5,417,225; 6,448,571; and 6,653,648. Disclosed is a shielding system for customized protection from X-rays. The shielding system is mounted on an x-ray producing gantry. The gantry, if desired, can be tilted so that it may lie in a plane that is about, for example, plus or minus 45 degrees from a vertical plane. In use over the various orientations of such machines there is a need to isolate the technician or physician from X-radiation. To do this, shielding devices such as lead or other radio-opaque curtains are often used. But such curtains need to be suspended from a movable rail so that their placement is predictable, yet adjustable regardless of machine orientation. One way to achieve the goal of reliably supporting such shielding devices is to suspend them from a rail that is mounted in a foundational block which is affixed preferably to an upright or other face of the movable gantry. Optionally the block can be secured to a gantry-mounted track that enables the block's position to be further adjusted. If desired, multiple foundational blocks may be affixed to the gantry. Facing an upright, usually vertical plane of the gantry is a back surface (A) of each foundational block. A post-receiving aperture extends between the top surface (C) and an opposing bottom surface (D). Each aperture is configured to receive a proximal end region of an arcuately movable rail from which, for example, a radio-opaque curtain can be suspended if desired under the influence of gravity in a vertical plane. One or more detents are defined by the top surface (C). At least some of the detents are configured to be in registration with one or more lugs that extend radially from a proximal end region of the bracket. Upon registration, there is little or no twisting movement of the rail in relation to the foundational block. Extending from the proximal section of the rail is a distal section. In use, a rail can be turned within an associated aperture. Regardless of gantry orientation, one or more radio-opaque curtains can be suspended from a given rail in a desired position from the distal section by twisting the proximal end region, rotating it in relation to the associated aperture and then seating one or more lugs in a suitable detent. As required, detailed embodiments of the present invention are disclosed herein; however, it is to be understood that the disclosed embodiments are merely exemplary of the invention that may be embodied in various and alternative forms. The figures are not necessarily to scale; some features may be exaggerated or minimized to show details of particular components. Therefore, specific structural and functional details disclosed herein are not to be interpreted as limiting, but merely as a representative basis for teaching one skilled in the art to variously employ the present invention. Disclosed is a shielding system 10 for use in alternative configurations as user-selected, customized shielding from X-rays. The shielding system 10 is mounted on a base such as a gantry 12. In use, the usually near-vertical plane of the gantry 12 can be tilted if desired through an angular range of about 45 degrees from a vertical plane. In the shielding system 10 associated with a given gantry, there are one or more foundational blocks 26 (see, e.g. FIG. 1). Facing the block-mounting mounting surface 14 of the gantry 12 is a back surface (A) 28 of the foundational block (FIG. 3). A post-receiving aperture 32 extends between the top surface (C) 28 (FIG. 2) and an opposing bottom surface (D) 30 (FIG. 3). One or more detents 38 are defined within the top surface (C). Each detent 38 is configured to be in registration with a lug or pin that extends generally radially from the rail 42. Upon registration of the rail 42 within an associated detent 38, there is little or no twisting movement of the rail 42 in relation to the foundational block 26. One or more apertures 32 extend between the bottom surface (D) 28 and the opposing top surface (C) 30. Each major aperture is configured to receive a rail 42. Each rail 42 has a proximal section 58 (FIG. 1) that is received by a major aperture 32 of a foundational block 26. Extending from the proximal section 58 is a distal section 60. It will be appreciated that the distal section 60 need not be orthogonal with the proximal section 58. Other angular relationships between about 30 and 90 degrees may be desirable, depending on the installation. Regardless of gantry orientation, one or more radio-opaque curtains 62 can be hung from the distal section 60 to allow flexibility in adapting to operator and patient examination requirements. It will be appreciated that the curtains 62 are preferably formed from an x-ray absorbing material. Such curtains 62 may for example be 20″ long×26″ wide. But the curtains 62 can be of any length and width. Further, the curtains 62 may have any desired x-ray attenuation characteristics. In one embodiment, there is a knob 64 that is received at an end of the distal section 60 of the rail 42 for constraining lateral movement of the one or more radio-opaque curtains 62 along the distal section 60 of the rail 42. Preferably the apertures 32 terminate at the opposing top surface (C) 28 (FIG. 2) and slots 38 extend radially from the apertures 32. The slots 32 are configured to engage lugs 68 that extend radially from the proximal section 58 of the rail 42 to preclude a twisting motion of the rail 42 when seated within an associated aperture 32. This feature influences rail positioning regardless of gantry orientation. In most uses, a rail 42 can be twisted within an associated major aperture 32 so that the distal section 60 of the rail 42 can be made to extend substantially horizontally regardless of gantry orientation. Optionally, the rail position is secured by engaging a lug 68 within a slot 38. It will be appreciated that the distal section 60 of the rail 42 may be arcuately positioned so that the distal section 60 extends horizontally. Optionally, if desired, the distal section 60 may lie in parallel with a face of the gantry. In some cases, each gantry has one or two foundational blocks 26 for optimized shielding of the patient and operator from radiation. Each block 26 supports a rail 42. Preferably there are up to four blocks 26 per gantry. Placement of the block 26 is not restricted to a front face of the gantry. For example, one pair of blocks 26 may be positioned behind another pair of blocks so that one curtain 62 may be positioned partially behind another curtain. To secure a foundational block 26 in relation to the gantry, means for securement are provided. Such securement means include, for example, a threaded bolt, a screw, a rivet, cement or a glue. In some cases, the gantry orientation lies between plus and minus 45 degrees from a vertical plane. Although the foundation blocks 26 are depicted as generally brick-shaped, it will be appreciated that such blocks may be shaped in various geometries and sizes, and have non-parallel faces that may or may not be planar. 10 shielding system 12 gantry 26 foundational blocks 28 back (gantry-facing) surface (A) 30 bottom surface (D) 32 post-receiving aperture 38 detent 42 rail 58 proximal section of rail 60 distal section 62 radio-opaque curtain 64 knob 68 lugs While exemplary embodiments are described above, it is not intended that these embodiments describe all possible forms of the invention. Rather, the words used in the specification are words of description rather than limitation, and it is understood that various changes may be made without departing from the spirit and scope of the invention. Additionally, the features of various implementing embodiments may be combined to form further embodiments of the invention.
	description	1. Field of the Invention The present invention relates to an improvement of output power of an X-ray generation tube, an X-ray generation device, and an X-ray imaging system. 2. Description of the Related Art An X-ray generation tube is an X-ray source that is used for an X-ray generation device in an application of a medical diagnosis or a nondestructive inspection such as a foreign matter inspection. The X-ray generation tube includes an electron gun for emitting an electron beam, an anode for accelerating electrons, and a target for generating an X-ray by a collision of the electrons. The target is electrically connected to the anode. It is known that, in order to obtain a predetermined analysis resolution, the X-ray generation device includes a grid electrode having an electrostatic lens action for a purpose of decreasing a focus diameter of the electron beam to be formed on the target. Japanese Patent Application Laid-Open No. 2011-81930 describes an X-ray generation device including a lens electrode for focusing the electron beam, which is disposed between an electron emitting portion and the target. On the other hand, when the X-ray generation tube outputs an X-ray, approximately 1% of kinetic energy of electrons included in the electron beam irradiating the target is used for the X-ray, and most of the input energy is converted into thermal energy, with the result that a temperature of the target rises. The target of the X-ray generation tube is irradiated with the focused electron beam, and hence is apt to be thermally damaged at a region where current density distribution of the electron beam is largest. Japanese Examined Utility Model Application Publication No. H04-3384 discloses a method of reducing thermal damage to the focusing center portion of the target by devising a cathode structure. Japanese Examined Utility Model Application Publication No. H04-3384 discloses the method in which a cathode filament is formed into a spiral shape, and an end of the filament is positioned in the center portion of the spiral filament, so as to decrease the temperature at the center portion of the electron beam, and to reduce the current density of the electron beam of emitted thermal electrons. It is an object of the present invention to provide an X-ray generation tube capable of reducing thermal damage to the target and emitting an X-ray with high output intensity. Further, it is another object of the present invention to provide an X-ray generation device and an X-ray imaging system, in which a target has high life characteristics and output power is high. According to one embodiment of the present invention, there is provided an X-ray generation tube, including: a target for generating an X-ray through irradiation with an electron beam; an electron source provided opposed to the target; and a grid electrode having multiple electron passage apertures, in which the grid electrode is disposed between the target and the electron source so that a part of a source-side electron beam emitted from the electron source passes through the multiple electron passage apertures and irradiates the target, in which the source-side electron beam has a current density distribution, in which the grid electrode has an aperture ratio distribution, and in which a region of the source-side electron beam in which a current density is largest is aligned with a region of the grid electrode in which an aperture ratio is smallest. Further, according to one embodiment of the present invention, there is provided an X-ray generation device, including: the X-ray generation tube of one embodiment of the present invention; a tube voltage circuit to be electrically connected to each of the target and the electron source, so as to output a tube voltage to be applied between the target and the electron source; and a grid potential circuit for defining a voltage between the grid electrode and the target. Further, according to one embodiment of the present invention, there is provided an X-ray imaging system, including: the X-ray generation device of one embodiment of the present invention; and an X-ray detector for detecting an X-ray which is emitted from the X-ray generation device and passes through an object. Further features of the present invention will become apparent from the following description of exemplary embodiments with reference to the attached drawings. Embodiments of the present invention are described with reference to the drawings. FIG. 5A illustrates a structure of a related-art X-ray generation tube as a reference example. An X-ray generation tube 200 of this reference example includes an electron source 201 having a planate electron emitting portion. Only a layout of members necessary for comparison with the embodiment of the present invention is schematically illustrated. In FIG. 5A, an insulation tube constituting a barrel portion of the X-ray generation tube of this reference example is omitted. The electron source 201 generates a source-side electron beam 230 based on an extraction electrode 202, and a grid potential applied to a grid electrode 210 electrically connected to the extraction electrode 202. Electrons included in the generated source-side electron beam 230 are accelerated by an acceleration electric field formed by a tube voltage applied between an anode 204 and the electron source 201 so as to irradiate a target 205. A part of the source-side electron beam 230 irradiates the grid electrode 210, and the electron beam, which passes through multiple electron passage apertures 211 of the grid electrode 210, is focused by a focusing electrode 203 and irradiates the target 205 as a target-side electron beam 231. As a result, a focal point is formed on the target 205 in an area irradiated with the target-side electron beam 231, and hence an X-ray is emitted from the focal point. The grid electrode 210 of this reference example has the multiple electron passage apertures 211 formed in a uniform arrangement pattern as illustrated in FIG. 5B. In this specification, the “uniform arrangement” used in terms of an arrangement of the electron passage apertures means that an aperture ratio distribution of the electron passage apertures is uniform. The aperture ratio distribution is defined by at least one of an aperture area of the electron passage apertures or electron passage aperture arrangement density. The center portions of the electrodes constituting the X-ray generation tube 200 are disposed so as to align with a center axis 206 of the electron beam as illustrated in FIG. 5A. In the X-ray generation tube 200 of this reference example including the grid electrode 210 having the multiple electron passage apertures with a uniform aperture ratio, when a predetermined grid potential is applied to the grid electrode 210, a focal point corresponding to a current density distribution as Gaussian distribution illustrated in FIG. 2A is formed on the target 205. The current density distribution on the target 205 illustrated in FIG. 2A has the largest current density at the center axis 206. In this case, an electron beam center portion (hereinafter referred to as a focal point center) as an intersection of the target 205 and the center axis 206 has the highest temperature on the target 205. The fact that the source-side electron beam 230 has the current density distribution means that the electron beam has an irradiation density distribution on the target in a beam diameter direction of the source-side electron beam 230. Therefore, in the related-art X-ray generation tube 200, it is necessary to set a current upper limit of the electron beam irradiating the target 205 within a range that does not reach a heat resistance limit at the focal point center. As a method of enhancing the current upper limit of the electron beam irradiating the target 205, there are a method of improving heat resistance and heat dissipation of the anode 204 and the target 205, a method of reducing current density at the focal point center on the target 205, and the like. In the method of reducing the current density at the focal point center, because a thermal load at the focal point center is reduced, it is possible to enhance the upper limit of input energy higher than the conventional value. As one of the methods for the reduction, there is a method of increasing a focal point diameter. However, if the focal point diameter is increased, an imaging resolution is deteriorated. Therefore, it is an important point for obtaining higher output power of the X-ray to reduce the current density at the focal point center on the target 205 without increasing the focal point diameter, so as to increase the upper limit of the energy input to the target 205. For instance, the method described in Japanese Examined Utility Model Application Publication No. H04-3384 has a limitation in a shape of the electron emitting portion, a limitation of generation of a current density distribution in the electron beam, and the like. Therefore, it has been desired to provide a method of reducing thermal damage to the target so as to realize higher output power also in a cold cathode, an impregnated hot cathode, and the like, which are not a filament type. FIGS. 1A and 1B are diagrams illustrating an X-ray generation tube 100 according to a first embodiment of the present invention, and are schematic diagrams illustrating a layout of components necessary for describing the present invention. In FIG. 1A, an insulation tube constituting a barrel portion of the X-ray generation tube of this embodiment is omitted. An electron source 101 and an anode 104 constituting the X-ray generation tube 100 are fixed to the insulation tube (not shown). In addition, the X-ray generation tube of this embodiment has a transmission type structure in which the X-ray is extracted from a surface opposed to an electron incident surface of a target 105, but the present invention can be applied to a reflection type X-ray generation tube. In the grid electrode 210 of the X-ray generation tube 200 of the reference example described above, the electron passage apertures 211 are uniformly arranged as illustrated in FIG. 5B. Therefore, there is no difference of the aperture ratio between the center portion including an intersection of the grid electrode 210 and the center axis 206 of the electron beam and its peripheral part. In contrast, a grid electrode 110 of this embodiment has no electron passage aperture 111 at the center portion including an intersection of the grid electrode 110 and a center axis 106 of a source-side electron beam 130. In other words, the grid electrode 110 has a structure having an aperture ratio distribution in which the center portion including the intersection of the grid electrode 110 and the center axis 106 of the source-side electron beam 130 has a smaller aperture ratio than its peripheral part. The grid electrode 110 can be divided into, based on the aperture ratio distribution, an outer region 114 provided between an outer region outer periphery 116 and an inner region outer periphery 117, and an inner region 115 surrounded by the inner region outer periphery 117. The center portion including the intersection of the grid electrode 110 and the center axis 106 of the source-side electron beam 130 corresponds to the inner region 115 surrounded by the inner region outer periphery 117 of the grid electrode 110. The source-side electron beam 130 emitted from the electron source 101 has a current density distribution in the beam diameter direction of the electron beam. In this specification, the fact that the source-side electron beam 130 has a current density distribution means that the electron beam has an irradiation density distribution on the target in the beam diameter direction of the source-side electron beam 130. In the X-ray generation tube 100 of this embodiment, a region in which the source-side electron beam 130 has the largest current density is aligned with the center axis 106 of the electron beam. The grid electrode 110 has multiple electron passage apertures with an aperture ratio distribution as illustrated in FIG. 1B on a surface opposed to the electron source 101. In this specification, the fact that the grid electrode 110 has the aperture ratio distribution means that the electron beam has the aperture ratio distribution in the beam diameter direction of the source-side electron beam 130. The center axis 106 at which the source-side electron beam 130 has the largest current density is positioned so as to align with the region 115 in which the grid electrode 110 has the smallest aperture ratio. When appropriate potentials are applied to the electrodes constituting the X-ray generation tube 100 of this embodiment, there is formed a current density distribution on the target 105, which corresponds to electron irradiation density as shown in FIG. 2B by a broken line. A top hat type distribution shown in FIG. 2B corresponds to a current density distribution of a target-side electron beam 131 irradiating the target 105. In this specification, the focal point is defined as a region having a current density of 15% or higher of the largest value in the current density distribution of the target-side electron beam 131. A focal point diameter φb is a width in the beam diameter direction corresponding to a region having a current density of 15% or higher of the largest value in the current density distribution of the target-side electron beam 131. If a shape of the focal point of the target-side electron beam 131 is a circle centered at the center axis 106, the focal point diameter φb is matched with a diameter of the circle. In the graphs shown in FIGS. 2A and 2B, the vertical axis represents the current density of the target-side electron beam 131, and the horizontal axis represents a position from the focal point center on the electron incident surface side of the target 105. A conversion efficiency of irradiation current to X-ray intensity on the target 105 is substantially constant regardless of a position on the target 105. Therefore, intensity distributions of the X-rays emitted from the X-ray generation tube 200 and the X-ray generation tube 100 with respect to a position on the target are similar to the current density distributions shown in FIGS. 2A and 2B. The X-ray intensity distribution can be measured as follows. A pinhole mask (not shown) having a predetermined aperture is disposed 10 cm ahead of an X-ray emission window, and an X-ray detector (not shown) including X-ray detection elements arranged in a two-dimensional array is disposed 40 cm further ahead of the pinhole mask. Here, with reference to the normal to the X-ray emission window, intensity detected by the X-ray detector is recorded every time when a position of the pinhole mask is changed, to thereby obtain the X-ray intensity distribution. In this specification, a range indicating intensity of 15% or higher of the largest intensity value of the target-side electron beam 131 is defined as the focal point of the electron beam. Further, φa in FIG. 2A and φb in FIG. 2B correspond to focal point diameters of the X-ray generation tube 200 and the X-ray generation tube 100, respectively. According to the present invention, because the grid electrode 110 has the electron passage apertures 111 with the aperture ratio distribution, the target side current density distribution is changed from a normal distribution profile to a so-called top hat type profile. As a result, the X-ray generation tube of the present invention can secure heat resistance of the target 105 in the focal point center region and can enhance output intensity of the target 105. More specifically, the X-ray generation tube 100 of the present invention has a feature in that the grid electrode 110 has the multiple electron passage apertures 111 arranged so that the region having the largest current density of the source-side electron beam 130 is aligned with the region having the smallest aperture ratio of the grid electrode 110. The grid electrode 110 has a smaller aperture ratio in the center portion including the intersection of the grid electrode 110 and the center axis 106 of the source-side electron beam 130 than in its periphery, and hence the energy input to the target 105 becomes smaller as well. The electron passage apertures 111 of the grid electrode 110 only need to be arranged so that the aperture ratio in the center portion at which the grid electrode 110 crosses the center axis 106 of the source-side electron beam 130 is smaller than the aperture ratio in its peripheral part. The aperture ratio distribution of the grid electrode 110 can be formed by at least one of a surface density distribution of the electron passage apertures 111 or an aperture area distribution of the electron passage apertures 111. FIGS. 3A, 3B, 3C and 3D illustrate modified examples of the grid electrode 110 of the embodiment illustrated in FIGS. 1A and 1B. The embodiment illustrated in FIG. 3A is different from the embodiment illustrated in FIG. 1B in that the outer region 114 is defined by the outer region outer periphery 116 having a regular hexagonal shape. The grid electrode 110 of this embodiment has an aperture ratio distribution defined by the surface density distribution of the electron passage apertures 111 similarly to the embodiment illustrated in FIG. 1B, and has the inner region 115 having a lower aperture ratio than the outer region 114. The embodiment illustrated in FIG. 3B is different from the embodiment illustrated in FIG. 1B in that the inner region 115 has electron passage apertures 113 having a smaller aperture area than electron passage apertures 112 in the outer region 114. The grid electrode 110 of this embodiment has an aperture ratio distribution defined by the aperture area distribution of the electron passage apertures 111, and has the inner region 115 having a lower aperture ratio than the outer region 114. Further, the grid 110 of this embodiment is different from the embodiment illustrated in FIG. 1B also in that the outer region 114 is provided between the outer region outer periphery 116 and the inner region outer periphery 117 each having a substantially regular square shape. Each of the embodiments illustrated in FIGS. 3C and 3D is different from the embodiment illustrated in FIG. 1B in that the inner region 115 has an aperture ratio distribution defined by the surface density and aperture area distribution of the electron passage apertures 111 with respect to the outer region 114 in a circumferential direction and in a diameter direction of the grid 110. The grid electrode 110 of each of the embodiments illustrated in FIGS. 3C and 3D, in which the aperture ratio distribution is formed based on the surface density and aperture area distribution of the electron passage apertures 111, is preferred from the viewpoint of controllability of beam formation because continuous aperture ratio distribution can be formed. As the electron source 101, there is used an electron source whose electron emission amount can be controlled from the outside of the X-ray generation tube 100. An electron source that can be used for the electron source 101 is selected from the group consisting of hot cathodes of a metal filament type, an oxide filament type, and an impregnated type, and cold cathodes of a carbon nanotube type, a spindt type, and an MIM type. The embodiment of using the impregnated hot cathode for the electron source 101 is preferred from the viewpoint of symmetry or uniformity in an electron emission surface having a predetermined area. An extraction electrode 102 is an intermediate electrode disposed for forming an electric field in front of the electron emission source so as to control an electron emission amount from the electron source 101. The extraction electrode 102 of this embodiment is electrically connected to the grid electrode 110. The grid electrode 110 is an intermediate electrode disposed in front of the electron emission source 101 so as to form a beam profile of the source-side electron beam emitted from the electron source 101. It is not necessary that the grid electrode 110 and the extraction electrode 102 have the same potential, and the grid electrode 110 and the extraction electrode 102 may be connected to different voltage sources, respectively. The grid electrode 110 may be disposed at any position between the electron source 101 and the target 105, but the grid electrode 110 is preferred to be disposed in a vicinity of the extraction electrode 102 considering heat generation by the irradiating electrons. In the present invention, the grid electrode having the multiple electron passage apertures with the aperture ratio distribution may be disposed as a focusing lens electrode. The target 105 included in the X-ray generation tube 100 illustrated in FIG. 1A is a transmission type target. The transmission type target employs a self-supported thin film structure such as gold foil, or a lamination structure in which a target layer is disposed on the surface irradiated with the target-side electron beam 131, and the target layer is supported by a transmissive substrate made of a transmissive material. The transmission type target having the lamination structure includes the target layer disposed on the side of the transmissive substrate opposed to the grid electrode. The transmissive substrate of the transmission type target having the lamination structure is made of a light element having a smaller atomic number at least than a target metal contained in the target layer. For the transmissive substrate, beryllium, silicon carbide, diamond, or the like is used. The transmissive substrate made of diamond (hereinafter referred to as a diamond substrate) is particularly preferred as the transmissive substrate of the transmission type target for its physical properties such as a high X-ray transmittance (atomic number of 6 and low specific gravity), a high thermal conductivity (1,600 W/(mK)), and a high heat resistance (decomposition temperature of 1,200° C. or higher). For the diamond substrate, there is used single crystal diamond obtained by a high temperature high pressure synthetic method or the like, or polycrystal diamond obtained by a sintering method, a chemical vapor deposition method, or the like with use of microcrystal diamond as raw material. An outer shape of the transmissive substrate is a flat plate having a surface and an opposed surface. For instance, a rectangular parallelepiped shape or a disc shape is adopted. As to a disc transmissive substrate, by setting the diameter to 2 mm or more to 10 mm or less, it is possible to dispose a target layer on which a necessary focal point diameter can be formed. By setting the thickness of the diamond substrate to 0.5 mm or more to 2 mm or less, it is possible to secure radiation transmittance. If the diamond substrate having a rectangular parallelepiped shape is used, the above-mentioned range of the diameter may be read as a range of a short side length and a long side length of a surface of the rectangular parallelepiped. The target layer contains a metal element having a high atomic number, a high melting point, and a high specific gravity, and contains at least a metal selected from, for example, the group consisting of tantalum, tungsten, molybdenum, silver, gold, and rhenium. From the viewpoint of compatibility with the transmissive substrate, it is more preferred that the target layer contain at least one kind of a metal selected from the group consisting of tantalum, molybdenum, and tungsten in which the standard free energy of formation of the carbide thereof becomes negative. In addition, the target layer may be formed of a pure metal having a single composition or an alloy composition, or may be formed of a metal compound such as a carbide, nitride, or oxynitride of the metal. The thickness of the target layer is selected from the range of 1 μm or more to 12 μm or less. A lower limit and an upper limit of the thickness of the target layer are determined based on the viewpoints of securing the X-ray output intensity and reducing an interface stress, respectively. It is more preferred that the thickness be set within a range of 3 μm or more to 9 μm or less. For the purpose of constituting a part of the anode 104 of the X-ray generation tube 100, an anode member, a brazing filler metal, and a conductive electrode (not shown) are electrically connected to the target 105. The conductive electrode is a conductive member disposed as necessary for securing electric connection with the anode member. For the conductive electrode, a metal such as tin, silver, or copper, a metal oxide, or the like is used. The brazing filler metal has a function of supporting the target 105 with respect to the anode member as well as a function of electrically connecting the target layer and the anode member to each other. The brazing filler metal is an alloy containing gold, silver, copper, tin, and the like. The composition of the alloy is appropriately selected in accordance with the member to be bonded so that adhesiveness between the different types of materials such as the transmissive substrate, the conductive electrode, and the anode member can be secured. Next, in view of suppressing thermal damage at the focal point center of the target 105, a more preferred layout of the grid electrode 110 is described with reference to FIG. 4. FIG. 4 schematically illustrates an embodiment in which the electron beam is focused in a path from the electron source 101 to the target 105. The X-ray generation tube 100 can adjust the focal point diameter of the electron beam emitted from the electron source 101 by a focusing lens 120 disposed between the grid electrode 110 and the target 105. The focusing lens 120 of FIG. 4 may be constituted of one focusing lens electrode, or may be constituted of multiple electrodes. FIG. 4 illustrates a main part of the X-ray generation tube 100 forming an electron optical system, which has a crossover 124 at which the beam diameter of the target-side electron beam 131 becomes smallest. FIG. 4 illustrates multiple electronic optical virtual points, which are determined uniquely by the focusing lens 120 and its electrostatic parameters. This embodiment is different from the embodiment illustrated in FIGS. 1A and 1B at least in having the crossover 124. Further, the electrostatic parameters of the focusing lens 120 include positions, shapes, and potentials of electrodes constituting the focusing lens 120. The shape of the electrode of the focusing lens 120 includes a shape of the electron passage aperture, and a thickness of the electrode. The focusing lens 120 has a lens focal point at each of the front and the rear of the focusing lens 120, which corresponds to an image formation point of an object virtually disposed at infinity. In this specification, a lens focal point positioned on the electron source 101 side is referred to as a rear lens focal point 123a, and a lens focal point positioned on the target 105 side is referred to as a front lens focal point 123b. A point conjugate to the crossover 124 with respect to the focusing lens 120 is uniquely defined on the electron source 101 side. This point conjugate to the crossover 124 is referred to as a crossover conjugate point 122. In the same manner, a point conjugate to the focal point center on the target 105 with respect to the focusing lens 120 is uniquely defined on the electron source 101 side. This point conjugate to the focal point center is referred to as a focus center conjugate point 121. The crossover conjugate point 122 corresponds to a virtually indicated cathode position and is generally called a virtual cathode. By appropriately setting the electrostatic parameters of the focusing lens 120, it is possible to form the crossover 124 at which the beam diameter of the target-side electron beam 131 becomes smallest between the front lens focal point 123b and the target 105 as illustrated in FIG. 4. In this embodiment, concerning a distance from the focusing lens 120, the crossover conjugate point 122 is farther than the focus center conjugate point 121. In addition, the grid electrode 110 is disposed between the crossover conjugate point 122 and the focus center conjugate point 121. It is preferred to dispose the grid electrode 110 having the aperture ratio distribution close to the position of the focus center conjugate point 121. The reason is because, with this layout, an aperture shape of the grid electrode 110 projected to the target 105 can be formed with high reproducibility, and hence the current density at the focal point center on the target 105 is reduced. Therefore, it is more preferred to dispose the grid electrode 110 having the aperture ratio distribution so as to align with the position of the focus center conjugate point 121 in order to reduce the current density at the focal point center on the target 105. Further, the virtual points such as the crossover 124, the crossover conjugate point 122, and the focus center conjugate point 121 illustrated in FIG. 4 can be identified by calculation of electrostatic field in which dielectric constants, electrode potentials, sizes, and a layout relationship of members constituting the X-ray generation tube 100 are modeled. FIG. 6 illustrates an X-ray generation device 10 according to an embodiment of the present invention, from which an X-ray beam is extracted toward the front of an X-ray transmissive window 13. The X-ray generation device 10 of this embodiment includes, inside a housing container 11 that has the X-ray transmissive window 13, the X-ray generation tube 100 as the X-ray source and a driving circuit 5 for driving the X-ray generation tube 100. The driving circuit 5 applies a tube voltage between the electron source and the anode of the X-ray generation tube 100 to form an acceleration electric field between the target 105 and the electron emitting portion. By appropriately setting the tube voltage in accordance with a thickness of the target layer and an element type of the target metal, it is possible to select a type of radiation necessary for imaging. It is preferred that the housing container 11 for housing the X-ray generation tube 100 and the driving circuit 5 have sufficient strength as a container and be superior in heat dissipation. As a structural material thereof, a metal material such as brass, iron, or stainless steel is used. In this embodiment, an insulating liquid is filled in the space in the housing container 11 except for the X-ray generation tube 100 and the driving circuit 5. The insulating liquid is liquid having electric insulation and has a role of maintaining electric insulation in the housing container 11 and a role as a medium for cooling the X-ray generation tube 100. As the insulating liquid, it is preferred to use an electric insulating oil such as mineral oil, silicone oil, and perfluoro oil. Next, with reference to FIG. 7, a structural example of an X-ray imaging system 1 including the X-ray generation tube 100 of the present invention is described. A system control device 2 integrally controls the X-ray generation device 10 and a radiation detection device 6 for detecting an X-ray. A driving circuit 5 outputs various control signals to the X-ray generation tube 100 under control by the system control device 2. In this embodiment, the driving circuit 5 is housed together with the X-ray generation tube 100 in the housing container for housing the X-ray generation device 10, but may be disposed outside the housing container. The control signal output by the driving circuit 5 controls an emission state of the X-ray beam to be emitted from the X-ray generation device 10. The X-ray beam emitted from the X-ray generation device 10 is radiated to the outside of the X-ray generation device 10 after the radiation range thereof is adjusted by a collimator unit (not shown) having a movable stop, is caused to pass through an object 4, and is detected by a detector 8. The detector 8 converts the detected X-ray into an image signal and outputs the image signal to a signal processing portion 7. Under control by the system control device 2, the signal processing portion 7 performs predetermined signal processing on the image signal and outputs the processed image signal to the system control device 2. Based on the processed image signal, the system control device 2 outputs to a display device 3 a display signal for controlling the display device 3 to display the image. The display device 3 displays on a screen an image based on the display signal as a taken image of the object 4. The X-ray imaging system 1 can be used for a nondestructive inspection of an industrial product, or a pathological diagnosis of a human body or an animal. In the following, structures, actions, and effects of the present invention are described more specifically with reference to examples. The X-ray generation tube 100 of Example 1 is described with reference to FIG. 1A. Each component has a shape symmetric with respect to the center axis 106 except for the electron passage apertures 111 of the grid electrode 110. A generally-used flat surface impregnated type cathode having a diameter of 2 mm was used for the cathode 101, and the grid electrode 110 having a thickness of 75 μm as illustrated in FIG. 1B was disposed at a position apart from the cathode 101 by 0.8 mm. The electron passage apertures 111 of the grid electrode 110 each have a diameter of 300 μm and are arranged in a lattice pattern at a pitch of 350 μm, but there is no aperture at the center portion positioned on the center axis 106. The extraction electrode 102 was disposed on the rear side of the surface of the grid electrode 110 opposed to the electron source 101 and was electrically connected to the grid electrode 110. The extraction electrode has an aperture having a diameter of 5 mm and a thickness of 3 mm in the direction of the center axis 106 of the X-ray. Molybdenum was used for the grid electrode 110 and the extraction electrode 102 considering heat resistance. A focusing electrode 103 is disposed at a position apart from the extraction electrode 102 by 4 mm, and has an aperture of a thickness of 5 mm and a diameter of 10 mm. Stainless steel was used as material of the focusing electrode 103. The anode 104 was disposed at a position apart from the focusing electrode 103 by 40 mm. In addition, because high energy flows in, diamond was disposed on the center axis 106 considering heat dissipation and X-ray transmittance, and copper was used around the diamond. Further, on the diamond, a film made of tungsten having a thickness of 5 μm was formed as the target 105. In this way, it is possible to constitute the transmission type target in which the X-ray generated from the target 105 can be extracted to the outside through the diamond disposed just below the target 105. In the X-ray generation tube 100 configured as described above, a voltage of 0 V was applied to the electron source 101, a voltage of 200 V was applied to the extraction electrode 102 and the grid electrode 110, a voltage of 2,000 V was applied to the focusing electrode 103, and a voltage of 75 kV was applied to the anode 104 and the target 105. Then, it was confirmed that the target-side current density distribution profile of the focal point was formed to be recessed a little at the center portion as shown in FIG. 2B, that is, the current density at the focal point center was able to be suppressed. Therefore, a voltage of the extraction electrode was increased so as to increase the current up to a thermal limit of the target 105. As a result, it was possible to input energy up to the maximum of 1.7 kW. As Comparative Example, there was manufactured the X-ray generation tube 200 including the grid electrode 210 having the electron passage apertures 211 each having a diameter of 300 μm and arranged uniformly in a lattice pattern at a pitch of 350 μm. The X-ray generation tube 200 of Comparative Example had X-ray intensity in a Gaussian distribution as shown in FIG. 2A. The maximum value of the energy amount input to the target 205 in the X-ray generation tube 200 of Comparative Example was 1.5 kW. As understood from the above, the X-ray generation tube 100 of Example 1 was able to obtain higher output power of the X-ray than the X-ray generation tube 200 of Comparative Example by approximately 13%. As the grid electrode 110, a grid electrode similar to that of Example 1 except for the electron passage apertures 111 arranged in a honeycomb pattern as illustrated in FIG. 3A was prepared, and further the X-ray generation tube was manufactured similarly to Example 1 so as to perform the test. As a result, in this example as well, the X-ray intensity distribution as shown in FIG. 2B was obtained, and the energy input up to 1.7 kW was able to be performed. As the grid electrode 110, the grid electrode 110 as illustrated in FIG. 3B was manufactured, in which the electron passage apertures 111 are disposed at the same positions as those of FIG. 5B, but a diameter of an aperture near the center portion is different. This grid electrode 110 is the same as that of FIG. 5B except that a center aperture and four apertures adjacent to the center aperture in the up and down direction and in the right and left direction each have a diameter of 100 μm, and four apertures diagonally adjacent to the center aperture each have a diameter of 200 μm. This grid electrode 110 was prepared, and the X-ray generation tube was manufactured similarly to Example 1 so as to perform the test. As a result, in this example as well, the energy input up to 1.7 kW was able to be performed. While the present invention has been described with reference to exemplary embodiments, it is to be understood that the invention is not limited to the disclosed exemplary embodiments. The scope of the following claims is to be accorded the broadest interpretation so as to encompass all such modifications and equivalent structures and functions. This application claims the benefit of Japanese Patent Applications No. 2013-043842, filed Mar. 6, 2013, and No. 2014-029252, filed Feb. 19, 2014 which are hereby incorporated by reference herein in their entirety.
	description	This application claims priority under 35 U.S.C. § 119 from Japanese Patent Application No. 2019-172262, filed Sep. 20, 2019, the entire disclosure of which is herein expressly incorporated by reference. The present invention relates to a predictive model construction method and prediction method of a reactor water radioactivity concentration of a nuclear power plant. For example, a boiling water nuclear power plant and a pressurized water nuclear power plant are known as nuclear power plants (also simply referred to as plants). In the plants, for a main constituent member such as a reactor pressure vessel, stainless steel, nickel base alloy, or the like is used in a water contact portion that is configured to contact water in order to inhibit corrosion. Further, in the plants, a part of cooling water (hereinafter, also referred to as reactor water) existing in the reactor pressure vessel is cleaned up by a reactor water cleanup system, and metal impurities exist in a small amount in the reactor water are actively removed. Even if corrosion prevention measures are taken as described above, an existence of a very small amount of metal impurities remaining in the reactor water is unavoidable, and therefore a part of the metal impurities adhere to, as metal oxides, outer surfaces of fuel rods included in a fuel assembly. Irradiation of neutrons emitted from nuclear fuel materials in the fuel rods cause nuclear reactions of metal elements contained in the metal impurities adhered to the outer surfaces of the fuel rods, and the metal elements are turned into radionuclides such as cobalt 60, cobalt 58, chromium 51, and manganese 54. A part of the radionuclides adhered to the outer surfaces of the fuel rods in forms of oxides are eluted as ions in the reactor water in accordance with solubility of the oxides in which the radionuclides are incorporated. The radionuclides are re-emitted into the reactor water as insoluble particles, which are referred to as claddings. Apart of the radionuclides in the reactor water is removed by the reactor water cleanup system. However, radionuclides that are not removed are accumulated on surfaces of structure members that contact the reactor water while circulating in a reactor water recirculation system or the like together with the reactor water. As a result, radiation is emitted from the surfaces of the structure members, causing radiation exposure of a person who performs a periodic inspection. An exposure dose of the person is managed so as not to exceed a prescribed value for each person. However, in recent years, the prescribed value has been lowered, causing a necessity to economically reduce the exposure dose of each person as much as possible. In such a situation, it is an effective measure to predict the exposure dose at a time of a next periodic inspection, to make a shielding plan or a work personnel plan and to determine necessity of decontamination besides lowering a total exposure dose. In order to predict the exposure dose, it is necessary to predict a pipe dose, and since the pipe dose is strongly dependent on the reactor water radioactivity concentration during operation, it is important to predict a transition of the reactor water radioactivity concentration during an operation period of the plant. In addition, since the exposure dose of the next periodic inspection is predicted and used for a periodic inspection plan, it is necessary to perform the prediction as fast as possible. As a prediction method of the reactor water radioactivity concentration, for example, there is a water quality diagnostic system for reducing a dose rate of a primary reactor cooling water system using a physical and chemical simulation model as described in JP-A-64-063894 (PTL 1). The water quality diagnostic system predicts a future plant dose rate by using a simulation model (mass balance model) for estimating a change in radioactivity in the cooling water with a current water quality condition as an input, and diagnoses the current water quality condition based on a result of the prediction. In a technique of PTL 1, it is necessary for the simulation model and model parameters to be already existed and optimized. However, in reality, the model parameters change temporally, and as the operation continues, a machine or a material may be replaced and a difference between a model calculation value and an actual measured value may increase. A self-learning diagnosis and prediction device described in JP-A-06-289179 (PTL 2) has a function of automatically repairing model parameters in response to a temporal change such as a specification and a characteristic of a plant, so as to prevent degradation of prediction accuracy and improve a model by self-learning. In a technique described in PTL 2, optimization and contribution of model parameters of a model set based on a physical model or a chemical model are adjusted, and it is necessary to prepare the model in advance. Therefore, it is necessary to describe, as a mathematical expression, a correlation between a state quantity of a prediction target and a state quantity used as input. The correlation of the state quantities represented by a mass balance model is optimized by optimizing the model parameters. However, although the correlation is considered, an appropriate model cannot be constructed when the correlation is complicated and it is difficult to represent the correlation by a mathematical expression. Therefore, it is required to accurately predict the state quantity of the plant even when a correlation is complicated and cannot be represented by a mathematical expression. The invention was made in view of such a background, and an object of the invention is to provide a highly accurate predictive model construction method and prediction method of a reactor water radioactivity concentration of a nuclear power plant. In order to solve the above problems, the predictive model construction method is a predictive model construction method of a reactor water radioactivity concentration predictive model construction device that predicts a radioactive metal corrosion product concentration in reactor water of a nuclear reactor in a nuclear power plant, and the predictive model construction method includes: the predictive model construction device executing a step of calculating a plant state quantity prediction value to be calculated by using a physical model that describes plant state quantities of the nuclear power plant including a flow rate of feedwater and a metal corrosion product concentration in feedwater of the reactor water in the nuclear reactor, and a step of causing a machine learning model to learn learning data for supervised learning so as to construct a predictive model, and the learning data for supervised learning includes the plant state quantity prediction value and a plant state quantity that is able to be actually measured and includes at least one of the flow rate of feedwater, the metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, a radioactive metal corrosion product concentration in reactor water, a flow rate of reactor water cleanup system, duration of stay in reactor core in a fuel assembly, and electrical output of the reactor water in the nuclear reactor as input data and includes a radioactive metal corrosion product concentration in the reactor water which is an actual measured value as output data. In order to solve the above problems, the prediction method is a prediction method of a reactor water radioactivity concentration prediction device that predicts a radioactive metal corrosion product concentration in reactor water of a nuclear reactor in a nuclear power plant, the prediction device stores a predictive model that is obtained by causing a machine learning model to learn learning data for supervised learning, the learning data for supervised learning includes a plant state quantity that is able to be actually measured and includes at least one of a flow rate of feedwater, a metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, a radioactive metal corrosion product concentration in reactor water, a flow rate of reactor water cleanup system, duration of stay in reactor core in a fuel assembly, and electrical output of the reactor water in the nuclear reactor and a plant state quantity prediction value calculated by using a physical model that describes plant state quantities of the nuclear power plant including the flow rate of feedwater and the metal corrosion product concentration in feedwater of the reactor water in the nuclear reactor as input data and includes a radioactive metal corrosion product concentration in the reactor water which is an actual measured value as output data, and the prediction method includes the prediction device executing a step of calculating the plant state quantity prediction value based on the plant state quantities by using the physical model, and a step of inputting the calculated plant state quantity prediction value and a plant state quantity that includes at least one of the flow rate of feedwater, the metal corrosion product concentration in feedwater, the metal corrosion product concentration in reactor water, the radioactive metal corrosion product concentration in reactor water, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output of the reactor water in the nuclear reactor into the predictive model as input data and calculating a radioactive metal corrosion product concentration in the reactor water as output data. According to the invention, it is possible to provide a highly accurate predictive model construction method and prediction method of the reactor water radioactivity concentration of the nuclear power plant. A nuclear power plant to be predicted and a simulation model (mass balance model) for estimating a change in radioactivity in reactor water will be described before a reactor water radioactivity concentration prediction device in an embodiment for carrying out the invention is described. <Overview of Nuclear Power Plant> FIG. 1 is an overall system configuration diagram of a nuclear power plant P100 according to the present embodiment. A schematic configuration of the nuclear power plant P100 (for example, a boiling water nuclear power plant) to which the reactor water radioactivity concentration prediction device according to the present embodiment is applied will be described with reference to FIG. 1. The nuclear power plant P100 includes a nuclear reactor P1, a turbine P3, a condensate water device P4, a nuclear reactor cleanup system, and a feedwater system. The nuclear reactor P1 installed in a nuclear reactor storage vessel P11 includes a reactor pressure vessel P12 with a reactor core P13 built in. A cylindrical reactor core shroud P15 installed in the reactor pressure vessel P12 surrounds the reactor core P13. A plurality of fuel assemblies (not shown) are loaded in the reactor core P13. Each fuel assembly includes a plurality of fuel rods filled with a plurality of fuel pellets produced from a nuclear fuel material. An annular downcomer P17 is formed between an inner surface of the reactor pressure vessel P12 and an outer surface of the reactor core shroud P15. A plurality of internal pumps P21 are installed at a bottom of the reactor pressure vessel P12. Impellers of the internal pumps P21 are disposed below the downcomer P17. The feedwater system includes a condensate water pump P5, a condensate water cleanup device P6, a feedwater pump P7, a low pressure feedwater heater P8, and a high pressure feedwater heater P9 installed in this order from the condensate water device P4 toward the reactor pressure vessel P12 in a feedwater pipe P10 connecting the condensate water device P4 and the reactor pressure vessel P12. A hydrogen injection device P16 is connected to the feedwater pipe P10 between the condensate water cleanup device P6 and the feedwater pump P7 by a hydrogen injection pipe P18. An on-off valve P19 is provided on the hydrogen injection pipe P18. The nuclear reactor cleanup system includes a cleanup system isolation valve P23, a cleanup system pump P24, a regenerative heat exchanger P25, a non-regenerative heat exchanger P26, and a reactor water cleanup system P27 installed in this order in a stainless steel cleanup system pipe (stainless steel member) P20 connecting the reactor pressure vessel P12 and the feedwater pipe P10. A residual heat removal system provided in the nuclear power plant P 100 includes a residual heat removal system pipe P28 whose one end portion is connected to the reactor pressure vessel P12 and is connected to the downcomer P17, and whose another end portion is connected to an inside of the reactor pressure vessel P12 and is above the reactor core P13. A residual heat removal system pump P29 and a heat exchanger (cooling device) P30 are installed in the residual heat removal system pipe P28. One end of the cleanup system pipe P20 is connected to the residual heat removal system pipe P28 upstream of the residual heat removal system pump P29. Cooling water (reactor water) existing in the downcomer P17 in the reactor pressure vessel P12 is pressurized by the internal pumps P21 and guided to a lower plenum below the reactor core P13. The reactor water is supplied from the lower plenum to the reactor core P13, and is heated by heat generated by fission of the nuclear fuel material contained in the fuel rods of the fuel assembly. A part of the heated reactor water is turned into steam. The steam is guided from the reactor pressure vessel P12 through a main steam pipe P2 to the turbine P3 to rotate the turbine P3. A power generator (not shown) connected to the turbine P3 is rotated to generate electric power. The steam discharged from the turbine P3 is condensated by the condensate water device P4 and is turned into water. The water is supplied as feedwater into the reactor pressure vessel P12 through the feedwater pipe P10. The feedwater flowing through the feedwater pipe P10 is pressurized by the condensate water pump P5, impurities are removed by the condensate water cleanup device P6, the water is further pressurized by the feedwater pump P7, and heated by the low pressure feedwater heater P8 and the high pressure feedwater heater P9. Extraction steam extracted from the main steam pipe P2 and the turbine P3 at an extraction pipe P14 is supplied to the low pressure feedwater heater P8 and the high pressure feedwater heater P9 respectively, and serves as a heating source of the feedwater. A certain percentage of the reactor water is cleaned up by a reactor water cleanup system since the reactor water in the reactor pressure vessel P12 contains a metal corrosion product contained in the feedwater and a product generated by corrosion of structural material inside the reactor pressure vessel P12. The reactor water in the reactor pressure vessel P12 is supplied to the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26 through the cleanup system pipe P20 branched from the residual heat removal system pipe P28 by driving of the cleanup system pump P24, and is cooled to about 50° C. by the heat exchangers. After the metal corrosion product contained in the reactor water is removed by allowing the cooled reactor water to pass through the reactor water cleanup system P27, and the reactor water is heated by the regenerative heat exchanger P25, the heated reactor water merges with the feedwater flowing in the feedwater pipe P10 and is supplied to the reactor pressure vessel P12. All control rods (not shown) are inserted into the reactor core when the operation of the nuclear reactor P1 is stopped. The fission reaction of the nuclear fuel material is stopped due to the insertion of all the control rods, and the operation of the nuclear reactor P1 is stopped. Although residual heat in the reactor core P13 and in machines in the reactor pressure vessel P12 is removed by evaporation of the reactor water, the reactor core P13 and the machines in the reactor pressure vessel P12 are cooled by using the residual heat removal system when a temperature of the reactor water drops to about 150° C., since efficiency of heat removal by evaporation of the reactor water decreases when the temperature drops to some extent. That is, the reactor water in the reactor pressure vessel P12 is supplied to the heat exchanger P30 through the residual heat removal system pipe P28 under driving of the residual heat removal system pump P29, and then the reactor water is cooled by the heat exchanger P30 and returned to the reactor pressure vessel P12. The reactor water during operation of the nuclear reactor contains a radioactive metal corrosion product, including cobalt 60, causing adherence to the structural material in accordance with the concentration thereof, and causing radiation exposure of a person who performs periodic inspection due to radiation from radionuclides adhering to the structural material. According to the reactor water radioactivity concentration prediction device according to the present embodiment, a radioactive metal corrosion product concentration can be predicted with high accuracy. By using the prediction, an operator of the plant can design a plan to inhibit the exposure by reducing the reactor water radioactivity concentration during operation, laying a shield, or performing chemical decontamination. <Overview of Mass Balance Model> FIG. 2 is a diagram showing a mass balance model 140 (see FIG. 3 described later) of a migration behavior of a metal corrosion product to the reactor water according to the present embodiment. The mass balance model 140 is a physical model that describes, according to the macro mass conservation law, dynamic behaviors of allowing a metal corrosion product contained in the feedwater and a metal corrosion product generated due to corrosion of structural materials inside and outside the reactor in contact with the reactor water to be re-adhered to surfaces of the fuel rods or surfaces of the structural materials inside and outside the reactor where the reactor water is interposed, and removing the metal corrosion products out of the system by the reactor water cleanup system. The solid line arrow and the dashed line arrow in FIG. 2 indicate migration by cladding and ions. The mass balance model 140 of the metal corrosion product is described by a group of simultaneous differential equations shown in Equations 1 to 8 below. d ⁢ C d ⁢ t = 1 V ⁢ { F f ⁢ C f + X + ζ ⁢ ⁢ M + ζ P 1 ⁢ m 1 + ζ P 2 ⁢ m 2 - ( δ ⁢ V + β ⁢ V + δ P 1 ⁢ S 1 + δ P 2 ⁢ S 2 ) ⁢ C } ( Equation ⁢ ⁢ 1 ) d ⁢ M d ⁢ t = δ ⁢ V ⁢ C - ζ ⁢ ⁢ M ( Equation ⁢ ⁢ 2 ) d ⁢ m 1 dt = δ P 1 ⁢ S 1 ⁢ C - ζ P 1 ⁢ m 1 ( Equation ⁢ ⁢ 3 ) d ⁢ m 2 dt = δ P 2 ⁢ S 2 ⁢ C - ζ P 2 ⁢ m 2 ( Equation ⁢ ⁢ 4 ) d ⁢ R d ⁢ r = 1 V ⁢ { Y + ζ ⁢ A + ζ P 1 ⁢ Γ 1 + ζ P 2 ⁢ Γ 2 - ( δ ⁢ V + β ⁢ V + δ P 1 ⁢ S 1 + δ P 2 ⁢ S 2 + λ ) ⁢ R } ( Equation ⁢ ⁢ 5 ) d ⁢ A d ⁢ t = G ⁢ M + δ ⁢ R ⁢ V - ( ζ + λ ) ⁢ A ( Equation ⁢ ⁢ 6 ) d ⁢ Γ 1 d ⁢ t = G 1 ⁢ m 1 + δ P 1 ⁢ S 1 ⁢ R - ( ζ P 1 + λ ) ⁢ A ( Equation ⁢ ⁢ 7 ) d ⁢ Γ 2 d ⁢ f = δ P 2 ⁢ S 2 ⁢ R - ( ζ P 2 + λ ) ⁢ A ( Equation ⁢ ⁢ 8 ) In the above equations, meanings of variables and parameters are as follows. C: metal corrosion product concentration in reactor water (concentration of, for example, iron, nickel, cobalt, and the like) t: time point V: retention amount of reactor water Ff: flow rate of feedwater Cf: metal corrosion product concentration in feedwater X: generation rate of metal corrosion product generated by corrosion of in-reactor structural materials ζ: elution or release constant of fuel rod adhesive substances ζp1: elution or release constant of in-reactor structural material adhesive substances ζp2: elution or release constant of out-reactor structural material adhesive substances M: amount of metal corrosion product accumulated on fuel rods m1: amount of metal corrosion product adhered to and accumulated on surfaces of in-reactor structural materials m2: amount of metal corrosion product adhered to and accumulated on surfaces of out-reactor structural materials δ: adhesive constant for fuel rods β: removal rate in nuclear reactor cleanup system δp1: adhesive constant for in-reactor structural materials δp2: adhesive constant for out-reactor structural materials S1: surface area of in-reactor structural materials S2: surface area of out-reactor structural materials R: radioactive metal corrosion product concentration in reactor water (concentration of, for example, cobalt 60, cobalt 58, manganese 54, and the like) Y: generation rate of radioactive metal corrosion product generated by corrosion of in-reactor structural materials A: amount of radioactive metal corrosion product accumulated on fuel rods Γ1: amount of radioactive metal corrosion product adhered to and accumulated on surfaces of in-reactor structural materials Γ2: amount of radioactive metal corrosion product adhered to and accumulated on surfaces of out-reactor structural materials λ: decay constant of radioactive metal corrosion product G: production rate of radionuclides on fuel rods G1: production rate of radionuclides on in-reactor structural materials Among the above variables, C, Cf, Ff, R, and Γ2 are state quantities that can be measured during operation, and M, A, and Γ1 are state quantities that can be measured when the fuel rods are taken out from the inside of the reactor at a time of stopping of periodic inspection and the like. V, S1, and S2 are plant parameters unique to the plant. λ, G, and G1 are physical constants determined in accordance with nuclides of the radioactive metal corrosion product, and X, Y, ζ, ζp1, ζp2, δ, δp1, δp2, and β are model parameters in principle. m1 and m2 are state quantities that are usually difficult to be measured since the product adhered from the water side and the product generated by corrosion of the structural materials cannot be distinguished. In a related art, the model parameters are adjusted so as to allow calculated values and actual measured values of the state quantities such as C, M, and R on a left side to match, and future estimated values of Cf are given as inputs and the state quantities on the left side are calculated and predicted by using the adjusted model parameters. <Reactor Water Radioactivity Concentration Prediction Device: Overall Configuration> Hereinafter, a reactor water radioactivity concentration prediction device for predicting a radioactivity concentration of the reactor water in the nuclear power plant P100 according to the present embodiment will be described. The reactor water radioactivity concentration prediction device predicts concentrations of cobalt 60, cobalt 58, chromium 51, manganese 54, and the like in the reactor water. Simulation and machine learning technology based on the mass balance model 140 described by Equations 1 to 8 are used in the prediction. Specifically, a result of the simulation is used as apart of the input to a machine learning model. FIG. 3 is a functional configuration diagram of a reactor water radioactivity concentration prediction device 100 according to the present embodiment. The reactor water radioactivity concentration prediction device 100 is a computer, and includes a control unit 110, a storage unit 120, and an input/output unit 160. The reactor water radioactivity concentration prediction device 100 receives operation data such as a thermal output from a process computer P110 used in the nuclear power plant P100 and output data from measuring devices P120 installed in the nuclear power plant P100. The storage unit 120 includes a random access memory (RAM), a read only memory (ROM), a solid state drive (SSD), and the like, and stores a predictive model 130, a mass balance model 140, and a plant state quantity database 150. The predictive model 130 is a learning model of machine learning, and is, for example, a neural network. The mass balance model 140 is the simulation model described by Equations 1 to 8, and is the equations or the parameters that describe the model. Alternatively, the mass balance model 140 may be regarded as a program for executing the simulation. The plant state quantity database 150 stores plant state quantities including plant data, feedwater data, and reactor water quality data of the nuclear power plant P100 that serve as values of the variables and parameters included in Equations 1 to 8. The plant state quantity database 150 also stores other plant state quantities such as duration of stay in reactor core of the fuel assemblies loaded in the reactor core P13 (see FIG. 1) and electrical outputs. The input/output unit 160 receives data from the process computer P110 and the measuring devices P120 and stores the data in the plant state quantity database 150. The input/output unit 160 includes a display, a keyboard, and a mouse (not shown), receives operations from the user of the reactor water radioactivity concentration prediction device 100, and displays data such as a prediction result. The control unit 110 is configured by a central processing unit (CPU), and includes a learning unit 111, a prediction unit 112, and a simulation unit 113. The learning unit 111 performs learning processing (see FIG. 6 described later) by using the data stored in the plant state quantity database 150 as data for supervised learning (learning data), and generates the predictive model 130 for predicting the radioactivity concentration of the reactor water. The prediction unit 112 inputs the data stored in the plant state quantity database 150 to the generated predictive model 130, and executes prediction processing (see FIG. 9 described later) for predicting the radioactivity concentration of the reactor water. The simulation unit 113 executes a simulation based on the mass balance model 140 described by Equations 1 to 8. A simulation execution result is input data of the predictive model 130. Details of processing of the learning unit 111 and the prediction unit 112 will be described later with reference to FIGS. 4 to 9. <Data for Supervised Learning> FIG. 4 is a diagram showing a configuration of original data of the data for supervised learning of the predictive model 130 in the learning processing according to the present embodiment. FIG. 5 is a diagram showing the input data and output data of the predictive model 130 included in the data for supervised learning according to the present embodiment. Here, an example will be described, in which one set of data for supervised learning is generated from plant state quantities 151 and cobalt 60 concentrations 152 for 30 days. A data set 210 described in FIG. 4 is an original data set of one set of data for supervised learning whose start date is the first day, and includes plant state quantities on the first day to the 30th day, and the reactor water radioactivity concentrations of cobalt 60 (also referred to as cobalt 60 concentrations) on the 60th day. The pieces of data are stored in the plant state quantity database 150. Plant state quantities that can be actually measured (actually measured data), such as a flow rate of feedwater, a metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, a radioactive metal corrosion product concentration in reactor water, a flow rate of the reactor water cleanup system, duration of stay in reactor core of the fuel assembly, and an electrical output, which are included in each plant state quantity from the first day to the 30th day, are included as input data of the predictive model 130. The data for supervised learning which is input data of the predictive model 130 corresponds to an arrow from the plant state quantity 151 to the predictive model 130 shown in FIG. 5. The amount of metal corrosion product (M) and the amount of radioactive metal corrosion product (A) accumulated on the fuel rods and calculated by using the mass balance model 140 based on the flow rate of feedwater, the metal corrosion product concentration in feedwater, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output, which are included in the plant state quantities from the first day to the 30th day, are included as the input data (also referred to as plant state quantity prediction values) of the predictive model 130. The data for supervised learning which is input data of the predictive model 130 corresponds to an arrow from the mass balance model 140 to the predictive model 130 shown in FIG. 5. Although not directly appearing in Equations 1 to 8, the flow rate of reactor water cleanup system, the duration of stay in reactor core of the fuel assembly, and the electrical output affect the variables and parameters. For example, the adhesive constant for the fuel rods (δ) is dependent on the electrical output, and the removal rate (β) in the nuclear reactor cleanup system is calculated based on the flow rate of reactor water cleanup system. The cobalt 60 concentration (actual measured value) on the 60th day is included as the output data (radioactive metal corrosion product concentration in reactor water) of the predictive model 130. The data for supervised learning which is output data of the predictive model 130 corresponds to an arrow from the cobalt 60 concentration 152 to the predictive model 130 shown in FIG. 5. The input data and the output data described above are one set of data for supervised learning for the predictive model 130. Similarly, one set of data for supervised learning is generated from the data set 220 having the second day as the start date. By repeating the process hereinafter, 30 sets of data for supervised learning for the predictive model 130 are generated from the plant state quantities of the first day to the 59th day and the cobalt 60 concentrations from the 60th day to the 89th day. <Reactor Water Radioactivity Concentration Prediction Device: Learning Processing> FIG. 6 is a flowchart of the learning processing to be executed by the learning unit 111 according to the present embodiment. In step S11, the learning unit 111 executes steps S12 to S13 for each predetermined start date (first day to 30th day). In step S12, upon receiving an instruction from the learning unit 111, the simulation unit 113 sets the plant state quantity 151 as input data, and executes the simulation based on the mass balance model 140. Specifically, the simulation unit 113 calculates the amount of metal corrosion product (M) and the amount of radioactive metal corrosion product (A) accumulated on the fuel rods, based on the flow rate of feedwater, the metal corrosion product concentration in feedwater, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output which are included in the plant state quantities 151 from the start date to the 30th day by using the mass balance model 140 (Equations 1 to 8). In step S13, the learning unit 111 generates the data for supervised learning. Specifically, the learning unit 111 generate the data for supervised learning in which the plant state quantities that can be actually measured such as the flow rate of feedwater, the metal corrosion product concentration in feedwater, the metal corrosion product concentration in reactor water, the radioactive metal corrosion product concentration in reactor water, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output which are included in the plant state quantities 151 for the 30 days and the amount of metal corrosion product and the amount of radioactive metal corrosion product accumulated on the fuel rods that are calculated in step S12 are input data, and the cobalt 60 concentration 152 after 59 days from the start date is output data. In step S14, if the learning unit 111 executes steps S12 to S13 for every predetermined start date, the processing proceeds to step S15, and if there is an unprocessed start date, the processing returns to step S12 and the unprocessed start date is processed. In step S15, the learning unit 111 trains the predictive model 130 using the data for supervised learning generated in step S13 (causes the predictive model 130 to learn the data for supervised learning), and constructs the predictive model 130. The predictive model 130 for predicting the cobalt 60 concentration can be constructed by the above-described prediction processing. Next, processing of predicting the cobalt 60 concentration using the predictive model 130 will be described. <Reactor Water Radioactivity Concentration Prediction Device: Prediction Processing> FIG. 7 is a diagram showing a configuration of original data of input data of the predictive model 130 in prediction processing according to the present embodiment. FIG. 8 is a diagram showing the input data and the output data of the predictive model 130 in the prediction processing according to the present embodiment. In FIG. 7, a reference day is an execution day of the prediction processing, and the prediction unit 112 predicts the cobalt 60 concentration 30 days after the reference day based on the plant state quantities for the past 30 days including the reference day. Specifically, plant state quantities that can be actually measured (actually measured data), such as the flow rate of feedwater, the metal corrosion product concentration in feedwater, the metal corrosion product concentration in reactor water, the radioactive metal corrosion product concentration in reactor water, the flow rate of reactor water cleanup system, the duration of stay in reactor core of the fuel assembly, and the electrical output, which are included in respective plant state quantities from the 29th day before the reference day to the reference day, are included as the input data of the predictive model 130. The input data corresponds to an arrow from a plant state quantity 153 to the predictive model 130 shown in FIG. 8. The amount of metal corrosion product (M) and the amount of radioactive metal corrosion product (A) accumulated on the fuel rods that are calculated by using the mass balance model 140 based on the flow rate of feedwater, the metal corrosion product concentration in feedwater, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output, which are included in the plant state quantities from the 29th day before the reference day to the reference day, are included as the input data (plant state quantity prediction values) of the predictive model 130. The input data corresponds to an arrow from the mass balance model 140 to the predictive model 130 shown in FIG. 8. The cobalt 60 concentration 30 days after the reference day is included as the output data of the predictive model 130. The output data corresponds to an arrow from the predictive model 130 to a cobalt 60 concentration 154 shown in FIG. 8. FIG. 9 is a flowchart of the prediction processing to be executed by the prediction unit 112 according to the present embodiment. In step S21, upon receiving an instruction from the prediction unit 112, the simulation unit 113 sets the plant state quantity 153 as input data, and executes the simulation based on the mass balance model 140. Specifically, the simulation unit 113 calculates the amount of metal corrosion product (M) and the amount of radioactive metal corrosion product (A) accumulated on the fuel rods based on the flow rate of feedwater, the metal corrosion product concentration in feedwater, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output, which are included in the plant state quantities 153 for the past 30 days including the reference day, by using the mass balance model 140 (Equations 1 to 8). In step S22, the prediction unit 112 inputs the input data into the predictive model 130. Specifically, the prediction unit 112 inputs the plant state quantities that can be actually measured such as the flow rate of feedwater, the metal corrosion product concentration in feedwater, the metal corrosion product concentration in reactor water, the radioactive metal corrosion product concentration in reactor water, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output, which are included in the plant state quantities 153 for the past 30 days including the reference day, and the amount of metal corrosion product and the amount of radioactive metal corrosion product accumulated on the fuel rods that are calculated in step S21 into the predictive model 130. In step S23, the prediction unit 112 predicts the cobalt 60 concentration by executing the predictive model 130 and acquiring the output of the cobalt 60 concentration 154. When a reference point is current, the cobalt 60 concentration 154 can be predicted from the plant state quantity 153, which is a past actual value (actual measured value). When the reference point is set in the past, the prediction can be verified. Specifically, accuracy of the prediction can be evaluated by calculating the cobalt 60 concentration 154 based on the plant state quantity 153, which is the past actual value, and comparing the calculated cobalt 60 concentration 154 with the cobalt 60 concentration of an actual value. For example, the accuracy of the prediction can be evaluated by comparing the cobalt 60 concentration on the 30th day before the reference point and a cobalt 60 concentration today. <Modification: Period of Input/Output Data to Predictive Model> In the above-described embodiment, although the data input to the predictive model 130 (see FIGS. 5 and 8) is data of 30 days, the invention is not limited thereto, and the data may be, for example, based on plant state quantities in a longer period. Although the cobalt 60 concentration 30 days after the reference day is predicted, the cobalt 60 concentration in a farther future, for example, 90 days after the reference day, may be predicted. The number of days to be predicted is not limited to one day, and may be, for example, a period from 30 days to 60 days after the reference day. The predictive model 130 is constructed by creating the predictive model 130 in accordance with a desired period of the input data, or an interval between the reference day and a day of predicting (prediction day) the cobalt 60 concentration, and thus the cobalt 60 concentration in the desired period of the input data or an interval between the reference day and a prediction day can be predicted. In the above-described embodiment, plant state quantities between the reference day and the prediction day are not included in the input data. The plant state quantity that is predetermined by an operation plan of the nuclear power plant or predictable may be added to the input data as a planned value. FIG. 10 is a diagram showing a configuration of original data of the data for supervised learning of the predictive model 130 in learning processing according to a modification of the present embodiment. FIG. 11 is a diagram showing a configuration of original data of input data of the predictive model 130 in the prediction processing according to the modification of the present embodiment. In the learning processing, plant state quantities of 29 days may be further added to the original data of the input data of the data for supervised learning in addition to the plant state quantities of 30 days from the start date (see FIG. 4). In the prediction processing, the plant state quantities, which are planned values, from the reference day to the prediction day, are used as the original data of the input data in addition to the plant state quantities of the past 30 days including the reference day (see FIG. 7). For example, the flow rate of feedwater, the metal corrosion product concentration in feedwater, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output are determined by the operation plan (predictable) or can be adjusted by an operation of the nuclear power plant, and may be added to the input to the predictive model 130. The amount of metal corrosion product and amount of radioactive metal corrosion product accumulated on the fuel rods that can be calculated based on these plant state quantities by using the mass balance model 140 may also be added to the input to the predictive model 130. The reactor water radioactivity concentration prediction device 100 can predict the cobalt 60 concentration with higher accuracy by adding input data on days close to the prediction day. <Design of Operation Plan Using Reactor Water Radioactivity Concentration Prediction Device> In order to design an operation plan for inhibiting the cobalt 60 concentration to a desired value or less, the prediction processing may be executed repeatedly while changing the plant state quantity 153 (see FIG. 8) serving as a planned value. For example, an iron concentration in the feedwater exists as the plant state quantity 153. Cobalt 60 is incorporated into a metal oxide containing iron oxide as a main component on the surface of the fuel rods, and the cobalt 60 in the reactor water is eluted therefrom. Therefore, it is considered that the cobalt 60 concentration is likely to be affected by the iron concentration of the feedwater. The iron concentration in the feedwater can be adjusted, and the cobalt 60 concentration can be changed by making a plan and setting the iron concentration in the feedwater to a desired value (planned value at a predicted time point). FIG. 12 is a diagram showing a change in prediction of cobalt 60 concentration data accompanying a change in plant state quantity 153 according to the present embodiment. An upper graph in FIG. 12 is a graph showing the iron concentration in the feedwater included in the plant state quantity 153, data before a time point t is actual data (actually measured data), and the estimated iron concentration to be input as three planned values after the time point t is indicated by a dashed line, a one-dot chain line, and a two-dot chain line. A lower graph of FIG. 12 is a prediction result of the cobalt 60 concentration data corresponding to the plant state quantities including the feedwater iron concentration shown in the upper graph of FIG. 12. It can be seen that the cobalt 60 concentration decreases as the feedwater iron concentration increases. Accordingly, by repeatedly executing the prediction processing while changing the plant state quantity of the planned value such as the feedwater iron concentration, the planned value of the plant state quantity for achieving the desired cobalt 60 concentration can be obtained, and the operation plan of the nuclear power plant P100 can be designed. <Features of Learning Processing and Prediction Processing> A generation source of cobalt 60 in the reactor water is cobalt adhered to the fuel rods, and the cobalt receives neutron irradiation from the fuel rods and is turned into cobalt 60. The cobalt 60 is incorporated into the metal oxide containing iron oxide as the main component on the surfaces of the fuel rods, and the cobalt 60 in the reactor water is eluted therefrom. Therefore, a temporal change in the cobalt 60 concentration in the reactor water is strongly affected by the amount of the metal oxide formed on the surfaces of the fuel rods. The plant state quantity is a value that cannot be measured during operation of the nuclear power plant, and can be measured after the fuel rods are taken out from the reactor core P13 (see FIG. 1) while the nuclear power plant is stopped, such as during the periodic inspection. The input data of the predictive model 130 in the above-described embodiment includes the amount of metal corrosion product (M) and the amount of radioactive metal corrosion product (A) accumulated on the fuel rods that are calculated by using the mass balance model. As compared with a model in which only the actual value (measured value) of the plant state quantity is input, the prediction accuracy of the cobalt 60 concentration can be improved by adding the plant state quantities that cannot be measured during operation but strongly affect the cobalt 60 concentration to the input data of the predictive model 130. As compared with the prediction by the mass balance model, the prediction is performed in consideration of the correlation between the plant state quantities that are not included in the mass balance model, and the accuracy is improved. <Modification: Prediction Processing Using a Plan Pattern> In FIG. 12, the reactor water radioactivity concentration prediction device 100 predicts the cobalt 60 concentration by manually inputting a planned value of a reactor water iron concentration. The planned value of the reactor water iron concentration may be prepared in advance, and the cobalt 60 concentration may be predicted for each planned value, and the prediction result may be output. A pattern of the planned value is, for example, a pattern of migration to an iron concentration that is higher/lower by a predetermined value/a predetermined ratio than a current reactor water iron concentration, a pattern of a length of a period until the migration, or a pattern of a combination thereof. For example, there is a pattern that maintains the concentration after a current reactor water iron concentration is increased by 1 parts per billion (ppb, μg/l) in one day, a pattern that maintains the concentration after the current reactor water iron concentration is reduced by 20% in half a day, and a pattern that increases 0.01 ppb for each day. FIG. 13 is a functional configuration diagram of a reactor water radioactivity concentration prediction device 100A according to the modification of the present embodiment. As compared with the reactor water radioactivity concentration prediction device 100 (see FIG. 3), the reactor water radioactivity concentration prediction device 100A further stores a plan pattern 121 in the storage unit 120, and includes a prediction unit 112A instead of the prediction unit 112. The plan pattern 121 is a pattern of the planned value of the iron concentration described above. The prediction unit 112A executes prediction processing by using a plan pattern described in FIG. 14 to be described later. FIG. 14 is a flowchart of the prediction processing using the plan pattern according to the modification of the present embodiment. In step S31, the prediction unit 112A executes step S32 for each plan pattern 121 of the reactor water iron concentration. In step S32, the prediction unit 112A executes the prediction processing (see FIG. 9) by setting the plant state quantity indicated in the plan pattern of the reactor water iron concentration as original data of the input data. In step S33, the prediction unit 112A proceeds to step S34 when step S32 is executed for all plan patterns 121 of the reactor water iron concentration, and returns to step S32 and processes an unprocessed plan pattern if there is an unprocessed plan pattern 121 of the reactor water iron concentration. In step S34, the prediction unit 112A outputs the result of the prediction processing of step S32 (displays the result on a display of the input/output unit 160). According to the above-described reactor water radioactivity concentration prediction device 100A, the user (operator of the plant) of the reactor water radioactivity concentration prediction device 100A can grasp how a prediction value of the cobalt 60 concentration changes in accordance with the planned value of the reactor water iron concentration. The planned value included in the plan pattern 121 is not limited to the reactor water iron concentration, and may be another plant state quantity affecting the cobalt 60 concentration. <Modification: Output Data of Mass Balance Model to be Input to Predictive Model> In the embodiment described above, the input data of the predictive model 130 calculated by the mass balance model 140 is the amount of metal corrosion product (M) and the amount of radioactive metal corrosion product (A) accumulated on the fuel rods. In the reactor water activity concentration prediction device 100, a metal corrosion product concentration (C) and a radioactive metal corrosion product concentration (R) of the reactor water may be calculated by using the mass balance model 140 as input data of the predictive model 130. Specifically, the radioactive metal corrosion product concentration (C) and the radioactive metal corrosion product concentration (R) of the reactor water calculated in step S12 of the learning processing (see FIG. 6) are added to the input data of the data for supervised learning for the predictive model 130 in step S13. The input data corresponds to the arrow from the mass balance model 140 to the predictive model 130 shown in FIG. 5. Similarly as in the prediction processing (see FIG. 9), the metal corrosion product concentration (C) and the radioactive metal corrosion product concentration (R) of the reactor water calculated in step S21 are added to the input data for the predictive model 130 in step S22. The input data corresponds to the arrow from the mass balance model 140 to the predictive model 130 shown in FIG. 8. The prediction accuracy of the cobalt 60 concentration of the predictive model 130 can be improved by adding the data predicted by the mass balance model 140 as the input data to the predictive model 130. <Modification: Metal for Concentration Prediction> In the above-described embodiment, the reactor water radioactivity concentration prediction device 100 learns and predicts the cobalt 60 concentration. In addition to the cobalt 60, the reactor water radioactivity concentration prediction device 100 can also learn and predict cobalt 58 and manganese 54 in a similar manner. <Modification: Separation of Device that Executes Learning Processing and Prediction Processing> The invention is not limited to the above-described embodiments, and modifications can be made without departing from the scope thereof. For example, the learning processing and the prediction processing executed by the reactor water radioactivity concentration prediction device 100 may be executed by separated devices. A predictive model construction device may execute the learning processing to construct a predictive model, and a prediction device that acquires the constructed predictive model may execute the prediction processing. <Other Modification> In the above-described embodiments, although data for each day is set as the input data, the invention is not limited thereto, and the input data may be, for example, data for every two days or every six hours. An interval of the input data may not be limited to a fixed interval, for example, an interval of input data that is far from the prediction day may be increased. By increasing the interval and reducing the number of the input data, the reactor water radioactivity concentration prediction device 100 can increase speeds of the learning processing and the prediction processing. In the above-described embodiment, amass balance model is used as a model of the migration behavior of the metal corrosion product. In addition, a physical or chemical model suitable for a pressurized water nuclear power plant or another type of reactor (for example, heavy water reactor) may be used. Although some embodiments of the invention have been described above, these embodiments are merely examples, and the technical scope of the invention is not limited thereto. The invention can be implemented in various other embodiments, and various omissions, substitutions, and changes can be made without departing from the spirit of the invention. The embodiments and modifications thereof are included in the scope and gist of the invention described in the description, and are also included in the disclosure described in the claims and equivalents thereof.
	claims	1. A phase-contrast X-ray system for imaging an object, said system comprising: a source for emitting a collimated white beam; means for filtering out a photon energy lower than a selected energy level from the collimated white beam, thereby producing an unmonochromatized beam to be irradiated through the object, wherein a phase of the unmonochromatized beam shifts as the unmonochromatized beam irradiates through the object as a result of refractive index; and detecting means for detecting an unmonochromatized beam image having passed through the object by detecting shifts in the phase of the unmonochromatized beam as the unmonochromatized beam irradiates trough the object. 2. The phase-contrast X-ray system according to claim 1 , further comprising a processor for obtaining an image of the object based on an output of the detecting means. claim 1 3. The phase-contrast X-ray system according to claim 1 , wherein the collimated white beam emitting source includes a synchrotron radiation source. claim 1 4. The phase-contrast X-ray system according to claim 1 , wherein the detecting means includes a scintillation crystal, an optical microscopy objective and a CCD camera. claim 1 5. The phase-contrast X-ray system according to claim 1 , wherein the detecting means is an electronic imaging detector. claim 1 6. The phase-contrast X-ray system according to claim 5 , wherein the electronic imaging detector includes a charge coupled device. claim 5 7. A method of imaging an object using a phase-contrast X-ray system, said method comprising the steps of: extracting a collimated white beam from a source; filtering out photon energies lower than a selected energy level from the collimated white beam, thereby producing an unmonochromatized beam; irradiating the object with the unmonochromatized beam such that a phase of the unmonochromatized beam shifts as the unmonochromatized beam irradiates through the object as a result of refractive index; and detecting an unmonochromatized beam image which has passed through the object by detecting shifts in the phase of the unmonochromatized beam as the unmonochromatized beam irradiates through the object. 8. The method in accordance with claim 7 , further comprising processing an output of the detecting step in order to obtain an image. claim 7 9. The method in accordance with claim 7 , wherein the collimated white beam is extracted from a synchrotron radiation source. claim 7 10. The method in accordance with claim 7 , wherein the object is a thick hydrated biological object. claim 7 11. The method in accordance with claim 7 , wherein the selected photon energy level is about 10 keV. claim 7
	summary
	abstract	A charged-particle beam irradiation device, which irradiates an object to be irradiated with a charged-particle beam, includes a scanning member that scans the object to be irradiated with the charged-particle beam; an irradiation amount setting unit that sets an irradiation amount of the charged-particle beam at a plurality of target scanning positions on a scanning line of the charged-particle beam with which the scanning member scans the object to be irradiated; and a scanning speed setting unit that sets a target scanning speed of the charged-particle beam at each of the target scanning positions on the basis of the irradiation amount set by the irradiation amount setting unit.
060552964	summary	BACKGROUND OF THE INVENTION The present invention relates to an improved radiographic grid for use in an x-ray apparatus, especially for use in an x-ray mammography apparatus. More particularly, it relates to a radiographic grid having foil disposed about individual lamella for reducing lamellae density artifacts. It has been well known since the early days of radiography that secondary or scattered x-rays reduce the contrast of an x-ray image. The low difference in x-ray absorption characteristics between cancerous and non-cancerous tissue has made mammography particularly susceptible to imaging problems caused by scattered radiation. A conventional Bucky grid, consisting of a series of lead foil strips separated by strips of x-ray semi-transparent spacers, helps remove scattered radiation from radiographic fields. The thin strips of x-ray radiation absorbing material are called lamellae and are substantially aligned with the incident course of the radiation from the x-ray source, with the x-rays being transmitted through the gaps between the lamellae. The grid is positioned between the object being analyzed and the image receptor (or film) to reduce scattered radiation, thereby improving image contrast on the film. Radiographic grids have been subject to various recent improvements. For example, U.S. Pat. No. 4,901,335 to Ferlic et al. teaches a reciprocating grid having at least a 90% open area at all positions of travel to transmission of directly incident x-ray radiation (i.e. radiation perpendicular to the tangent of the direction of travel of the grid at the point of incidence). The lamellae are individually positioned and aligned with respect to each other in a grid housing and then a cover sheet, substantially covered with an adhesive, is pressed down onto the edges of the lamellae. Radiographic grids have proved to be a highly useful tool for removing scattered radiation from radiographic fields. However, it can be demonstrated that x-ray images produced with radiographic grids contain a "straight line" density artifact apparently associated with the lamellae. These lines correspond to the distance between individual lamella. Obviously, these line-shaped densities are undesirable as the goal with mammography or any other x-ray application is to eliminate all density related noise so that the resulting image is a true depiction of the patient's status. Primary radiation is orientated in the same axis as the lamellae and passes between them to reach the film. Scattered radiation arises from many points within the patient, and is multidirectional, so that most of it is absorbed by the lamellae, and only a small amount passes between them. The lamellae line artifacts are not characteristic of prim or scattered radiation, but rather are subsequent to secondary radiation. The lamellae line artifacts are analogous to either a wave or a tertiary radiation. The lamellae line artifacts appear resultant from or subsequent to the scattered radiation and are likely an additional emission from the lead or metal based lamellae. The basis for concluding that the lamellae line artifacts are tertiary radiation or wave-related is due to the observation that the lamellae emission which produces the density artifact occurs subsequent or resultant to the secondary radiation. Further, tests have shown that the emission appears to occur only in a downward direction (with respect to the position of the grid) as the lamellae line artifact does not materialize on film placed above the grid. Finally, tests have demonstrated that the line artifact is orientated with respect to the central ray of the x-ray source. In other words, the density line artifact is produced on the left side of lamellae positioned on the far right side of the film. Conversely, the line artifact density is produced on the right side of the lamellae positioned to the far left side of the film. Regardless of whether the artifact originates from a third order energy source, this density, like all other densities associated with radiographic grids, should be eliminated. Little research has been done to find a solution. In fact, a search of standard references revealed no literature describing the effect. A substantial need exists for a radiographic grid apparatus which eliminates the recently identified lamellae line artifact. SUMMARY OF THE INVENTION The present invention provides a radiographic grid configured to reduce or eliminate lamellae line density artifacts. The radiographic grid includes a grid housing and a plurality of x-ray radiation absorbing lamellae disposed within the grid housing. The ends of the lamallae furthest from the x-ray source are coated with a thin metal foil which acts to reduce the lamellae line density artifact otherwise found on the x-ray image.
	abstract	X-ray diagnostic system. The system includes a source of x-rays which communicates with an x-ray beam concentrator spaced apart from the x-ray source and disposed for receiving x-rays from the x-ray source. An x-ray spectrometer is disposed for receiving x-rays from the concentrator. In a preferred embodiment, the concentrator is formed of a cylindrical spiral of a metal-coated plastic material having a surface for reflecting x-rays. In another embodiment, the concentrator includes a plurality of concentric nested cylinders of a metal-coated plastic material for reflecting x-rays. In yet another embodiment, the concentrator is a glass capillary bundle. The concentrator allows the spectrometer to be spaced away from the source of x-rays such as scanning electron microscope.
048517020	summary	BACKGROUND OF THE INVENTION This invention relates to radiation-blocking containers for shielding of radioactive solutions stored within substantially non-shielding vessels. Generally, small vials or microcentrifuge vessels (microtubes) are commonly used to store and transfer small amounts of radioactive material such as .sup.32 P and .sup.125 I isotopes, which produce high energy .beta.-particles. These radioactive materials are often combined with other chemical and biochemical ingredients and the resulting mixtures within these microtubes are subjected to incubations at a variety of temperatures. Microtubes are typically made of polyethylene or polypropylene and include a lid which snaps shut. These vessels have thin walls and provide little shielding from the emitted radiation for the laboratory worker manipulating the samples. Existing radiation shields specifically designed to hold these microtubes consist of a heavy block of radiopaque material such as Plexiglass.TM. which may include multiple drilled-out cylindrical holes for holding several microtubes. Heavy radiopaque cannisters with lids are also used to store single microtubes. SUMMARY OF THE INVENTION The invention features a radiation shield suitable for preventing radiation or radioisotopes from reaching a person using one or more sample vessels holding a radioactive solution. The shield includes a radiopaque container, having a top and a bottom opening, and a means for supporting the vessel within the container. The container blocks .beta.-radiation emitted through the sides of the vessel while allowing continued contact between the vessel and the environment via the bottom opening of the container. In preferred embodiments, the container permits incubation liquid to enter the bottom opening of the container and surround that portion of the vessel holding radioactive solution; the container is a sleeve into which the vessel fits; the sleeve has a substantially flat bottom surface which allows it to rest in a free-standing position on a flat surface; the sleeve has a flange which allows it to be suspended in a rack, and permits incubation liquid to enter the bottom opening of the sleeve when it is suspended in the rack, and surround that portion of the vessel containing the radioactive solution; the container includes a rack for supporting a plurality of vessels, and a radiopaque shield is formed along the perimeter of the rack leaving the bottom of the rack open; the container includes a radiopaque cover for shielding radiation emanating through the top opening. In yet another embodiment, the container is a block of radiopaque material having a plurality of bores, which traverse the block, and the container is transparent. This invention provides radiation shielding for vessels, microtubes, and other small vessels, with the sleeves or racks extending from the lid of a microtube to a point some distance below the microtube to minimize lateral exposure of radiation to the laboratory worker. For example, the laboratory worker can safely hold a sleeve by its sides without significant exposure of the hand to radioactivity. The geometry of the sleeves (open at the bottom) permits liquid in any incubator bath to directly enter the sleeve and surround at least the lower part of the microtube housed within the sleeve to allow accurate temperature control of the radioactive sample. Thus, the sleeve simultaneously provides an individual a portable shield which permits safe hand-manipulation of the vessel as well as direct liquid incubation of the shielded vessel. Hand and finger exposure to radioactivity often occurs during pipeting operations involving hand-held microtubes and other small vessels. Such exposure is encountered in formulating radioactive cocktails for DNA sequencing reactions, conducting immuno precipitation reactions with radio-iodinated reagents, and hand-loading samples into various analytical devices to separate various radioactive molecular species. Manipulations of these sorts are almost impossible to complete without hand and finger contact with the small vessels bearing the radioactive solutions even when proper .beta.-radiation shielding is being utilized. For these operations, the transparent sleeve provides both lateral radiation shielding of the microtube and physical separation between the lid of a microtube and the fingers holding the outer sleeve. This separation minimizes the risk of radioactive contamination to the hand caused by liquid residue on the lid of the microtubes accidentally contacting the gloved hand. Using the sleeves during microtube vortexing and other microtube manipulation steps further reduces radiation exposure of the fingers and the hands. Thus, although narrow columns of radiation are permitted to escape upward and/or downward from the vessels, these columns do not create any great risk of exposure to radiation. On the contrary the shields allow easier manipulation of vessels and thus lower the risk of radioactive spills from the vessels caused by the awkwardness of other types of shielding. Transparent shielding also makes manipulation of the vessels easier since the user need not guess when he is touching the radioactive solution with, for example, a pipetting device. Other features and advantages of the invention will be apprent from the following description of the preferred embodiments, and from the claims. DESCRIPTION OF THE PREFERRED EMBODIMENTS The Figures are first briefly described.
	abstract	A method of separating Osmium from Iridium, including receiving a powdered mixture of Osmium and Iridium, oxidizing the Osmium of the powdered mixture, capturing the oxidized Osmium in a trapping solution, reducing the oxidized Osmium from the solution to release the Osmium.
	description	The present invention generally relates to ion implantation, and more particularly to a method for real-timely monitoring ion implantation. Ion implantation is a very important technique in the manufacture of integrated circuit, memory, flat plan display, solar cell, and so on. The accuracy of performing the ion implantation influences on final yielding rate of the manufacture significantly. In general, the ion implantation is performed by making a relative motion between an ion beam and a workpiece, such as wafer and glass plate, and then the ion implantation result is inaccurate if the projection of the ion beam on the workpiece and/or the relative motion between the ion beam and the wafer is not qualified enough. Hence, it is important to monitor how the workpiece is implanted by the ion beam. Some conventional approaches are popularly used for monitoring the ion beam. One conventional approach is monitoring an ion beam when the ion beam is projected into a Faraday cup before or after scanning the ion beam through a workpiece. However, the ion beam is not monitored during an implantation period of scanning the workpiece by the ion beam, and then the variation on the ion beam during the implantation period can not be monitored. Another conventional approach is inferring the ion beam by analyzing a distribution of ions implanted in the workpiece after finishing an implantation. Clearly, the ion beam still is not monitored real-timely. Still another conventional approach is monitoring an ion beam from one or more Faraday cups close to, such as located behind, the workpiece, so that at least partial ion beam not implanted into the workpiece may be monitored by the Faraday cup during the period of scanning the workpiece by the ion beam. And then, the measured profile is compared with a pre-determined profile to monitor the variation of the ion beam during the scanning period. Herein, the difference between the measured profile and the pre-determined profile may indicate the quality of the practical ion implantation on the workpiece by the ion beam. The approach is somewhat real-timely than previous approaches, but no more ion beam message may be acquired from the measured profile. One more conventional approach is using a profiler to measure the ion beam during a beam turning period, so that some ion beam geometric messages may be acquired before the workpiece being implanted by the ion beam. For example, the ion beam shape, the ion beam height, the ion beam width and the ion beam center. This approach is more accurate owing to the function of the profile, but still can not real-timely monitor the ion beam. Besides, all these conventional approaches only can monitor whether the ion beam quality, at most provide some messages of the ion beam, but none can be used to tune the ion implantation during the implantation period. Accordingly, there is a need for having approach for more real-timely monitoring the ion beam, even for acquiring more ion beam message and tuning the ion implantation during the implantation period. The present invention is directed to a method for monitoring ion implantation much real-timely by monitoring a profile having numerous signals respectively relevant to numerous relative positions between the ion beam and the workpiece. Moreover, the present invention also is directed to some applications of the profile, no matter by directly analyzing the profile or by other way to utility the profile. One embodiment provides a method for monitoring ion implantation. The method has at least the following steps. First, provide an ion beam and a workpiece. Next, implant the workpiece by the ion beam and generate a profile having numerous signals relevant to respectively numerous relative positions between the ion beam and the workpiece, wherein the profile has at least a higher portion, a gradual portion and a lower portion. Therefore, analyze the profile without referring to a pre-determined profile, so as to monitor the ion implantation much real-timely. Another embodiment provides a method for monitoring ion implantation. The method has at least the following steps. First, provide an ion beam and numerous workpieces. Then, implant one or more workpieces by the ion beam in sequence. Herein, for each implanted workpiece, a profile having numerous signals relevant to respectively numerous relative positions between the ion beam and the workpiece is generated and analyzed without referring to a pre-determined profile. Herein, each generated profile has at least a higher portion, a gradual portion and a lower portion. Next, generate a reference being a function of these profiles corresponding to these implanted workpieces. After that, implant another workpiece by the ion beam and generate an another profile, wherein the another profile has numerous signals relevant to respectively numerous relative positions between the ion beam and the another workpiece, wherein the another profile has at least a higher portion, a gradual portion and a lower portion. And then, analyze the another profile without referring to the pre-determined profile to generate an analyzed result. Finally, compare the analyzed result of the another profile with the reference, so as to the variation of the ion beam, or the variation of the ion implantation, during a period of implanting these implanted workpieces and the another workpiece may be monitored much-timely. In at least a specific embodiment, the profile may be a current curve formed with numerous current values measured at different relative positions by a Faraday cup close to the workpiece. In at least another specific embodiment, the profile may be a capacitance-related current curve formed with numerous capacitance-related current values measured at different relative positions by a capacitance meter electrically coupled with the workpiece. In other specific embodiments, to monitor the ion beam quality and/or to improve the control of implanting the workpiece by the ion beam, at least one of the following steps may be performed: (a) calculate an ion beam width and/or an ion beam height of the ion beam according to a span of the gradual portion of the profile, even calculate an ion beam contour and/or an ion beam center by using the spans of many gradual portions generated on many scanning lines where the ion beam is scanned through the workpiece accordingly; (b) convert the gradual portion to generate a current distribution of the ion beam, even to adjust how the workpiece is scanned by the ion beam accordingly; (c) monitor whether there is any glitch on the profile or on the converted current distribution of the ion beam, even to adjust/stop the ion implantation accordingly; and (d) change the relative moving direction between the ion beam and the workpiece when all currently measured signals are located on the higher portion and fix the relative moving direction at least a currently measured signal is located on the gradual portion, even to assign at least a turn around point of at least a specific scanning line at a specific position wherein a distance between the specific position and an edge of the workpiece along the specific scanning line is essentially equal to a span of the gradual portion when said workpiece is scanned by said ion beam along a plurality scanning lines over said workpiece. Still in other specific embodiments that numerous workpieces are implanted by the ion beam, at least one of the fowling steps may be performed: (a) after the reference is generate, implant other workpieces by the ion beam when the corresponding profile of each of the other workpieces is compared with the reference; (b) tune the ion beam whenever a difference between the analyzed result and the reference is un-acceptable, wherein the reference is updated by using the tuned ion beam and then other un-implanted workpieces are implanted and monitored by comparing the corresponding profile of each of the other un-implanted workpieces with the updated reference; (c) tune one or more ion implantation parameters whenever a difference between the analyzed result and the reference is un-acceptable, wherein the reference is updated by using the tuned ion beam and then other un-implanted workpieces are implanted and monitored by comparing the corresponding profile of each of the other un-implanted workpieces with the updated reference, wherein these implantation parameters may include ion beam energy, ion beam direction, ion beam diverse, position of a workpiece right be implanted by ion beam, and an alignment between ion beam and a workpiece right be implanted by the ion beam; and (d) compare the reference with a measured result measured by using a profiler to measure the ion beam, where these un-implanted workpieces will not be implanted if a difference between the reference and the measured result is un-acceptable. Accordingly, how the workpiece is implanted by the ion beam may be monitored much real-timely, because the profile may be generated simultaneously during a period of implanting the workpiece by the ion beam. Moreover, by analyzing the profile, not only the quality of the ion beam may be further monitored but also how the workpiece is implanted by the ion beam may be further improved. Reference will now be made in detail to appended embodiments of the present invention. Examples of these embodiments are illustrated in the accompanying drawings. While the invention will be described in conjunction with these specific embodiments, it will be understood that it is not intended to limit the invention to these embodiments. In fact, it is intended to cover alternatives, modifications, and equivalents as may be included within the spirit and scope of the invention as defined by the appended claims. In the following description, numerous specific details are set forth in order to provide a through understanding of the present invention. The present invention may be practiced without some or all of these specific details. In other instances, well-known process operations are not described in detail in order not to obscure the present invention. FIG. 1 illustrates a block diagram of a method for monitoring ion implantation according to an embodiment of the present invention. FIG. 2A to FIG. 2C illustrate respectively different relative positions between the ion beam and the workpiece during an implantation period according to an embodiment of the present invention. FIG. 3 illustrates a current curve relevant to the relative positions as illustrated in FIG. 2A to FIG. 2C. FIG. 4A to FIG. 4C illustrate respectively different relative positions between the ion beam and the workpiece during an implantation period according to another embodiment of the present invention. In addition, FIG. 5 illustrates a capacitance-related current curve relevant to the relative positions as illustrated in FIG. 4A to FIG. 4C. Referring to FIG. 1 first, the method for monitoring ion implantation according to an embodiment of the present invention comprises the following steps. First, provide an ion beam and a workpiece (S110). Next, implant the workpiece by the ion beam and generate a profile having numerous signals relevant to respectively numerous relative positions between the ion beam and the workpiece, wherein the profile has at least a higher portion, a gradual portion and a lower portion (S120). Finally, analyze the profile without referring to a pre-determined profile (S130). Note that the profile is generated during an implantation period of implanting the workpiece by the ion beam. Hence, the ion implantation may be monitored much real-timely. In an embodiment, the profile is a current curve as illustrated in FIG. 3. The current curve is formed with numerous current values measured from a Faraday cup 132 close to the workpiece 200 at different relative positions between the ion beam 112 and the workpiece 200 as illustrated in FIG. 2A to FIG. 2C. In another embodiment, the profile is a capacitance-related current curve as illustrated in FIG. 5. The capacitance-related current curve is formed with numerous capacitance-related current values measured from a capacitance meter having two electrodes 122/124 electrically coupled with the workpiece 200 at different relative positions between the ion beam 112 and the workpiece 200 as illustrated in FIG. 4A to FIG. 4C. In detail, referring to FIG. 2A and FIG. 3 first, when a relative position between the ion beam 112 and the workpiece 200 is as illustrated in FIG. 2A, whole the ion beam 112 is projected outside the workpiece 200 and may be projected on the Faraday cup 132. It means the ion beam 112 is not crossing an edge of the workpiece 200 at this time. Hence, a higher portion A1 with a substantially stable current value equal to a total current of the ion beam 112 as illustrated in FIG. 3 is measured. And then referring to FIG. 2B and FIG. 3, when a relative position between the ion beam 112 and the workpiece 200 is as illustrated in FIG. 2B, a part of the ion beam 112 is projected on the workpiece 200, and the other part of the ion beam 112 may be projected on the Faraday cup 132. It means the ion beam 112 is crossing an edge of the workpiece 200 at this time. Hence, a gradual portion B1 as illustrated in FIG. 3 is measured, wherein a current value of the gradual portion B1 is varying with a ratio of the ion beam 112 projected on the workpiece 200. Reasonably, the gradual portion B1 is gradually decreasing when the overlap between the ion beam 112 and the workpiece 200 is gradually increasing. In other words, in other non-illustrated situation, other gradual value is gradually increasing when the overlap between the ion beam 112 and the workpiece 200 is gradually decreasing. In addition, referring to FIG. 2C and FIG. 3, when a relative position between the ion beam 112 and the workpiece 200 is as illustrated in FIG. 2C, whole the ion beam 112 is projected on the workpiece. In other words, the ion beam 112 is not crossing an edge of the workpiece 200 at this time, too. Hence, a lower portion C1 with a substantially stable current value equal to 0 as illustrated in FIG. 3 is measured. In the other embodiment, referring to FIG. 4A to FIG. 4C first, where the workpiece 200 is held by a chuck with two electrodes 122 and 124 positioned closed to two opposite sides of the chuck. Such design causes an equivalent capacitor formed between the electrode 122 and the edge of the workpiece 200, and also causes another equivalent capacitor formed between the electrode 124 and the edge of the workpiece 200. Thereafter, an original current may flow through the electrode 122, the workpiece 200 and the electrode 124 in sequences, so that a capacitance-related current may be measured by a capacitance meter electrically coupled with the workpiece 200. As usual, such design is used to monitor whether the workpiece 200 is held by the chuck exactly, because the capacitance of the two equivalent capacitors are dependent strongly on the relative positions between the workpiece 200 and the two electrodes 122/124. However, when at least partial ion beam 112 is projected on the workpiece 200, the electrons applied to neutralize the ion beam 112 (or the implanted workpiece 200), even the ions of the ion beam 112, may provide an additional current path to the original current, so that partial original current is grounded by the additional current path and then the measured capacitance-related current is degraded. Accordingly, the measured capacitance-related current will become a function of the relative position between the ion beam 112 and the workpiece 200, because at least the electrons for neutralizing are moved together with the ion beam 112. Therefore, when a relative position between the ion beam 112 and the workpiece 200 is as illustrated in FIG. 4A, whole the ion beam 112 is projected outside the workpiece 200 and then a higher portion A2 substantially equal to the original current as illustrated in FIG. 5 is measured. And then referring to FIG. 4B and FIG. 5, when a relative position between the ion beam 112 and the workpiece 200 is as illustrated in FIG. 4B, a part of the ion beam 112 is projected on the workpiece 200 and then a gradual portion B2 as illustrated in FIG. 5 is measured. Clearly, at this time, the measured capacitance-related current value is decreased when the overlap ratio is increasing and then the grounded current is increased, in the versa. In addition, referring to FIG. 4C and FIG. 5, when a relative position between the ion beam 112 and the workpiece 200 is as illustrated in FIG. 4C, whole the ion beam 112 is projected on the workpiece and then a lower portion C2 as illustrated in FIG. 5 is measured. Herein, the lower portion C2 is substantially equal to a difference between the original current and the maximized grounded current. Note that the profile in the present invention is not limited to have one or more gradual portions, also is not limited to be a U-shape profile, as illustrated in FIG. 3 and FIG. 5. Indeed, only the gradual portion corresponds to the situation that the ion beam is overlapped with an edge of the workpiece, and then the analysis on the gradual portion is more important and useful for monitoring how the workpiece is implanted by the ion beam. Hence, the number of the gradual portion and the contour of the profile may be flexibly adjusted. In a word, a higher portion, a decreasing gradual portion, a lower portion, an increasing gradual portion and another higher portion are measured in sequence in an ideal situation when the workpiece is scanned by the ion beam along a scan line. However, the ion beam may be interfered by a robot arm for moving the workpiece (or moving the chuck) through the ion beam or affected by other factor(s), so that the profile may have only a decreasing gradual portion or an increasing gradual increasing portion in practice. In some non-illustrated embodiments, when the workpiece is scanned by the ion beam back and forth, there are numerous profiles each corresponding to a scan path from a side of the workpiece through the workpiece surface to an opposite side of the workpiece one and only one time. In such case, an average profile, or an average gradual portion, acquired by averaging these profiles, or acquired by averaging these gradual portions, may be used as an equivalent profile (or equivalent gradual portion). Accordingly, by monitoring the gradual portion B1 and/or B2, which is generated during a period that the ion beam 112 is crossing an edge of the workpiece 200, the ion beam projected on the workpiece 200 may be monitored much real-timely and then some application(s) of the measured profile may be performed much real-timely. Of course, the application(s) of the measured shape is not limited by the number of the gradual portion or by whether the gradual portion is increasing or decreasing. Every application may be achieved by using only a gradual portion, or by using an average of at least two gradual portions. One application of the measured profile is converting the gradual portion to generate a current distribution of the ion beam 112 along an ion beam cross-section. FIG. 6A illustrates a relationship between the measured profile and the current distribution. As well-known, the current distribution current should have a peak close to the center of the ion beam cross-section area and is gradually decreased to the boundary of the ion beam cross-section. Hence, as the relationship shown in FIG. 6A, the measured profile corresponds to the higher portion when the ion beam cross-section is totally located outside the workpiece 200 (i.e., the ion beam 112 is totally projected outside the workpiece 200), the measured profile corresponds to the gradual portion when the ion beam cross-section is partially located inside the workpiece 200 (i.e., the ion beam 112 is moving through the edge of the workpiece 200), and the measured profile corresponds to the lower portion when the ion beam) cross-section is totally located inside the workpiece (i.e., the ion beam 112 is totally projected on the workpiece 200). Moreover, owing to the current distribution having gradual and non-linear shape, the variation of the gradual portion also has non-linear and gradual shape. Indeed, when the overlap ratio between the ion beam 112 and the workpiece 200 is gradually increased, the absolute value of the slope of the gradual portion is firstly increased and then decreased. Therefore, as shown in FIG. 6B, a current distribution may be converted from the measured profile, from at least a gradual portion of the measured profile. Accordingly, the ion beam current distribution among the cross-section of the ion beam 112 may be real-timely monitored during the implantation period of implanting the workpiece 200 by the ion beam 112 in a more real-timely manner. At most a short period of moving whole the ion beam 112 through the edge of the workpiece 200 is required to real-timely monitor the ion beam 12. In other words, after the ion beam 112 is provided before the implantation period, any variation of the ion beam 112 may be more real-timely monitored. To compare with these conventional approaches mentioned above, the application of the measured profile provides a more real-timely monitoring on the geometric messages of the ion beam. And then, both the adjustment on the ion beam before being implanted into the workpiece and these used parameters (such as the distance between the neighboring scan lines and the scan velocity along the scan path) may be efficiently monitored and adjusted. Another application of the measured profile is using the gradual portion to acquire the geometric messages of the ion beam 112. As discussed above and shown in FIG. 6C, the gradual portion is corresponding to the size of the ion beam 112 along a relative moving direction between the ion beam 112 and the workpiece 200. Therefore, by referring to FIG. 6C and FIG. 6D, when the relative moving direction is parallel to a minor axis of the ion beam 112 and the movement is overlapped a diameter D1 of the workpiece 200, an ion beam width W of the ion beam 112 may be acquired simply by measuring a span of the gradual portion. Alternatively, when the relative moving direction is parallel to a major axis of the ion beam 112 and the movement is overlapped with another diameter of the workpiece 200, an ion beam height H of the ion beam 112 may be acquired simply by measuring the span of the gradual portion along the used diameter. Another related application of the measured profile still is using the gradual portion to acquire the geometric messages of the ion beam 112. As shown in FIG. 6D, when the relative moving direction is parallel to a diameter of the workpiece 200 but the movement is not overlapped with the diameter (for example, the movement is overlapped with the line Lo parallel to the diameter Do), the variation of the measured profile is started before the ion beam 112 being overlapped with the cross point Po between Lo and the edge of the workpiece 200 and is not immediately finished after the ion beam 112 is not overlapped with the cross point Po. In other words, in such situation, the measured profile is a function of at least the size and the shape of the ion beam 112, the contour and the size of the workpiece 200 and the position of the line Lo. Accordingly, by scanning the workpiece 200 through the ion beam 112 along numerous lines respectively parallel to a direction (such as the direction parallel to the major axis of the ion beam 112) and numerous lines respectively parallel to another direction (such as the direction parallel to the minor axis of the ion beam 112), numerous different widths/heights corresponds to different lines are acquired. Therefore, as the converting mechanism introduced above (as shown in FIG. 6A and FIG. 6B), the contour of the ion beam 112 is calculated and then the ion beam width, the ion beam height and the center of the ion beam 112 on both directions may be acquired. After that, to compare with the initial contour or the ideal contour of the ion beam 112, an alignment between the ion beam 112 and the workpiece 200 may be adjusted, even the scan path of the ion bean 112 through the workpiece 200 may be adjusted, also the ion beam 112 may be adjusted by adjusting the operation of an ion source, a mass analyzer and/or a beam optics. To compare with these conventional approaches mentioned above, these applications of the measured profile provide a more real-timely monitoring on the current distribution on the ion beam cross-section. And then, the practical implanted dose on the implanted workpiece may be precisely monitored. A further application of the measured profile is monitoring a variation of the ion beam 112 and providing a channel to prevent/improve improper ion implantation. As shown in FIG. 6E and FIG. 6F, no matter by monitoring the converted ion beam current distribution or by measuring the measured profile, any monitored glitch indicates an abnormal variation is happed. For example, the ion beam source may be unstable, sot that the ion beam current distribution is unstable. For example, the moving velocity of the workpiece may be not stable, so that the ion beam current projected into the workpiece is suddenly changed. Hence, by monitoring the appearance of any glitch, whether the workpiece 200 is properly implanted by the ion beam 112 may be real-timely monitored during the implantation period. Therefore, the ion implantation may be paused to avoid improper implantation on the workpiece 200, and the ion implanter and/or the implanting parameters may be modified and/or corrected accordingly. To compare with these conventional approaches mentioned above, the application of the measured profile provides a channel to real-timely monitoring the ion beam and/or the ion implantation by analyzing the measured profile itself. And then, the ion beam and/or the ion implantation may be flexibly adjusted. An optional application of the measured profile is to improving the efficiency of the ion implantation. In the prior art, as shown in FIG. 6G, the turn around points 604 are predetermined and assigned to be the some because the real relative position between the workpiece 200 and the ion beam 112 usually can not be real-timely monitored during the implantation period of scanning the ion beam through the workpiece 200 along the scan path 603. Hence, to minimize the risk of tuning the relative movement between the ion beam 112 and the workpiece 200 early and then improperly implanting the ion beam 112 to the workpiece 200, the safety distance between the edge of the workpiece 200 and each turn around point 604 along each scan line of the scan path 603 usually is significantly larger than the ion beam width W and/or the ion beam height H. Then, for the upper portion and lower portion shown in FIG. 6G, the distance between the turn around points and the edge of the workpiece 200 is significantly larger the ion beam size. Therefore, the required time to scan through these safety distances is larger, and then the waste time, even the waste ion beam 112, is un-avoidably increased. In contrast, in this invention, the measured profile is generated by scan the ion beam 112 through the workpiece 200 along a scan line of the scan path 603, and then the currently measured signals corresponds to the location of the ion beam 112 should be only a portion of the measured profile no matter the ion beam 112 is located on which portion of the scan line. Indeed, at least one of currently measured signals is located on the gradual region when the ion beam 112 is partially overlapped with the workpiece 200, all currently measured signals are located on the higher portion when the ion beam 112 does not overlapped with the workpiece 200, and all currently measured signals are located on the lower portion when the ion beam 112 is totally overlapped with the workpiece 200. Therefore, to move scan the ion beam 112 along a fixed scanning direction when the ion beam 112 is overlapped with the workpiece 200 and to change the scanning direction or to move to another scan line after the ion beam 112 is not overlapped with workpiece 200 with shorted safety distance, it is possible to change a relative moving direction between the ion beam 112 and the workpiece 200 when all currently measured signals are located on the higher portion and to fix the relative moving direction when at least a currently measured signal is located on the gradual portion. Further, when the workpiece 200 is scanned by the ion beam 112 along numerous scanning lines over the workpiece 200, at least a turn around point of at least a specific scanning line may be assigned at a specific position wherein a distance between the specific position and an edge of the workpiece 200 along the specific scanning line is essentially equal to a span of the gradual portion. Accordingly, as shown in FIG. 6H, the safety distance between each turn around point and the edge of the workpiece 200 may be reduced to be almost equal to the ion beam size. And then, the waste time, even the waste ion beam 112, may be minimized. In addition, the measurement of the profile may meet some troubles in the real world. For example, the noise may induce the measured values at different relative positions oscillated, especially oscillates at the gradual portion where the measured value should be clearly changed among different relative positions. Moreover, sometimes, at least one relative position(s) has no proper measured value owing to improper measurement or other reasons. Therefore, in an embodiment, the measured profile may be modified by using an N-points smoothing process, wherein N is a positive integer, so that a modified profile has measured value(s) at all relative positions and then the above applications may be performed more easily and effectively. Furthermore, in another embodiment, when at least one specific relative position(s) has no proper measured signal, a correcting process may be performed to modify the measured profile. Herein, the correcting process may have at least one of following approaches. One approach is generating the profile by only these signals measured at other relative positions, i.e. any relative position with no proper sign are skipped. Another approach is generating a pseudo signal for each specific relative position by extrapolating from at least two signals measured at other relative positions, and then generating the profile by both these measured signals and the pseudo signal. One another approach is generating a pseudo signal for each specific relative position by interpolating from at least two signals measured at other relative positions, and then generating the profile by both these measured signals and the pseudo signal. The above embodiments are focused on how to real-timely monitor the ion implantation on a signal workpiece, also on some applications about how to efficiently handle the ion beam and adjust the ion implantation. However, some other embodiments may be used to more real-timely monitor the ion implantation of numerous workpieces, especially to monitor these workpieces by using only the monitoring results of these workpieces. As usual, a lot of workpieces may be implanted by using the same implantation parameters to form same implant region on each implanted workpiece. It is a natural requirement of the mass production. Therefore, after one or more of these workpieces are implanted and monitored by the above embodiments in sequence, at least the geometric messages of the ion beam acquired from these corresponding measured profile may be used to establish a reference. Then, during the implantation of the other workpieces from the same lot, the reference may be used to determine whether the practical implantation parameters; values are varied. Referring to FIG. 7 first, the method for monitoring ion implantation according to an embodiment of the present invention comprises the following steps. First, provide an ion beam and numerous workpiece (S710). Then, implant one or more the workpieces by the ion beam. For each implanted workpiece, a profile has numerous signals relevant to respectively numerous relative positions between the ion beam and the workpiece is generated and analyzed without referring to a pre-determined profile. Herein, each profile comprises a higher portion, a gradual portion and a lower portion (S720). Next, generate a reference being a function of these profiles corresponds to these implanted workpieces (S730). And then, implant another workpiece by the ion beam and generate another profile. Herein, the another profile has numerous signals relevant to respectively numerous relative positions between the ion beam and the another workpiece, also has a higher portion, a gradual portion and a lower portion (S740). After that, analyze the another profile without referring to the pre-determined profile to generate an analyzed result (S750). Finally, compare the analyzed result of the another profile with the reference (S760). Of course, all implantation parameters' values should be assigned initially before the step (710) and all these workpiece should be arranged to have the same implant result if each workpiece is implanted perfectly. Hence, the difference between the reference and the analyzed result of the another profile may be used to determine whether the practical implant result on the another workpiece is similar with the practical implant result(s) on these workpieces implanted before. However, in the real world, the implantation on these workpiece may be imperfect. The ion beam current may be varied during a period of implanting these workpieces, the motor used to drive each workpiece through the ion beam may be unstable during the period of implanting these workpiece, and so on. Therefore, after the Step (760), the embodiment may have some optional variations. Of course, all these optional variations may be used independently or mixed. One optional variation is repeating step (S740), step (S750) and step (S760) in sequence until all these workpieces are implanted by the ion beam. In other words, after one or more workpiece is used to build a reference, all other workpieces are implanted by the ion beam in sequence. Besides, for each workpiece implanted after the reference being built, a measured profile is acquired and analyzed as discussed in the above embodiments. Hence, for each workpiece implanted after the reference being built, by comparing the analyzed result of the corresponding profile with the reference, the practical implant result may be real-timely monitored without further analyzing the implanted workpiece after the ion implantation. For example, when both the reference and the analyzed result are the ion beam height, the ion beam width, the ion beam center (no matter the X-axis center or on the Y-axis center), the reference may be used to decide the acceptable ion beam shape/contour range. Then, whether the analyzed result of each workpiece implanted later is located in the acceptable ion beam shape/contour range provide a good index on the implant result of these workpieces implanted later. Another optional variation is repeating step (S740), step (S750) and step (S760) in sequence until a difference between the reference and the analyzed result of a specific workpiece is un-acceptable (such as the difference is larger than a threshold). In such condition, as usual, the ion beam current or the ion beam shape is significantly varied. Hence, the ion beam is tuned before implanting other workpiece(s). Note that the tuned ion beam may be not thoroughly equal to the original ion beam, although both are qualified enough to implant these workpieces. Hence, the step (S720) and the step (S730) may be performed again to update the reference, i.e., to generate a new reference. After that, steps (S740), (S750) and (S760) may be repeated in sequence for other workpieces not yet implanted. Of course, when a difference between the updated reference and the analyzed result of another specific workpiece un-acceptable, the updated reference may be updated again before other un-implanted workpieces are implanted later. Of course, there are other potential reasons may induce the un-acceptable difference. For example, the driving mechanism used to move and rotate the workpiece may by un-stably operated, and the alignment between the ion beam and the workpiece to be implanted may be varied. Hence, other similar variation may tune the operation of the driving mechanism or tune the alignment, even to tune one or more ion implantation parameters, but not only tune the ion beam. As an example, the implantation parameter may be at least one of the following: ion beam energy, ion beam direction, ion beam diverse, position of a workpiece right is implanted by the ion beam, and an alignment between the ion beam and a workpiece right is implanted by the ion beam, and so on. Still another optional variation is comparing the reference with a measured result measured by using a profiler to measure the ion beam, and then processed steps (S740), (S750), and (S760) in sequence only when a difference between the reference and the measured result is acceptable. The profile is a popular and exact method to measure the ion beam, but it may be not real-timely enough. Hence, the optional variation uses the profile to check whether the reference is exact enough, and then uses the reference to monitor the implantation on the other workpieces in a more real-timely manner. In these embodiment capable of monitoring numerous workpieces, the step (S720) and the step (S730) are similar with these embodiments capable of real-timely monitoring a workpiece. Hence, the details are not disclosed and discussed again, but only main contents may be used are described as the following: (a) The reference may be an average current distribution along an ion beam cross-section acquired by averaging one or more current distributions where each is converted from a gradual value of a profile. (b) The reference is an average ion beam geometric message acquired by averaging one or more ion beam geometric messages of these profiles, wherein the ion beam geometric messages may be an ion beam contour, an ion beam height, an ion beam width, and an ion beam center. (c) The ion beam width may be a span of a gradual portion of a profile when a relative moving direction between the ion beam and the workpiece coincides with a minor axis of the ion beam and crosses a diameter of the workpiece, the ion beam height may be a span of a gradual portion of a profile when the relative moving direction coincides with a major axis of an ion beam and crosses a diameter of a workpiece, and the ion beam contour and the ion beam center may be converted from numerous ion beam widths as a function of vertical position and numerous ion beam heights as a function of horizontal position which are acquired by scanning the workpiece through the ion beam along numerous first lines parallel to a specific diameter of the workpiece and along numerous second lines vertical to the specific diameter of the workpiece. (d) Use a Faraday cup close to the workpiece to measure numerous current values at these relative positions, so that at least one profile is formed as a current curve. (e) Use a capacitance meter electrically coupled with the workpiece to measure numerous capacitance-related current values at there relative positions, so that at least one profile is formed as a capacitance-related current curve. In summary, the invention proposes a method for monitoring ion implantation. During a period of scanning an ion beam through a workpiece, numerous signals are measured at different relative positions between the ion beam and the workpiece by a Faraday cup, a capacitance meter, or even other devices. Depends on the overlap ratio between the ion beam and the workpiece, the profile has at least a higher portion, a gradual portion and a lower portion. After that, by directly analyzing the profile without referring to a pre-determined profile, the ion implantation may be monitored in a much real-time manner. For example, by using the span of the gradual portion, and by using the relative motion messages, at least one of the ion beam width, the ion beam height, the contour and the center of the cross-section of the ion beam may be found out. For example, because the profile is a function of both the ion beam current implanted into the workpiece and the relative position between the workpiece and the ion beam, by converting the profile, an ion beam current distribution among the cross-section of the ion beam may be found. In addition, the ion beam may be adjusted or terminated when at least one glitch is appeared on the profile or the current distribution. Furthermore, the turn around point of each scanning line of a scan path may be adjusted flexibly and precisely by using ion beam width/width corresponding to the span of the gradual portion. Although specific embodiments of the present invention have been described, it will be understood by those of skill in the art that there are other embodiments that are equivalent to the described embodiments. Accordingly, it is to be understood that the invention is not to be limited by the specific illustrated embodiments, but only by the scope of the appended claims.
056569025	claims	1. A robot comprising: a motor chamber; first and second arm member assemblies, each supported by the motor chamber at a first end thereof for circumferential movement about a common axis; the first and second arm member assemblies each having a second end coupled to a common robot element; and a drive mechanism mounted in the motor chamber, the drive mechanism having drive members, each coupled to a respective one of the first and second arm member assemblies to controllably move the first ends of the first and second arm assemblies: a robot chamber; a motor chamber comprising a side wall structure extending within the confines of the robot chamber and being in an airtight relation in respect to the robot chamber; first and second arm member assemblies, each supported at a first end thereof by the side wall structure of the motor chamber within the robot chamber for circumferential movement about a common axis; the first and second arm member assemblies each having a second end coupled to a common robot element within the robot chamber; and a drive mechanism mounted within the motor chamber and having drive members, each coupled through the side wall structure of the motor chamber to a respective one of the first and second arm member assemblies to controllably move the first ends of the first and second arm assemblies: a robot chamber; a motor chamber comprising a wall structure extending within the confines of the robot chamber and being in an airtight relation in respect to the robot chamber; first and second arm member assemblies within the robot chamber, each supported at a first end thereof by the motor chamber for circumferential movement about a common axis; the first and second arm member assemblies each having a second end coupled to a common robot element within the robot chamber; and a drive mechanism mounted within the motor chamber and magnetically coupled through the wall structure of the motor chamber to the first and second arm member assemblies to controllably, selectively move the first ends of the first and second arm assemblies: 2. A robot comprising: 3. A robot comprising:
	abstract	Systems and methods for providing a molten salt reactor can include a graphite reactor core that defines an internal space, with multiple fuel wedges being received in the internal space, and with the wedges each defining a fuel channel extending from a first end to a second end of each of the wedges. The reactor can further include a fuel pin rod that defines an internal fuel conduit and that is disposed between at least two of the wedges. The reactor core can also define a fuel ingress port and a fuel egress port. The reactor core can further be rotatably received within a reactor housing such that the ports are configured to become at least one of more occluded and less occluded as the reactor core rotates.
	claims	1. A treating method for treating a radioactive liquid waste containing a surface active agent, comprising the steps of: adding hydrogen peroxide to said radioactive liquid waste containing said surface active agent; charging ozone to said radioactive liquid waste while heating said radioactive liquid waste to a temperature of 50xc2x0 C. or higher, thereby forming hydroxy radicals in said radioactive liquid waste by decomposition of said ozone and said hydrogen peroxide; and decomposing the surface active agent by said hydroxy radicals. 2. A treating method as defined in claim 1 , wherein said ozone is charged in a plurality of aeration vessels supplied with said radioactive liquid waste, in which said ozone is charged into one of the aeration vessels and then ozone discharged therefrom is charged into another aeration vessel. claim 1 3. A treating method as defined in claim 2 , wherein bubbles of ozone charged in said aeration vessels are broken. claim 2 4. A treating method as defined in claim 3 , further comprising a step of removing a solid component contained in said radioactive liquid waste after said decomposing step. claim 3 5. A treating method as defined in claim 2 , wherein an alkali solution is added to said radioactive liquid waste and pH of said radioactive liquid waste supplied to said aeration vessel is adjusted to 7 or higher. claim 2 6. A treating method as defined in claim 5 , further comprising a step of removing a solid component contained in said radioactive liquid waste after said decomposing step. claim 5 7. A treating method as defined in claim 2 , further comprising a step of removing a solid component contained in said radioactive liquid waste after said decomposing step. claim 2 8. A treating method as defined in claim 1 , wherein an alkali solution is added to said radioactive liquid waste. claim 1 9. A treating method as defined in claim 8 , further comprising a step of removing a solid component contained in said radioactive liquid waste after said decomposing step. claim 8 10. A treating method as defined in claim 1 , further comprising a step of removing a solid component contained in said radioactive liquid waste after said decomposing step. claim 1 11. A treating method for treating a radioactive liquid waste containing a surface active agent, comprising the steps of: charging ozone to a radioactive liquid waste containing an organic substance including the surface active agent while hydrogen peroxide is present and is dissolved in said radioactive liquid waste, and while heating said radioactive liquid waste to a temperature of 50xc2x0 C. or higher; forming hydroxy radicals in said radioactive liquid waste by decomposition of said ozone and said hydrogen peroxide; and decomposing said surface active agent by said hydroxy radicals.
	summary
	claims	1. An electrical power source comprising a semiconductor, the semiconductor comprising a diamond material and a radioactive source embedded within the diamond material, wherein the radioactive source comprises a beta-emitting radioisotope and atoms of the radioisotope are substitutionally or interstitially integrated into the diamond material,wherein the diamond material comprises a plurality of regions in the form of layers within a continuous crystal lattice of the diamond material, andwherein at least one layer of the diamond material comprises the radioactive source and at least one layer of the diamond material does not comprise the radioactive source. 2. The electrical power source of claim 1, wherein the radioactive source embedded within the diamond material is formed of one or more of tritium, 14C, 10Be and phosphorus-33. 3. The electrical power source according to claim 1, wherein the diamond material has a layered structure with at least one layer comprising the radioactive source and at least one layer which does not comprise the radioactive source. 4. The electrical power source according to claim 1, wherein the radioactive source is provided in a layer of diamond having a thickness in a range 50 nanometres to 150 micrometres. 5. The electrical power source according to claim 1, wherein the diamond material includes a 13C diamond region which comprises isotopically purified diamond material having an increased 13C content compared to natural isotopic abundance. 6. The electrical power source according to claim 5, wherein the 13C diamond region is in the form of a layer having a thickness in a range 2 nanometres to 2 millimetres. 7. The electrical power source according to claim 5, wherein the 13C diamond region has an atomic concentration of 13C of at least 2%, 3%, 4%, 5%, 10%, 20%, 50%, 75%, 85%, 95%, 99%, or 99.9%. 8. The electrical power source according to claim 1, wherein the diamond material includes a 12C diamond layer comprising a boron-doped 12C diamond layer. 9. The electrical power source according to claim 8, wherein the 12C diamond layer has a thickness in a range 200 nanometres to 2 millimetres. 10. The electrical power source according to claim 1, wherein the diamond material includes a tri-layer structure comprising a layer of 14C containing diamond, a layer of 12C diamond, and a layer of 13C diamond. 11. The electrical power source according to claim 1, wherein the diamond material has a layered structure comprising one or more layers of the diamond material, and wherein at least one layer of the one or more layers comprises an isotopic layer within the diamond material. 12. The electrical power source according to claim 1, wherein the diamond material has a single substitutional nitrogen concentration of no more than 5 ppm, 1 ppm, 500 ppb, 300 ppb or 100 ppb in at least one region thereof. 13. The electrical power source according to claim 1, wherein the diamond material in which a radioactive source is embedded is a synthetic diamond material in which radioisotope atoms are integrated during formation of the synthetic diamond material.
	claims	1. A method of driving plasma ions and electrons in a field reversed configuration (FRC) magnetic field comprising the steps ofgenerating an FRC about an rotating elongate annular layer of plasma of ions and electrons axially extending within a cylindrical chamber along the longitudinal axis of the chamber, the plasma having a density of 1014 per cubic centimeters or more and wherein the ions in the rotating layer of plasma orbit in betatron orbits normal to the longitudinal axis of the chamber,creating an electric potential wave that penetrates the rotating layer of plasma and rotates in the same direction as the azimuthal velocity of ions in the rotating layer of plasma, wherein the electric potential wave has a wavelength one or more orders of magnitude greater than the radius of the chamber,trapping ions in the rotating layer of plasma in the electric potential wave, andincreasing the momentum and energy of the trapped ions. 2. The method of claim 1 wherein the step of creating an electric potential wave includes energizing a plurality of elongate electrodes forming a cylindrical surface within the chamber. 3. The method of claim 2 wherein the plurality of elongate electrodes form an elongate cyclotron. 4. The method of claim 2 wherein the cyclotron is a quadrupole cyclotron. 5. The method of claim 2 wherein the cyclotron is a dipole cyclotron. 6. The method of claim 1 further comprising the step of injecting neutral atoms into the plasma.
049869558	description	DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows a fuel assembly comprising This fuel assembly comprises of spacer grids 2 spaced in the longitudinal direction of the assembly guide tubes 3, to which the grids 2 are rigidly fixed, an upper joining piece 4 and a lower joining piece 5 which are fixed to the end of the guide tubes 3. The fuel rods or fuel rods of the assembly which are shorter than the guide tubes 3 are disposed in the framework so as to form a bundle in which the rods are disposed parallel to one another. The rods are held laterally by the spacer grids 2 so as to form a uniform network with squared mesh in the transverse sections of the assembly. The spacer grids 2 also retain the fuel rods 6 in their longitudinal direction by gripping means disposed at the level of each of the cells receiving a fuel rod and consisting of bosses projecting inwards in the cell and of resilient springs pressing the fuel rod with a certain pressure against the bosses. The spacer grids 2 comprise an outer frame of generally square form, consisting of small plates assembled together at their corners. The fuel rods disposed against the inner surface of the small plates of the frame of the spacer grids form the peripheral rods, such as the rod 6a. FIG. 2 shows the removal device 10 which makes it possible to remove the peripheral rods of a fuel assembly 1 disposed inside a storage pool for fuel assemblies, even if these peripheral fuel rods have several successive pieces in their length. The fuel assembly 1 rests, by means of its lower joining piece 5, on a support 11, the positioning of the assembly in a vertical position in which the rods are removed being facilitated by an insertion device which is flared towards the top 12, fixed on the base 11 and which receives the lower part of the assembly. A vessel 13 for recovering pieces of fuel rods 6'a is placed on the support 11, near the assembly 1. The upper part of the assembly is at a depth, below the upper level 14 of the pool, which is substantially equal to three meters, this depth of water ensuring effective biological protection for an operator 15 performing the removal operations from the platform 16 of a control station 18 fixed on the upper part of a lateral wall 19 of the storage pool. The device according to the invention comprises a rod 20 of great length disposed vertically in the vicinity of the wall 19 of the pool and connected, at its upper part 21, to a vertical support which is integrally attached to the platform 16. A carriage 22 is mounted so as to be movable in the longitudinal direction of the rod 20 and may be displaced in this longitudinal direction by a winch 23 which may be activated by a handle used by the operator 15. By means of a displacement device 24, the carriage 22 carries a support 25 on which is fixed the tool for removing the peripheral fuel rods of the assembly 1 and at least one video camera 26. The removal tool is remotely controlled by remote control means 28 of known type, such as ball remote controls comprising a sheath inside which a flexible element is mounted so as to be movable, and which makes it possible to provide a pulling or pushing action which is remotely controlled, for example manually, by a wheel 29 fixed on the railing of the platform 16 of the control station 18. Ball remote controls 30 and 31 also make it possible to remotely control the displacements of the tool support 25 by means of the displacement device 24 and to ensure the orientation control of the video cameras, such as 26, respectively. In FIG. 2, the carriage 22 and the displacement device 24 are in a position which makes it possible to remove a peripheral rod 6a from the assembly 1, the upper joining piece 4 of the assembly being removed. The device for the support and positioning of the tool support 25 is the subject of a patent application filed jointly by the company FRAMATOME and the company COGEMA on the same day as the present patent application. FIGS. 3, 4 and 5 show the displacement device 24 of the tool support 25 which consists of a crossed carriage displacement assembly. This displacement assembly is fixed on the carriage 22 so as to be movable in the vertical direction and comprises a first carriage, or lower carriage, which is movable in a first horizontal direction substantially perpendicular to the wall 19 of the pool, on which a second carriage, or upper carriage, carrying the support device 25, is mounted so as to be movable in a second horizontal direction substantially parallel to the wall of the pool 19. The vertical displacement carriage 22 and the crossed carriage assembly 24 make it possible to place the support 25 and the removal tool carried by this support at the level of a zone of a piece of rod to be removed located between two spacer grids, according to which the piece of rod is grasped and gripped in order for it to be removed from the framework of the assembly. The displacements of the crossed carriage assembly 24 are ensured by ball remote controls 30 whose upper part is accessible from the control station 18. A video image of the work zone provided by a camera such as 26, enables the operator to place the support 25 and the tool for removing the rods in the desired position. The removal tool comprises a body 32 fixed on a bracket 33 forming one of the elements of the tool support 25. The body 32 of the tool will be described in greater detail with reference to FIG. 10. A shaft 35 is fixed on a part of the body 32 by means of a support 36 and of screws 37 and 38. The shaft 35 is fixed in a vertical position, i.e., a position parallel to the axis of the rod 20 which corresponds to the direction of displacement of the carriage 22. The part 40 of the removal tool which grasps and grips the fuel rods 6a located at the periphery of a fuel assembly consists of tongs 41 which are visible on a larger scale in FIGS. 6 and 7. The tongs 41 comprise two arms 43 and 44 which are articulated, at one of their ends, on the shaft 35. As may be seen, in particular, in FIG. 3, the arm 43, or female arm, has the form of a rectangle in which the second arm 44, or male arm, fits in the complete closure position of the tongs shown in FIG. 7. The tongs 41 also comprise a guide plate 45 rigidly fixed to the shaft 35 above the female arm 43 by means of a pin 46. The two parallel parts of the rectangle forming the female arm 43 of the tongs 41 are traversed, in their part which is situated below the guide plate 45, by oblong openings 47 inclined relative to the longitudinal axis of the arm 43, in the direction of the male arm 44. Similarly, the male arm 44 is traversed by an oblong opening 48 inclined relative to its longitudinal axis and directed towards the female arm 43. The oblong openings 47 and 48 are superposed, at their end part, so as to provide a bore which permits the passage of a spindle 50 for positioning the tongs perpendicular to the arms 43 and 44. The guide plate 45 comprises an oblong cavity 49 on its lower surface in a longitudinal direction corresponding to the direction of the axis 51 of symmetry of the tongs 41. At its upper part, the cavity 49 covers the part common to the oblong apertures 47 and 48 forming the bore for the passage of the positioning spindle 50. The positioning spindle 50, which may be seen in 6A and 6B, comprises a central part 50a in the form of a smooth rod, a threaded end part 50b and a head 50c having two flats 52, as may be seen in FIG. 6B. The head 50c of the spindle 50 is engaged in the cavity 49 of the guide plate 45. Two washers 54 and 54' are placed around the smooth part 50a of the spindle 50, in position in the oblong apertures in the female arm 43 of the tongs, and, on the threaded end 50b of the spindle 50 projecting relative to the lower part of the arm 43 are engaged a fixing nut 53 and a counternut 53'. The positioning spindle 50 is thus held in the oblong openings 47 and 48 and in the cavity 49 whilst remaining movable inside these openings and cavities during the maneuvering of the tongs. This permits accurate positioning of the male arm 44 and of the female arm 43, the longitudinal axes of which remain perfectly symmetrical relative to the axis 51 of symmetry of the tongs during the displacements of the tongs between the open position thereof shown in FIG. 6 and the completely closed position thereof shown in FIG. 7. The positioning spindle 50 comes up against the end of the apertures 47 and 48 in the completely open position of the tongs shown in FIG. 6, which makes it possible perfectly to define this open position and the orientation of the arms 43 and 44 on the support 25. In the completely open position of the tongs shown in FIG. 6, the male arm 44 is exactly parallel to a row of peripheral rods 6a of the assembly. The arms 43 and 44 comprise end jaws 57 and 58, respectively, in the vicinity of the articulation shaft 35. In the open position of the tongs shown in FIG. 6, the distance between the inner surfaces of the jaws 57 and 58 is slightly greater than the diameter of a rod 6a. On the other hand, the end parts of the jaws 57 and 58 are produced in the form of thin flats which may be inserted between two successive rods 6a, as may be seen in FIG. 4, in which the tongs 41 are shown in an open position, as in FIG. 6. At its end opposite to the jaw 57 of the tongs, the female arm 43 is traversed, in its two parallel parts, by openings 59. Similarly, at its end opposite to the jaw 58, the male arm 44 is perforated by a circular opening 60. A cover 41, which may be seen, in particular, in FIGS. 4 and 5, comprises two arms placed on either side of the female arm 43 and comprises openings permitting the articulated mounting of the cover 61 on the end part of the female arm 43 by means of spindles 62 engaged in the openings 59 in the female arm 43 and immobilized by pins in the cover 61. The cover 61 comprises a threaded hole in its end part, inside which is screwed the threaded end 64a of a sheath 64 of a ball remote control whose flexible element 65, mounted so as to be movable in translation by the balls in the sheath 64, is connected, at its threaded end 65a to a nut 66 engaged in the opening 60 in the male arm 44 of the tongs 41. FIG. 9 shows the end of the ball remote control 28 located opposite the gripping tongs 41 at the level of the control station 18 of the removal device shown in FIG. 2. The end of the sheath 64 is fixed on the support 66 of a displacement device of the movable flexible element 65 of the ball remote control. The body 66 of the displacement device is fixed by means of a bracket 67 and two half-flanges 68 and 69 which are clamped together by screws 70 on the railing of the platform 16 of the control station 18. A screw 72, one end of which is integrally attached to the wheel 29, is placed inside the housing 66 and makes it possible to drive a carriage 73 in translation along the length of the casing, on which carriage is fixed, by means of nuts 73', the end of the flexible element 65 of the ball remote control. The displacement of the flexible element 65 of the ball remote control inside the sheath 64 makes it possible to close the tongs on a rod 6a after these tongs have engaged the rod, as may be seen in FIG. 4. A stop 74 makes it possible to arrest the movement of the screw 72 when the displacement of the arms of the tongs makes it possible to obtain the complete closure of these tongs, as may be seen in FIG. 7. In this position, the rod 6a may be slightly crushed between the jaws 57 and 58 of the tongs, as may be seen in FIG. 8. This deformation makes it possible to improve the grasping force of the tongs on the rod during extraction. However, it is necessary to avoid too great a deformation, which could lead to a breakage or tearing of the rod 6a. In the completely closed position of the tongs, as may be seen in FIG. 7, the positioning spindle 50 has stopped at the end of the oblong apertures 47 and 48, which coincide. FIG. 10 shows a pushing device which may be mounted on the body 32 of the removal tool in order to push a fuel rod which has broken inside a spacer grid. The piece of rod must be pushed outside the spacer grid in order to be seized by the tongs 41. The body 32 of the tool comprises a central part 32a on which is mounted, in the vertical direction, a slide bush 75 comprising two parts, in each of which is mounted a ball bush 76. A shaft 77 is mounted so as to slide in the bush 75 and comprises a central part on which a rack 77' is machined. A pushing fork 78 is fixed by means of screws 79 to the central part of the sliding shaft 77. The fork 78 carries, at its outer end opposite to the shaft 77, a thrust device 80 whose diameter corresponds substantially to the outer diameter of the sheath of a rod 6a. Support forks, such as 81, are placed above and on either side of the pushing fork 78. The support forks 81, whose shape may be seen in FIG. 11 with reference to an alternative embodiment, comprise, at their end, inner recesses which may be engaged over the peripheral rods 6'a located on either side of the rod 6a on which pushing and removal is performed. The flexible element 82 of a ball remote control, whose sheath 83 is fixed on a part of the support 25 by means of an arm 84 and a bracket 85, is connected to an end part of a rack 86 mounted so as to be movable in the central part 32a of the body 32 and so as to engage with a pinion 87 whose rotation results in the rotation of a pinion 88 and, by means of a second pinion 89, engaging with the rack 77', the vertical displacement of the sliding shaft 77 and of the pushing fork 78. The displacement of the movable flexible element 82 of the ball remote control in the direction of the pushing action results in a downward displacement of the sliding shaft 77, of the pushing fork 78 and of the thrust device 80. The ball remote control 82, 83 may be activated from the control station 18 by the operator 15. FIG. 11 shows an alternative embodiment of the pushing device in FIG. 10, the pushing fork 78' disposed between the support forks 81 comprising a recess 90 whose diameter is slightly greater than the diameter of the sheath of a pencil and an inner recess 91 at its end which is capable of engaging on the outer surface of a fuel rod 6 situated on the first inner row of fuel rods opposite the rod 6a being removed. The pushing fork 78' is controlled in the same manner as the pushing fork 78 in order to push on the end of a rod projecting relative to a spacer grid. In order to implement the removal device according to the invention, the operator displaces the carriage 22 and the crossed carriage device, so as to place the jaws of the tongs in an open position on either side of the rod or of the piece of pencil 6a to be removed, as may be seen in FIG. 4. The operator then activates the wheel 29 of the remote control 28 in order to obtain the desired gripping of the tongs on the rod. The tongs are then in their position shown in FIG. 8. If the spindle of the tongs is mounted on a sliding shaft, such as the shaft 77, which may be displaced by a remote control and a rack and pinion assembly, it is possible to measure, from the control station, the removal force on the rod by a dynamometric device pushing on the flexible element of the ball remote control for displacing the shaft 77 situated at the level of the control station 18. The use of dynamometric springs makes it possible directly to measure the pushing force necessary to displace the pencil or the piece of rod to be removed. If the removal force does not exceed a predetermined value, the rod or the piece of rod is then completely removed by raising the displacement assembly 24 along the rod 20 by the carriage 22 driven by the winch 23. When the piece of rod has been completely removed, the column 20 is displaced laterally and then the tongs are opened so that the piece of rod 6'a is recovered in the vessel 13. When the piece of rod to be removed is longer than the gap separating two grids, the rod is first slid on the side where the previous piece of rod has been removed, in order to completely release one of the end parts of the rod from the corresponding grid. The crossed carriage device is then used to pull the rod slightly outwards from the assembly grid and it is then removed by displacing the carriage 22 in the vertical direction. The rod is pulled in an oblique direction relative to its axis of origin, but the flexion angle of this rod is sufficiently small for the flexibility of the sheath to be able to permit removal. Use may also be made of a device comprising means for support and displacement consisting of two crossed carriage assemblies. A first crossed carriage assembly is equipped with the removal tool comprising the gripping tongs. The second crossed carriage device is equipped with a tool module comprising cutting tongs which make it possible to cut the sheath of the rod. The work operation consists, by using the crossed carriage device equipped with the cutting tongs tool module, in cutting the rod at a sufficient height above a grid so that it can be removed at a later stage. The piece of cut rod is held with the gripping tongs and the piece of rod held by the tongs is cut at a short distance from the second spacer grid. The piece of rod disposed between the two grids is released from the assembly and deposited in the vessel. It is thus possible successively to remove various parts of rod between two successive spacer grids. The mounting of the cutting tongs tool module may be designed so as to come into position vertically above the removal tongs while maneuvering the crossed carriage device, which makes it possible to work on the same piece of rod with the cutting tongs and with the gripping tongs. As explained hereinabove, if a fuel rod is broken inside a spacer grid, it is necessary to push this piece of rod so that it can be seized by the tongs. This removal operation may be implemented by using a support comprising two crossed carriage devices. One of the crossed carriage devices is equipped with the pushing tool shown in FIG. 10. The second crossed carriage device is equipped with the removal tool comprising the gripping tongs, shown in FIGS. 3, 4 and 5. The device and the method according to the invention thus make it possible to remove all the peripheral rods of a fuel assembly which have suffered any damage which may lead to a complete breakage. Dismantling of the fuel assemblies is thus avoided, which assemblies may be reused in the core of the reactor. The operating costs of the plant are thus reduced. The gripping tongs may have a form different from that which has been described, these tongs being activated remotely by a control device other than a ball remote control. It is also possible to use any form of tongs comprising displacement adjustment means and immobilizing means of any type. It is also possible to use any related pushing or pulling device in order to remove pieces of rod in any case which may arise. The invention also applies in the case of all fuel assemblies for light-water reactors consisting of bundles of parallel fuel rods held in a framework.
	summary
	abstract	A particle beam therapy system comprises a charged particle beam generator for generating a charged particle beam, two or more treatment rooms provided with respective irradiation devices for irradiating the charged particle beam, a beam line for transporting the charged particle beam extracted from the charged particle beam generator to the irradiation device in selected one of the two or more treatment rooms, a beam detection processing/control unit for monitoring a beam state of the charged particle beam in one of the two or more irradiation devices, and a selector for switchably selecting one of the irradiation devices which is to be monitored by the beam detection processing/control unit. The selector is controlled such that the selector establishes connection with the irradiation device in the selected one treatment room to which the charged particle beam is transported through the beam line. The system configuration can be simplified while maintaining the operation efficiency.
	claims	1. A control system for a radiation therapy system comprising a moveable patient positioner and a moveable radiation nozzle, wherein the control system generates control signals to induce movement of at least one of the patient positioner and the radiation nozzle and wherein the control system further receives position signals from a plurality of external measurement devices, each of the plurality of external measurement devices being arranged to independently provide position information indicative of a current spatial position of one or more of the patient positioner, of the radiation nozzle, and of fixed reference objects and wherein the control system calibrates the position signals from each of the plurality of external measurement devices with each other and to the fixed reference objects and wherein the control system further generates the control signals so as to induce at least one of the patient positioner and the radiation nozzle to move to achieve a desired translational and rotational alignment of the patient positioner with respect to the radiation nozzle. 2. The control system of claim 1, wherein the control system comprises:a motion control module that generates the control signals;a 6-D module that determines 6-dimensional spatial positions of at least the patient positioner and the radiation nozzle with respect to the fixed reference objects; anda command and control module in communication with the motion control and 6-D modules. 3. The control system of claim 2, further comprising a patient registration module in communication with the command and control module wherein the patient registration module determines a registration transformation between a computer tomography orientation and a treatment session set-up pose. 4. The control system of claim 1, wherein the system further receives position signals from the external measurement devices indicating a current position of one or more movable imagers that can be arranged to obtain image data of at least a portion of the patient positioner and wherein the system determines a position of the one or more imagers with respect to the fixed reference objects. 5. The control system of claim 4, wherein the control system receives position signals from the one or more moveable imagers indicative of a spatial position of a target iso-center of a patient affixed to the patient positioner and wherein the control system generates the control signals so as to induce the patient positioner to move so as to align the target iso-center at a desired translational and rotational position. 6. The control system of claim 1, wherein the control system receives position signals from two or more external measurement devices arranged to provide independent position information for each of the patient positioner and the radiation nozzle respectively from multiple perspectives. 7. The control system of claim 1, wherein the position signals from the plurality of external measurement devices comprise direction vectors and wherein the control system calculates intersections of pluralities of the direction vectors. 8. The control system of claim 7, wherein the control system further determines a translational location and rotational orientation of each of the patient positioner and of the radiation nozzle based at least in part on the calculated intersections of the direction vectors. 9. The control system of claim 1, wherein the system further receives independent secondary local position signals indicative of the current position of the patient positioner and of the radiation nozzle and wherein the system generates the control signals as a function of both the position signals from the external measurement devices and from the secondary position signals. 10. A method of determining positions of movable components of a radiation therapy system, the method comprising:externally measuring positions of a plurality of fixed monuments;defining a fixed frame of reference based on the plurality of fixed monuments;externally measuring position of one or more monuments of at least one movable component of a radiation therapy system;calibrating the measured position of the one or more monuments of the at least one movable component with respect to the defined fixed frame of reference; andcalculating a translational location and rotational orientation of the at least one movable component with respect to the fixed frame of reference. 11. The method of claim 10, comprising externally measuring the position of the one or more monuments of the at least one movable component via two or more external measurement devices so as to perform a binocular measurement. 12. The method of claim 10, wherein the externally measuring the position of the one or monuments of the at least one movable component comprises performing direction measurements so as to determine a plurality of direction vectors. 13. The method of claim 12, comprising calculating intersections of the plurality of direction vectors and wherein calculating the translational location and rotational orientation is based at least in part on the calculated intersections. 14. The method of claim 10, comprising:externally measuring positions of one or more monuments of a plurality of movable component of the radiation therapy system; andcalibrating the measured position of the one or more monuments of the plurality of movable components with respect to each other and with respect to the fixed frame of reference. 15. The method of claim 10, further comprising generating movement commands to induce the at least one movable component to move to a desired translational location and rotational orientation. 16. The method of claim 15, further comprising:generating first movement commands to induce the at least one movable component to move to a first desired translational location and rotational orientation with respect to the fixed frame of reference;externally remeasuring position of the one or more monuments of the at least one movable component after movement according to the first movement commands occurs; andrecalculating the translational location and rotational orientation of the at least one movable component with respect to the fixed frame of reference. 17. The method of claim 16, further comprising:generating second movement commands to induce at least second movable component to move to a second desired translational location and rotational orientation with respect to the fixed frame of reference;externally remeasuring position of the one or more monuments of the at least second movable component after movement according to the second movement commands occurs; andcalculating the translational location and rotational orientation of the at least second movable component with respect to the fixed frame of reference. 18. The method of claim 10, further comprising independently performing at least second local position measurements of the at least one movable component.
	claims	1. A reactor vessel handling method comprising the steps of reinforcing an overhead traveling crane installed above a reactor vessel in a reactor containment vessel of a pressurized water reactor containing said reactor vessel by supporting said overhead traveling crane with reinforcing members installed in said reactor containment vessel, lifting said reactor vessel from said reactor containment vessel up to an operating floor in said reactor containment vessel by using the reinforced overhead traveling crane, tilting down said reactor vessel at said operating floor, and then carrying out said reactor vessel in the state of being tilted-down through an opening provided in a side wall of said reactor containment vessel to the outside of said reactor containment vessel. 2. A reactor vessel handling method according to claim 1 , wherein an equipment carrying-in opening in said reactor containment vessel is enlarged and utilized as said opening. claim 1 3. A reactor vessel handling method comprising the steps of reinforcing an overhead traveling crane installed above a reactor vessel in a reactor containment vessel of a pressurized water reactor containing said reactor by supporting said overhead traveling crane with reinforcing members installed in said reactor containment vessel, lifting said reactor vessel from said reactor containment vessel up to an operating floor in said reactor containment vessel by using the reinforced overhead traveling crane, tilting down said reactor vessel at said operating floor, and carrying out said reactor vessel in the state of being tilted-down through an opening provided in a side wall of said reactor containment vessel to the outside of said reactor containment vessel; and then carrying in a new reactor vessel in a state of being tilted-down to said operating floor within said reactor containment vessel through said opening, and moving said new reactor vessel in a state of being tilted-up from said operating floor to said installed position by using said reinforced overhead traveling crane. 4. A reactor vessel handling method comprising the steps of, in a state in which an overhead traveling crane installed above a reactor vessel in a reactor containment vessel of a pressurized water reactor containing said reactor vessel is reinforced by supporting the overhead traveling crane with reinforcing members installed in said reactor containment vessel, carrying in a new reactor vessel in a state of being tilted-down to said operating floor within said reactor containment vessel through an opening provided in a side wall of said reactor containment vessel, and moving said new reactor vessel in a state of being tilted-up from said operating floor to an installed position of said new reactor vessel in said reactor containment vessel by using said reinforced overhead traveling crane. 5. A reactor vessel handling method according to claim 4 , wherein said reinforcing members for said overhead traveling crane are removed after moving said new reactor vessel to said installed position. claim 4 6. A reactor vessel handling method according to claim 1 , wherein said reinforcing members are installed to said operating floor. claim 1 7. A reactor vessel handling method according to claim 1 , wherein said tilting-down of the reactor vessel is carried out such that the reactor vessel is tilted along with a radiation shield surrounding said reactor vessel. claim 1 8. A reactor vessel handling method according to claim 1 , wherein said tilting-down of the reactor vessel is carried out on a car set up on said operating floor, and said carrying-out of the reactor vessel is carried out by using said car. claim 1 9. A reactor vessel handling method according to claim 8 , wherein said tilting-down of the reactor vessel is carried out such that the reactor vessel is tilted along a radiation shield surrounding said reactor vessel. claim 8 10. A reactor vessel handling method according to claim 9 , wherein said tilting-down of the reactor vessel is carried out by engaging a shaft attached to said radiation shield as a center of rotation with a shaft bearing member provided on said car. claim 9 11. A reactor vessel handling method according to claim 3 , wherein said reinforcing members are installed on said operating floor. claim 3 12. A reactor vessel handling method according to claim 3 , wherein said tilting-down of the reactor vessel is carried out such that the reactor vessel is tilted along with a radiation shield surrounding said reactor vessel. claim 3 13. A reactor vessel handling method according to claim 3 , wherein said tilting-down of the reactor vessel is carried out on a car set up on said operating floor, and said carrying-out of the reactor vessel is carried out by using said car. claim 3 14. A reactor vessel handling method according to claim 13 , wherein said tilting-down of the reactor vessel is carried out such that the reactor vessel is tilted along with a radiation shield surrounding said reactor vessel. claim 13 15. A reactor vessel handling method according to claim 14 , wherein said tilting-down of the reactor vessel is carried out by engaging a shaft attached to said radiation shield as a center of rotation with a shaft bearing member provided on said car. claim 14 16. A reactor vessel handling method according to claim 3 , wherein said carrying-in of the new reactor vessel is carried out by laying said new reactor vessel on said car. claim 3 17. A reactor vessel handling method according to claim 4 , wherein said reinforcing members are installed on said operating floor. claim 4 18. A reactor vessel handling method according to claim 4 , wherein said carrying-in of the new reactor vessel is carried out by laying said new reactor vessel on a car. claim 4
	summary
	claims	1. A method for obtaining a shielding element for minimizing the penumbra of a scanning hadron pencil beam outside a target area, said hadron beam being scanned over the target area by an irradiation unit, said beam having a width (σ), the method comprising:(i) defining a closed or open contour of said target area;(ii) providing a block having a longitudinal thickness configured to block the passage of said hadron pencil beam and having a lateral surface perpendicular to said longitudinal thickness;(iii) forming an aperture of a shape similar to said contour of said target area, the aperture extending the length of the longitudinal thickness of said block in order to let through said hadron pencil beam, said aperture forming a longitudinal internal surface; and(iv) trimming said block so as to form a longitudinal external surface around said longitudinal internal surface, said longitudinal internal and longitudinal external surfaces delimiting a side wall, the side wall having a side thickness everywhere greater than at least once the width (σ) of the hadron pencil beam, and, in at least one location, less than five times the width (σ) of the hadron pencil beam. 2. The method for obtaining a shielding element according to claim 1, wherein the longitudinal external surface is similar to said longitudinal internal surface. 3. The method for obtaining a shielding element according to claim 1, wherein the method further comprises forming a support configured to attach said shielding element in said irradiation unit. 4. The method for obtaining a shielding element according to claim 3, wherein the support comprises an edge located on said longitudinal external surface formed during said trimming step. 5. The method for obtaining a shielding element according to claim 3, wherein the support comprises a base made in a low density material and having an aperture greater than or equal to said aperture of a shape similar to said contour and crossing said longitudinal thickness of said block, so as to let through said beam. 6. A shielding element obtained according to the method of claim 1. 7. An irradiation unit configured to scan a hadron pencil beam from a source over a target area, said irradiation unit comprising the shielding element according to claim 6, positioned between said source and said target area, so as to minimize the penumbra of the hadron pencil beam outside said target area. 8. A method for scanning a target area with a hadron pencil beam from a source, the method comprising placing the shielding element obtained according to claim 6 in an irradiation unit between said source and said target area, so as to minimize the penumbra of said hadron pencil beam outside said target area. 9. The shielding element obtained according to the method of claim 1, wherein the hadron pencil beam has a width (σ) of from about 3 to about 12 mm when exiting the irradiation unit. 10. The shielding element obtained according to the method of claim 1, wherein the longitudinal external surface is similar to said longitudinal internal surface. 11. A method of scanning a target area with a scanning hadron pencil beam, the method comprising:positioning a shielding element between a scanning hadron pencil beam irradiation unit and the tumor target area, the shielding element having a longitudinal thickness configured to block the passage of said scanning hadron pencil beam, the shielding element comprising:an aperture of a shape similar to an open or closed contour of said target area, the aperture extending the length of the longitudinal thickness of the shielding element in order to let through said scanning hadron pencil beam, said aperture forming a longitudinal internal surface of the shielding element,a longitudinal external surface around said longitudinal internal surface, said longitudinal internal and longitudinal external surfaces delimiting a side wall of the shielding element,wherein the side wall has a side thickness everywhere greater than at least once the width (σ) of the scanning hadron pencil beam, and, in at least one location, less than five times the width (σ) of the scanning hadron pencil beam. 12. The method of scanning a target area according to claim 11, wherein the scanning hadron pencil beam has a width (σ) of from about 3 to about 12 mm when exiting the scanning hadron pencil beam irradiation unit. 13. The method of scanning a target area with a scanning hadron pencil beam according to claim 11, wherein the longitudinal external surface is similar to said longitudinal internal surface. 14. A shielding element configured for use with a scanning hadron pencil beam irradiation unit whereby a scanning hadron pencil beam is scanned over a target area, said scanning hadron pencil beam having a width (σ), the shielding element having a longitudinal thickness configured to block the passage of said scanning hadron pencil beam, the shielding element comprising:an aperture of a shape similar to an open or closed contour of said target area, the aperture extending the length of the longitudinal thickness of the shielding element in order to let through said scanning hadron pencil beam, said aperture forming a longitudinal internal surface of the shielding element,a longitudinal external surface around said longitudinal internal surface, said longitudinal internal and longitudinal external surfaces delimiting a side wall of the shielding element,wherein the side wall has a side thickness everywhere greater than at least once the width (σ) of the scanning hadron pencil beam, and, in at least one location, less than five times the width (σ) of the scanning hadron pencil beam. 15. The shielding element according to claim 14, wherein the scanning hadron pencil beam has a width (σ) of from about 3 to about 12 mm when exiting a scanning beam irradiation unit. 16. The shielding element according to claim 14, wherein the longitudinal external surface is similar to said longitudinal internal surface.
051241161	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to the drawings, it is seen in FIG. 1-3 that the invention is generally indicated by the numeral 10. Grid key 10 is formed in the shape of a rectangular bar to provide a main body portion 12 having first and second ends 14, 16. Main body portion 12 is bent near first end 14 to form a 45 degree angle in main body portion 12. In the preferred embodiment the bend in main body portion 12 is approximately 0.19 inch from first end 14 and has a bend radius of approximately 0.09 inch. First end 14 is provided with two rectangular extensions or tabs 18 that extend outwardly in opposite directions from first end 14 at a right angle to main body portion 12 to define a T-shape at first end 14. In the preferred embodiment tabs 18 are integral with main body portion 12 for ease of fabrication and strength. The following terms will be used for the sake of clarity in referring to the various dimensions of tabs 18 and main body portion 12. The dimension of tabs 18 from first end 14 back toward the bend in main body portion 12 shall be indicated as the length of tabs 18. The dimension across both tabs 18 shall be indicated as the width across tabs 18 with the corresponding dimension across main body portion 12 being indicated as the width of main body portion 12. The thickness of tabs 18 and main body portion 12 is that dimension perpendicular to the width. During loading of fuel rods into the exterior grid cells of a fuel assembly grid key 10 is used in the following manner. As seen in FIG. 4, first end 14 is inserted into window 20 of exterior grid strip 22. In this first position tabs 18 are parallel to the longitudinal axis of grid assembly 24, exterior grid strip 22, and interior grid strip 26. Each grid assembly 24 used in a nuclear fuel assembly to position fuel rods radially relative to each other is formed from a plurality of grid strips interleaved together to form cells between the strips that receive the fuel rods. Main body portion 12 is positioned such that the outside radius of the bend in main body portion 12, designated by the numeral 28, faces away from exterior grid strip 22 and the inside radius of the bend, designated by the numeral 30, faces exterior grid strip 24. Main body portion 12 is then rotated 90 degrees to a position where main body portion 12 is parallel with the longitudinal axis of grid assembly 24, tabs 18 are perpendicular to the longitudinal axis of grid assembly 24, and first end 14 is resting on the edge of interior grid strip 26. In this position tabs 18 extend beyond the edge of window 20. Main body portion 12 is then pulled away from exterior grid strip 22 and moved approximately 90 degrees in a vertical rotating motion. As seen in FIG. 5 and 6, this causes outside radius 28 to bear against interior grid strip 26 and tabs 18 to pull exterior grid strip 22 outward as main body portion 12 is moved. As main body portion 12 is moved, this causes rotation of tabs 18 from their first position where their length is between exterior grid strip 22 and interior grid strip 26 to a second position where the thickness of tabs 18 is between exterior grid strip 22 and interior grid strip 26. It is the thickness of tabs 18, which is greater than their length, that provides the camming effect in the keying or lifting of exterior grid strip 22. Dimple 32 on the interior side of exterior grid strip 22 is thus moved away the cell and a fuel rod may then be inserted into the cell without dimple 32 scratching the fuel rod. After a fuel rod is inserted grid key 10 is removed by reversing the steps used to allow insertion of the fuel rod. For ease of illustration only a portion of a grid assembly is shown. Grid key 10 may be formed from any suitable material such as 17-4 ph stainless steel or nylon or other plastics that are strong enough for such use and are easily produced by a molding process. The length and thickness of tabs 18 are adjusted according to the type of grid assembly and the space between interior grid strip 26 and exterior grid strip 22. As an example, in a zircalloy grid tested the dimension of tabs 18 from first end 14 back toward the bend (the length of tabs 18) is preferably 0.035 inch whereas in an inconel grid the preferred dimension is 0.068 inch. The width of main body portion 12 for such grids is 0.100 inch and 0.180 inch respectively while the thickness of main body portion 12 and tabs 18 is 0.059 and 0.078 inch respectively. The width across both tabs 18 for either type of grid is preferably 0.265 inch. The preferred length of main body portion 12 from the bend to second end 16 is 0.75 inch. Because many varying and differing embodiments may be made within the scope of the inventive concept herein taught and because many modifications may be made in the embodiment herein detailed in accordance with the descriptive requirement of the law, it is to be understood that the details herein are to be interpreted as illustrative and not in a limiting sense.
	summary
056087772	abstract	A slit radiography apparatus is provided with an absorption device comprising electrically controllable piezoelectric tongues. Upper and/or lower deflections of the tongues are limited to a maximum by trips outside the X-ray fan beam. More than one strip can be present as seen in the length direction from the fixed ends of the tongues. Thus deflections of the tongues are limited and great phase changes in the control signals without resultant oscillations can take place during the time the tongue are restricted in their movement by the strip.
	claims	1. A lens structure comprising: a substrate having a surface of predetermined curvature; and a film formed along a surface of the said substrate with multiple individual members each having at least one similar orientation relative to the portion of the substrate surface adjacent the member such that collectively the members provide predictable angles for diffraction of x-rays generated from a common source. 2. The lens structure of claim 1 wherein the film members each have a crystal orientation relative to an associated plane and the majority of the planes are each oriented with respect to a portion of said substrate surface adjacent the corresponding member at substantially the same angle. claim 1 3. The lens structure of claim 2 wherein the film is a polycrystalline structure comprising a plurality of grains having a fiber texture normal to the curvature of the substrate surface. claim 2 4. The lens of claim 1 wherein the film includes grains predominantly comprising Al with sufficient grains having a [ 111 ] direction normal to adjacent portions of the substrate surface such that the spatial distribution of the grains provides a fiber texture. claim 1 5. A reflective lens for converging x-rays comprising at least one curved surface of polycrystalline material. 6. The lens of claim 5 wherein the lens includes a reflective surface region of curvature for converging said x-rays into a beam of substantially parallel rays. claim 5 7. The lens of claim 5 wherein the lens includes a reflective surface region of curvature for converging said x-rays about a focal region. claim 5 8. A lens structure comprising: a polycrystalline film formed along a surface and having a curved plane fiber texture orientation-suitable to Provide parallel x-rays or suitable for focusing x-rays. 9. The structure of claim 8 wherein the film comprises lattice structures suitable for Bragg reflection along a sufficient portion of the surface to focus x-rays. claim 8 10. A method for transmitting x-rays to a region comprising: reflecting x-rays from a curved polycrystalline surface having a curved plane texture orientation based on Bragg diffraction. 11. The method of claim 10 wherein the step of reflecting the x-rays is performed about a surface curvature which converges the x-rays. claim 10 12. The method of claim 10 wherein the step of reflecting the x-rays is performed about a surface curvature which converges the x-rays into a beam of substantially parallel rays. claim 10 13. The method of claim 10 wherein the step of reflecting the x-rays is performed about a surface curvature which focuses the x-rays about a point. claim 10 14. A method for forming a Bragg reflecting surface comprising: providing a substrate having a surface of predetermined curvature; and forming a polycrystalline layer over the surface with the majority of individual crystalline grains having a common orientation with respect to the underlying substrate surface to provide a curved plane texture orientation of the type suitable for transforming divergent x-rays into Parallel or focusing radiation. 15. A device for translating x-rays, comprising: a polycrystalline surface region having crystal spacings suitable for reflecting a plurality of x-rays at the same Bragg angle along the region and transmitting the reflected x-rays to a reference position.
	claims	1. A nuclear light water moderated boiling water reactor comprising: a plurality of fuel assemblies arranged vertically in the reactor, the reactor being arranged such that water flows upwards through said fuel assemblies whereby part of the water is transformed into steam; at least one of the fuel assemblies comprising at least five fuel units stacked on top of each other, each fuel unit comprising a top tie plate and a bottom tie plate and a plurality of fuel rods extending between the top tie plate and the bottom tie plate; at least one of said fuel units having fuel rods of different diameters from fuel rods of another of said fuel units; and said at least one fuel assembly having a fuel channel with a substantially square cross section surrounding the stack of said at least five fuel units. 2. A nuclear light water moderated boiling water reactor according to claim 1 , wherein at least two of the fuel units in said at least one fuel assembly differ from each other in at least one of the following respects: claim 1 the fuel units have different numbers of fuel rods; the fuel units have different lattice configurations; the fuel units have different heights; the fuel units have bottom tie plates with different embodiments with respect to the mixing of coolant; and the fuel rods have different degrees of enrichment. 3. A nuclear light water moderated boiling water reactor according to claim 1 , wherein said fuel units are retained by a spring arrangement arranged in an upper part of the fuel assembly. claim 1 4. A nuclear light water moderated boiling water reactor according to claim 1 , wherein at least one fuel rod in said at least one fuel assembly comprises a top plug for attachment to the top tie plate of the fuel unit in which said at least one fuel rod is positioned and a bottom plug for attachment to the bottom tie plate of said fuel unit in which said at least one fuel rod is positioned, wherein, for receiving fissile gases, at least one of said top and bottom plugs is hollow and has an opening facing the interior of said fuel rod. claim 1 5. A nuclear light water moderated boiling water reactor according to claim 1 , wherein at least one fuel rod in said at least one fuel assembly has a coating protecting against abrasion damage to that part of the cladding tube of the fuel rod which is nearest the top tie plate of the fuel unit in which the fuel rod is positioned. claim 1 6. A nuclear light water moderated boiling water reactor according to claim 1 , wherein each of said fuel units comprises at least 80 fuel rod positions. claim 1 7. A nuclear light water moderated boiling water reactor according to claim 1 wherein the number of fuel units in said at least one fuel assembly is larger than six. claim 1 8. A method comprising: providing a fuel assembly in a nuclear light water moderated boiling water reactor arranged such that water flows upwards through said fuel assembly whereby part of the water is transformed into steam, said fuel assembly being arranged vertically in said reactor; providing at least five fuel units stacked on top of each other in said fuel assembly, each fuel unit comprising a top tie plate and a bottom tie plate and a plurality of fuel rods extending between the top tie plate and the bottom tie plate; providing, in at least one of said fuel units, fuel rods of different diameters from fuel rods of another of said fuel units; wherein said fuel assembly has a fuel channel with a substantially square cross section surrounding the stack of said at least five fuel units. 9. The method according to claim 8 , wherein at least two of the fuel units in said at least one fuel assembly differ from each other in at least one of the following respects: claim 8 the fuel units have different numbers of fuel rods; the fuel units have different lattice configurations; the fuel units have different heights; the fuel units have bottom tie plates with different embodiments with respect to the mixing of coolant; and the fuel rods have different degrees of enrichment. 10. The method according to claim 8 , wherein said fuel units are retained by a spring arrangement arranged in an upper part of the fuel assembly. claim 8 11. The method according to claim 8 , wherein at least one fuel rod in said fuel assembly comprises a top plug for attachment to the top tie plate of the fuel unit in which said at least one fuel rod is positioned and a bottom plug for attachment to the bottom tie plate of said fuel unit in which said at least one fuel rod is positioned, wherein, for receiving fissile gases, at least one of said top and bottom plugs is hollow and has an opening facing the interior of said fuel rod. claim 8 12. The method according to claim 8 , wherein at least one fuel rod in said fuel assembly has a coating protecting against abrasion damage to that part of the cladding tube of the fuel rod which is nearest the top tie plate of the fuel unit in which the fuel rod is positioned. claim 8 13. The method according to claim 8 , wherein each of said fuel units comprises at least 80 fuel rod positions. claim 8 14. The method according to claim 13 , wherein the number of fuel units in said fuel assembly is larger than six. claim 13
054597684	claims	1. In a nuclear reactor having an interior, a pressure vessel, a coolant conducting surface exposed to primary pressure, and a core, a safety device against overpressure failure of the pressure vessel upon insufficient cooling of the core, comprising: a differential-pressure-loaded pressure relief valve being set in the coolant conducting surface, said pressure relief valve having a hollow guide cylinder, a closure piece in the form of a differential-pressure piston being constructed as a hollow body and being longitudinally displaceable in said hollow guide cylinder between a closure position and an opening position, and a fusible stop sealing and retaining said differential-pressure piston in said closure position, said fusible stop melting due to a threshold temperature heat flow reaching said fusible stop upon reaching an upper threshold temperature in the interior of the reactor, for permitting said differential-pressure piston to move into said opening position. said pressure relief valve has a valve body with a wall and an inner periphery and said guide cylinder has an outer periphery, defining an annular duct remaining free as an overflow duct between said inner and outer peripheries; said pressure relief valve has vanes being disposed in said annular duct and joined to said wall of said valve body for holding said guide cylinder in a centered position in said valve body; and said overflow duct has an inlet cross section being normally sealed by said differential-pressure piston in said closure position and being cleared and released in said opened position. 2. The safety device according to claim 1, wherein the coolant conducting surface is a wall of the pressure vessel. 3. The safety device according to claim 1, wherein the coolant conducting surface is a pipeline connected to the pressure vessel. 4. The safety device according to claim 1, including a pressurizer discharge line being connected to said pressure relief valve and opening into a pressurizer relief tank. 5. The safety device according to claim 1, wherein said pressure relief valve has seating surfaces, said differential-pressure piston has sealing surfaces, and said fusible stop is disposed between said sealing surfaces and said seating surfaces. 6. The safety device according to claim 5, wherein said differential-pressure piston has peripheral piston surfaces, said guide cylinder has an inner periphery with guide surfaces, and said fusible stop is additionally disposed between said peripheral piston surfaces and said guide surfaces. 7. The safety device according to claim 1, wherein: 8. The safety device according to claim 1, wherein said guide cylinder has an end facing away from said differential-pressure piston, and said end has a bottom with a pressure relief orifice formed therein. 9. The safety device according to claim 1, wherein the coolant conducting surface is a primary coolant pipe having a wall, and said pressure relief valve is set into the wall near the pressure vessel. 10. The safety device according to claim 1, wherein the pressure vessel has primary coolant pipe sockets at a given level and a wall forming the coolant conducting surface, and said pressure-relief valve is set into the wall at the given level between the primary coolant pipe sockets. 11. The safety device according to claim 1, including a separate relief valve, and a pressure relief line being connected to said pressure relief valve and being constructed as a control line for said separate relief valve, said pressure relief valve and said control line having correspondingly low cross sectional dimensions.
	description	This application claims the benefit of provisional patent application No. 60/673,158, filed on Apr. 20, 2005. Quantum dots are semiconductor nanocrystals ranging from nanometers in size to a few microns. The size controls the number of electrons contained in the dot. Each quantum dot can contain from one to several thousand electrons. Since quantum dots are so small, quantum mechanical effects force the electron energy levels to be quantized. Quantum dots have sometimes been called artificial atoms because the electron quantum levels contained within a dot are similar to the electron orbitals in an atom. This quantization allows distinct wavelength (colors) of light to be emitted. Light or electric current typically excites them. The light emitted ranges from ultraviolet to visible to infrared, depending on the material and the size. This is a wavelength span of 350 to 2300 nanometers. The emission has a very narrow bandwidth of 30 nanometers, full width at half maximum (FWHM). Work is being conducted on using quantum dots as a replacement for LEDs. In addition to semiconductors, quantum dots can be made from metal. Presently, quantum dots are made by vacuum techniques such as molecular beam epitaxy (MBE) or chemical vapor deposition (CVD), or in aqueous solutions where a colloid is formed. Other techniques may be developed. Cadmium selenide (CdSe) is a common material for visible light quantum dots. A 3 nanometer CdSe dot emits 520-nanometer light that is green. Increasing the diameter to 5.5 nanometers increases the wavelength to 630 nanometers, the wavelength of red light. Quantum dots can be “tuned” by controlling their size to have any desired pure color of a desired wavelength. Other phosphors suitable for quantum dots include doped zinc sulfide (ZnS) compounds. Gold quantum dots have also been made. CdSe quantum dots may be coated with ZnS as a protective layer. Quantum dots are commercially available from companies such as Evident Technologies of Troy, N.Y. Quantum dots have a myriad of applications, including medical applications for tagging proteins and antibodies. The quantum dots fluoresce to map the proteins and antibodies. Other uses of quantum dots include photovoltaic solar cells, electroluminescent devices, the phosphorous of LED lights, thermoelectrics, inks, pigments and anti-counterfeiting materials, to mention just a few areas of current research and development. Quantum dots have been proven to be radiation resistant. It would therefore be advantageous if energy, namely alpha or beta particles from nuclear sources, could be utilized as the energy source to energize quantum dots for use in light sources having precise wavelengths and intensities, which precise light sources could be used, for example, to calibrate light sources. Other uses of these quantum dots include their uses in calibrating detector equipment such as ATP luminometers used for measuring the presence of ATP in swab samples, etc. The company Isotope Products Laboratories, of Valencia, Calif., currently supplies light sources using alpha and beta emitters as light calibration sources. Presently, scintillator manufacturers have a limited choice of wavelengths available. Quantum dots excited by alpha particles or beta particles would provide a light source with precise wavelengths ranging from the infrared to ultraviolet light. Besides permitting light sources to be made having a predetermined, single wavelength, such light sources could also, for example, allow multiple wavelengths of light to be emitted by using combinations of quantum dots. This is not currently available. Such light sources could be used as calibration sources for instruments using photomultiplier tubes or PIN photodiodes. Scintillator light sources are used in bacteria contamination detection systems, such as in ATP luminometers. Brighter light sources can be used to illuminate gun sights replacing fragile glass tubes containing tritium. The use of scintillator material limits the available plastic matrix to polyvinyltoluene (PVT). PVT is a delicate plastic whose surface is damaged by finger prints. However, with quantum dots, other plastics such as epoxy, various resins, and silicone can be used. Quantum dots can even be embedded in glasses. Light sources can be made using techniques such as spray coating or screen-printing. Spin or dip coating can also form films. Ink jet systems can also be used for film deposition. Quantum dots having desired qualities, (e.g., selected wavelength of light produced and intensity) can be combined in close association or proximity with one or more types of radionuclides in a matrix of optically translucent or transparent material. More particularly, one or more types of quantum dots that produce light of one or more desired wavelengths, respectively, are mixed into a translucent or transparent matrix, e.g., cured UV resin, air cured resins, along with one or more types of radionuclides. The radionuclides emit either alpha or beta particles. Examples of beta radiation emitting radionuclides include, but are not limited to, hydrogen 3 (tritium, or 3H), carbon 14 (14C), silicon 32 (32Si), nickel 63 (63Ni), and thallium 204 (204Tl). Examples of alpha radiation emitting radionuclides include, but are not limited to, polonium 210 (210Po), americium 241 (241Am) and thorium 232 (232Th). The radionuclide(s) will provide energy, in the form of emitted alpha or beta particles, which will energize the quantum dots and cause them to emit light at the desired wavelength(s) and intensity. Turning first to FIG. 1, there is shown a side view of an exemplary nuclear powered quantum dot light source 10. The exemplary nuclear powered quantum dot light source 10 preferably has a container 12 in which is located a quantum dot and radionuclide containing matrix 14. The container 12 can be formed in a desired size and shape, e.g., disk shaped, box shaped, etc., and of a desired material, such as metal, plastic, glass, etc. The container 12 has at least one side 16 which is open or light transmissive, with other sides 18 being opaque if desired. The sides 18 could be made to be reflective (e.g. mirrored) so that the light generated by the energized quantum dots is more efficiently reflected out of the container. A layer of material 20 not containing radioisotopes or quantum dots can optionally be provided over the at least one side 16. The layer of material 20 is made of clear or light transmissive material, such as a UV curable resin and can be provided to help ensure that physical human contact with any radioisotopes is eliminated. In addition or alternately, the entire container 12 can be encapsulated within a nonradioactive envelope. FIG. 2 is a diagrammatic and simplified representation of the exemplary matrix 14 containing quantum dots 30 and radionuclides 32 contained in a matrix material 34, such as a UV cured resin. Some other materials can include thermal sol-gel hybrids, UV sol-gel hybrids and plastic resins such as polycarbonate, polystyrene, PMMA (polymethylmethacrylate), and polyethylene. The ratio of the quantum dots 30 to radionuclide(s) 32 in the matrix material 34 can be determined as is required. The quantum dots are sized and engineered to produce light with a wavelength span of 350 to 2300 nanometers (ultraviolet to visible to infrared), with the size and material of the quantum dots determining the wavelength emitted. One or more different types of quantum dots can be used to provide either light at a single wavelength, or if desired, multiple wavelengths. The radionuclides 32 can be selected from radionuclides that emit either alpha or beta particles. Examples of beta radiation emitting radionuclides include, but are not limited to hydrogen 3 (tritium, or 3H), carbon 14 (14C), silicon 32 (32Si), nickel 63 (63Ni), and thallium 204 (204Tl). Examples of alpha radiation emitting radionuclides include, but are not limited to, polonium 210 (210Po), americium 241 (241Am) and thorium 232 (232Th). The radionuclide(s) will provide energy, in the form of emitted alpha or beta particles, that will energize the quantum dots and cause them to emit light at the desired wavelength(s) and intensity. FIG. 2 is provided as a simplified representation, but in actual construction, the amount of radionuclides relative to the number of quantum dots, and the concentrations of the radionuclides and radionuclides in the matrix will be selected so that the quantum dots are adequately energized and emit light of the proper wavelength and intensity. FIG. 3 is a side view of another exemplary nuclear powered quantum dot light source 50. In this embodiment, the nuclear powered quantum dot light source 50 is adapted for use in devices such as ATP luminometers, and comprises a test tube shaped holder 52 which contains a quantum dot containing matrix 54. The test tube shaped holder 52 has a sealed bottom 56 and an open top 58. A cap 60 is made of a radiolucent transparent material, such as most plastics and some glasses, and is used to seal off the holder 52 with the quantum dot containing matrix 54 contained therein, and prevents direct human contact with the quantum dot containing matrix 54. FIG. 4 is a side view of yet another exemplary nuclear powered quantum dot light source 70 that is similar to holder 50 of FIG. 3. This embodiment of nuclear powered quantum dot light source 70 comprises a test tube shaped holder 72 which contains a quantum dot containing matrix 74. The test tube shaped holder 72 has a sealed bottom 76 and a top 78 that is sealed off with a cap 80. Depending on the requirements, the cap 80 can be made of a radiolucent material and will seal off the holder 52 with the quantum dot containing matrix 54 therein, and thereby prevent direct human contact with the quantum dot containing matrix 54. With respect to all the holders discussed above, if desired, it is possible for the holders to have walls formed entirely of radiolucent material. Lastly, although all the embodiments are shown having a holder, it is possible to form the light source without a holder, such as by extruded or cast the material without a container. Such as use might be appropriate, for example, where the light source is placed in another device. Having thus described the exemplary embodiments of the present invention, it should be understood by those skilled in the art that the above disclosures are exemplary only and that various other alternatives, adaptations, and modifications may be made within the scope of the present invention. The presently disclosed embodiment is to be considered in all respects as illustrative and not restrictive. The scope of the invention being indicated by the appended claims rather than the foregoing description, and all changes which come within the meaning and range of equivalency of the claims are, therefore, intended to be embraced therein.
	summary
	description	Historically, engineered robust packaging systems for radioactive waste that incorporate radiation shielding have been designed, licensed and deployed on a project or application specific basis. This means that such a packaging system has to be designed for each individual project or application. The unique requirements for each project or application dictates the design of the packaging system. This makes the packaging system unsuitable for other projects and applications having different requirements. For example, a project may require a container made of double wall stainless steel integral welded shells, monolithic cast-in-place high density concrete shielding with steel reinforcing, and extensive machining of mating surfaces with bolted and welded lid. The stainless steel shells, shielding, lid, dimensions, and so forth are all specific to the project. The container cannot be used for another project that has different waste contents specification, confinement requirements, needs more or less shielding, or has different closure requirements. The use of a custom packaging system for each project or application causes other problems. The cost to design, demonstrate regulatory compliance, and fabricate a packaging system for each project is substantial and cost overruns are common. Fabricating a custom designed system is complex and there are often numerous fabrication nonconformances. The difficulty of fabricating the system often results in schedule overruns and delays. A number of representative embodiments are provided to illustrate the various features, characteristics, and advantages of the disclosed subject matter. The embodiments are provided in a variety of specific contexts although it should be understood that many of the concepts can be used in a variety of other settings, situations, and configurations. For example, the features, characteristics, advantages, etc., of one embodiment can be used alone or in various combinations and sub-combinations with the features, characteristics, advantages, etc., of one or more other embodiments. A modular packaging system for radioactive waste is structurally and mechanically robust, highly functional and configurable, and can be used for nearly all radioactive waste streams that require shielded packaging. It provides cradle-to-grave functionality for loading, interim storage, transport, and disposal of radioactive waste. It provides a platform that can be tailored in the field for batch-specific radioactive waste streams and includes uniform equipment interfaces that provide maximum operational flexibility to end users. The packaging system eliminates the conventional practice of developing custom packages for nearly every project and/or radioactive waste stream. It includes a standard modular container that can be configured using a catalog of features to package most types of radioactive waste. The basic process for configuring the container is as follows: (1) evaluate the specifications of the radioactive waste, (2) select a modular container grade and features, e.g., confinement boundary robustness, and the like, (3) select shielding material and thickness that corresponds to the specifications of the radioactive waste, and (4) select the features for the cavity of the modular container, e.g., liner, support framework for sub-containers, and the like. The modular container can be used to package radioactive waste from source to disposal including remote waste processing, remote container loading and handling, interim storage, off-site storage, and/or disposal by shallow land burial or in a geological repository. The modular container is capable of holding solid, granular, and wet radioactive waste. The modular container makes it unnecessary to handle and package the waste multiple times before final disposition. This lowers the lifecycle cost associated with managing radioactive waste. The modular container can be reused or disposed with the radioactive waste. It can also be configured by the end user to suit batch-specific waste streams. The modular container includes a standard enclosure envelope that can be configured in a variety of ways to meet the requirements of a specific project or application. Also, the modular container can include other components such as a liner to hold granular radioactive waste, a support framework to hold sub-containers of wet radioactive waste, and other support frameworks such as baskets, dividers, and the like to hold various types of solid radioactive waste, spent nuclear fuel (SNF), and high level waste (HLW). The modular container can include modular shielding inserts or members that can be used to adjust the shielding of the modular container to satisfy the requirements of a given project or application. The modular shielding inserts can be made of a variety of suitable materials and have any of a number of suitable thicknesses. The entire contents of all sections of the U.S. Code of Federal Regulations (CFR) and the International Atomic Energy Agency regulations referenced in this document are incorporated by reference. In the event of a conflict, the subject matter explicitly recited or shown in this document controls over any subject matter incorporated by reference. The incorporated subject matter should not be used to limit or narrow the scope of the explicitly recited or depicted subject matter. The Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. The Summary and the Background are not intended to identify key concepts or essential aspects of the disclosed subject matter, nor should they be used to constrict or limit the scope of the claims. For example, the scope of the claims should not be limited based on whether the recited subject matter includes any or all aspects noted in the Summary and/or addresses any of the issues noted in the Background. A packaging system for radioactive waste is modular in nature and can be tailored for a variety of radioactive waste. The packaging system is modular in that it can be deconstructed into a number of component parts or subsystems that can be mixed and matched in a variety of configurations. The components are able to connect, interact, fit together, and otherwise interoperate by adhering to an overall standardized design. The packaging system includes the following standardized subsystems and/or components: containers (including enclosure envelopes), interior shielding inserts, exterior shielding panels, interior loading baskets, impact limiters, interior liners, interior support frameworks, transport overpack system (including transport containers and transport impact limiters), disposal overpacks, and disposal vaults. Each subsystem or component can be configured separately and then used in conjunction with any other subsystem or component to provide a tremendous amount of flexibility to package a variety of radioactive waste. It should be appreciated that the standardized subsystems and components listed in the previous paragraph are provided by way of example and do not represent an exhaustive list of all the standardized subsystems and components of the packaging system. The packaging system can include additional standardized subsystems and components beyond those listed. Each standardized subsystems and components can be referred to as being modular because they are what make the packaging system modular. The packaging system includes standardized equipment handling interfaces such as standard forklift, crane rigging, and the like. It can be used for interim storage of radioactive waste as well as transport and final disposition by shallow surface burial and geological repository burial. The packaging system can handle any class of radioactive waste from Class A low level waste to high level waste. Radioactive waste can be classified according to a number of systems in use worldwide. It should be appreciated that some classifications use similar terminology but define the specifics of the waste differently. Despite this, radioactive waste can generally be divided into the following classifications. Low level waste (LLW) is generally radioactive waste that is suitable for near surface or shallow land disposal. This is a disposal option suitable for waste that contains such an amount of radioactive material that robust containment and isolation for limited periods of time up to a few hundred years are required. LLW covers a wide range of radioactive waste. It ranges from radioactive waste with an activity level that does not requiring shielding or particularly robust containment and isolation, to radioactive waste with an activity level such that shielding and more robust containment and isolation are necessary for periods up to several hundred years. Because LLW may have a wide range of activity concentrations and may contain a wide range of radionuclides, there are various design options for near surface disposal facilities. These design options may range from simple to more complex engineered facilities, and may involve disposal at varying depths, typically from the surface down to 30 m. They will depend on safety assessments and on national practices, and are subject to approval by the governing regulatory body. LLW can include low concentrations of long lived radionuclides. Although the waste may contain high concentrations of short lived radionuclides, significant radioactive decay of these will occur during the period of reliable containment and isolation provided by the site, the engineered barriers, and institutional control. The IAEA regulations defining LLW are set forth in IAEA CSG-1. In the U.S., LLW is radioactive waste that is defined by what it is not. It is radioactive waste not classified as high-level, spent fuel, transuranic or byproduct material such as uranium mill tailings. LLW has four subcategories: Classes A, B, C, and Greater Than Class C (GTCC), described below. On average, Class A is the least hazardous while GTCC is the most hazardous. The U.S. regulations defining Class B, C and GTCC are set forth in 10 CFR 61.55. Class A radioactive waste is the least radioactive of the four LLW classes. It is primarily contaminated with short-lived radionuclides. For example, it can have an average concentration of 0.1 Ci/ft3, Class B radioactive waste is contaminated with a greater amount of short-lived radionuclides than Class A. For example, it can have an average concentration of 2 Ci/ft3). Class C radioactive waste is contaminated with greater amounts of long-lived and short-lived radionuclides than Class A or B. For example, it can have an average concentration of 7 Ci/ft3. GTCC radioactive waste is the most radioactive of the low-level classes. It can have an average concentration of 300 to 2,500 Ci/ft3. TABLE 1Low Level Waste Classification TableClass AClass BClass CRadionuclide(Ci/m3)(Ci/m3)(Ci/m3)Total of all nuclides with less700No limitNo limitthan 5 years half lifeH-3 (Tritium)40No limitNo limitCo-60700No limitNo limitNi-633.570700Ni-63 in activated metal357007000Sr-900.041507000Cs-1371444600C-140.88C-14 in activated metal880Ni-59 in activated metal22220Nb-94 in activated metal0.020.2Tc-990.33I-1290.0080.08Alpha emitting transuranic nuclides 10 nCi/g 100 nCi/gwith half life greater than 5 yearsPu-241 350 nCi/g 3500 nCi/gCm-2422000 nCi/g20000 nCi/g Intermediate level waste (ILW) is radioactive waste that contains long lived radionuclides in quantities that need a greater degree of containment and isolation from the biosphere than is provided by near surface disposal. Disposal in a facility at a depth of between a few tens and a few hundreds of meters is indicated for ILW. Disposal at such depths has the potential to provide a long period of isolation from the accessible environment if both the natural barriers and the engineered barriers of the disposal system are selected properly. In particular, there is generally no detrimental effect of erosion at such depths in the short to medium term. Another important advantage of disposal at intermediate depths is that, in comparison to shallow surface disposal facilities suitable for LLW, the likelihood of inadvertent human intrusion is greatly reduced. Consequently, long term safety for disposal facilities at such intermediate depths will not depend on the application of institutional controls. Notably, ILW is a classification that is not used in the U.S. The IAEA regulations defining ILW are set forth in IAEA CSG-1. High level waste (HLW) is produced by nuclear reactors and include SNF and/or reprocessing waste. HLW contains such large concentrations of both short and long lived radionuclides that a greater degree of containment and isolation from the accessible environment is needed to ensure long term safety. Containment and isolation is usually provided by the integrity and stability of deep geological disposal, with engineered barriers. HLW generates significant quantities of heat from radioactive decay, and normally continues to generate heat for several centuries. Heat dissipation is an important factor that has to be taken into account in the design of geological disposal facilities. HLW typically has levels of activity concentration in the range of 104-106 TBq/m3 (e.g. for SNF recently discharged from power reactors). HLW includes conditioned waste arising from the reprocessing of SNF together with any other waste requiring a comparable degree of containment and isolation. At the time of disposal, following a few decades of cooling time, waste containing such mixed fission products typically has levels of activity concentration of around 104 TBq/m3. In the U.S., the regulations that define HLW are set forth in 10 CFR 60/63 In the U.S., transuranic waste (TRU) is radioactive waste that contains elements with atomic numbers (number of protons) greater than 92, the atomic number of uranium. The meaning of the term transuranic is above uranium. TRU includes only waste material that contains transuranic elements with half-lives greater than 20 years and concentrations greater than 100 nanocuries per gram. If the concentrations of the half-lives are below the limits, it is possible for waste to have transuranic elements but not be classified as TRU waste. The regulations defining transuranic waste are set forth in 10 CFR 61.55. There are also other classes of radioactive waste including special form material (10 CFR 71.75; 49 CFR 173.476; IAEA TS-G-1.1), special nuclear material (10 CFR 70.4), source material (10 CFR 40.4), and by product material (10 CFR 30.4). Standardized Container The packaging system includes a modular container 10, one example of which is shown in FIGS. 1-4. The modular container 10 includes a structural lid 18 (alternatively referred to as a top closure) positioned on a main body 30 to enclose a cavity 14 for the radioactive waste. The main body 30 includes side walls 16, side and corner wall support members 24 (alternatively referred to as side and corner tubes or support tubes), and a base 20 (alternatively referred to as a support base). The base 20 includes a base plate 26 (alternatively referred to as a base member) and base support members 22. The support members 22, 24 provide additional robustness, strength, and rigidity to the modular container 10. The walls 16, the structural lid 18, and the base plate 26 form the interior boundary of the cavity 14 and serve to define the enclosure envelope 12 (alternatively referred to as a main enclosure or confinement boundary). The modular container 10 can have any of a number of different configurations all of which are compatible with the other subsystems and/or components of the packaging system. Three specific configurations are described in greater detail and referred to as Grade A, B, and C modular container variants (the embodiment shown in FIGS. 1-4 corresponds to Grade B). The grades roughly correspond to the activity of the radioactive waste with Grade A being the most robust variant configured for use with the most active waste and Grade C being the least robust variant configured for use with the least active waste. It should be appreciated that the modular container 10 can have any number of grades or configurations. The different configurations of the modular container 10 are easy to assemble and can be inexpensively mass produced in large quantities compared to conventional containers. In one embodiment, the parts of the modular container 10 are self-jigging which simplifies fit-up and assembly. Something is generally considered self-jigging when its component parts incorporate design features that ensure each component, when assembled, remains in proper relationship throughout the fastening process (e.g., welding, bolting, and the like) without the aid of auxiliary fixtures. The dimensions and external features or interfaces of the modular container 10 are standardized for all grades, including the Grade A, B, and C variants. The external features are appurtenances on the exterior of the modular container 10 that facilitate remotely handling, moving, loading, lid placement, and/or stacking (as well as other operations) of the modular container 10. The features can include appurtenances such as standard lifting equipment, interfaces, and the like. In one embodiment, the modular container 10 includes openings 34 (alternatively referred to as bottom pockets) in the base 20 to receive the forks of a forklift. The openings 34 can have any suitable configuration that allows them to receive the forks. In one embodiment, the openings 34 fully capture the forks to reduce the likelihood of the modular container 10 toppling during movement. The modular container 10 can also be lifted using the openings 34 with a suitable spreader bar or sling. In another embodiment, the modular container 10 includes lifting members 28 (alternatively referred to as lifting lugs) on the structural lid 18 and/or the main body 30. The lifting members 28 can be used to remotely lift the structural lid 18 and the main body 30 together or separately and to guide stacking of the modular containers 10. The structural lid 18 can also include guide members 32 that guide placement of the structural lid 18 on the main body 30 by remote means as necessary. In the embodiment shown in FIGS. 1 and 3, there are two guide members 32 extending outward from the base of each lifting member 28 on the structural lid 18. The guide members 32 are spaced apart to allow the corresponding lifting member 28 on the main body 30 to pass between the guide members 32. The top of the lifting members 28 on the main body 30 are rounded so that the guide members 32 easily move to the side to align the structural lid 18 with the main body 30. The enclosure envelope 12 provides a robust confinement boundary for radioactive waste. The size and shape of the enclosure envelope 12 is standardized for all waste-forms and activity levels. In one embodiment, the enclosure envelope 12 is formed by coupling the structural lid 18 to the main body 30 with fasteners 42 such as bolts or the like. Once the structural lid 18 is in place, the fasteners 42 can be manually installed while the workers are frilly shielded from the radioactive waste by the main body 30 and the structural lid 18. In one embodiment, the structural lid 18 has a stepped design that forms a shear key that resists lateral and other loads and maintains the seal. For example, the main body can include a flange 36 coupled to the outside of the side walls 16 just below their upper edges. The structural lid 18 is stepped around the edges to extend over the upper edges of the side walls 16 and down to the flange 36 (using a spacer 38 in the embodiment shown in FIG. 4). The structural lid 18 is coupled to the flange 36 using one or more sealing members 44, which can be a gasket, O-ring, or the like depending on the application. The enclosure envelope 12 can be welded and leak tested. It should be appreciated that the various components of the modular container 10 and the packaging system as a whole can be fastened together in a variety of ways. Two of the most common ways include bolting and welding. It should be appreciated that any of the components of the packaging system can be coupled together using one or both of these techniques without explicitly reciting the same. The fasteners and/or fastening techniques used can be inspected (e.g., non-destructive examination of welds) and leak tested. It should also be appreciated that for purposes of this disclosure, the term “coupled” means the joining of two members directly or indirectly to one another. Such joining may be stationary in nature or movable in nature. Such joining may be achieved with the two members or the two members and any additional intermediate members being integrally formed as a single unitary body with one another or with the two members or the two members and any additional intermediate member being attached to one another. Such joining may be permanent in nature or alternatively may be removable or releasable in nature. The modular container 10 can include a filtered cavity vent 40 depending on the application. The vent 40 is typically included in situations where the pressure inside the cavity 14 has the potential of exceeding design conditions. The vent 40 prevents this from happening by allowing gas to escape. A filter is used to prevent radioactive material from escaping through the vent 40. The modular container 10 and any of its subsystems and/or components can be made of any suitable material. In general, the robustness and corrosion resistance of the material used to make the modular container 10 corresponds to the activity level of the radioactive waste. For example, the Grade C modular container can be made of lower cost materials such as structural carbon steel plate with comparatively reduced thickness and coated with decontaminable epoxy. The Grade A modular container variant 70 can be made of structural stainless steel plate with a comparatively increased thickness to provide increased structural capacity and corrosion resistance (no coating performance or maintenance issues over longer term) and to mitigate brittle fracture concerns. Standardizing the size and shape of the enclosure envelope 12 of the modular container 10 facilitates common operational interfaces and allows more economical non-structural materials to be utilized for the separate shielding inserts. Fabricating the enclosure envelope 12 to Type A transportation packaging standards avoids costly Type B transportation packaging fabrication for every container. This decouples modular container production manufacturing from more rigorous Type B transportation packaging licensing constraints. The modular container 10 can have any suitable shape so long as the other components and subsystems of the packaging system have a corresponding shape to preserve the modular nature of the system. It is preferable for the modular container 10 to have a cuboidal shape such as those shown in the Figures. The cuboidal shape of the modular container 10 with separate shielding inserts has a number of advantages relative to conventional cylindrical containers with concentric shells and integral shielding such as more efficient volume utilization, simpler loading, handling, and stacking, and the ease of fabrication and assembly sequencing. However, it should be appreciated that the modular container 10 can have other shapes such as cylindrical. The modular container 10 can be used with any type and/or form of radioactive waste that can physically fit in it. Examples of suitable types of radioactive waste include: solid waste—highly activated or surface contaminated components; granular waste—metallic fines, concrete ruble or excavated materials in drop-in liner; wet waste—stabilized liquid waste positioned in a support framework with one or more subcontainers; and other waste—smaller spent fuels, special form waste, and low to moderate pressure and heat generating wastes with application-specific inserts. The modular container 10 is especially useful for radioactive waste that exceeds Class A, but can also be used with Class A waste although such waste does not typically require such a robust engineered container. In one embodiment, the modular container 10 can be configured to be used with radioactive waste having higher concentrations of short lived isotopes such as Class B and C low level waste that has low concentrations of long-lived isotopes (see 10 CFR 61.55 and IAEA CSG-1). The modular container 10 can also be configured to be used with waste having high concentrations of short and/or long-lived isotopes such as greater than Class C waste (GTCC) (see 10 CFR 61.55), intermediate level waste (see IAEA CSG-1), transuranic waste (see 10 CFR 61.55), and high level waste (see 10 CFR 60/63). The modular container 10 can also be configured to hold special form material such as indispersible radioisotope material and sealed capsule containing radioisotope material (see 10 CFR 71.75 and 49 CFR 173.476, IAEA TS-G-1.1) It can also be configured to hold by-product material such as fuel and strategic nuclear material production waste such as source material tailings as well as byproduct waste from commercial, medical, or research activities (see 10 CFR 30.4). It should be appreciated that U.S. and IAEA regulations are typically referenced in this document with the understanding that other similar or corresponding regulations can be applicable depending on the jurisdiction. The modular container 10 can be configured to hold some types of SNF. This can be done using the Grade A container envelope and associated shielding inserts for the modular container 10 and using content-specific cavity features such as nuclear fuel specific loading baskets. In general, SNF that can be put in the modular container 10 are those that have compact geometry, lower decay heat flux, and lower pressure generation compared to LWR fuels. Examples include advanced gas-cooled reactor (AGR) oxide fuel, metallic uranium fuels such as materials testing reactor (MTR) and TRIGA research reactor fuels, natural uranium fuels such as Canada deuterium uranium (CANDU) and Magnox reactor fuels, as well as other defense and research reactor fuels that fit. The modular container 10 can be used for interim storage of radioactive waste on-site or at an off-site interim storage facility (indoor or outdoor interim storage). The Grades A, B, and C modular containers meet on-site interim storage, off-site transport as Type A or IP-2 packaging, and disposal requirements including accidental drop depending on the application. The lid closure system also satisfies Type A and IP-2 packaging requirements. It can also be used to transport the radioactive waste off-site if the waste is subject to Type A and/or IP-2 requirements at the time of loading or following decay in interim storage. If the waste is subject to the Type B requirements, then the modular container 10 can be packaged in a reusable transport overpack specifically designed for the modular container 10 that meets Type B requirements. TABLE 2Modular Container Storage, Transport, and DisposalModularContainerGradeInterim StorageOffsite TransportDisposalA10 CFR 72Type A as-is orGeologic disposal as-isoverpack as Type Bor in disposal overpackBType A plusType A as-is orGeologic disposal as-issite-specificoverpack as Type BCIP-2IP-2Surface disposal as-is Shielding Inserts The modular container 10 can include modular shielding inserts (alternatively referred to as modular shielding slabs) with variable thicknesses to customize the modular container 10 to the activity level of the radioactive waste. The modular container 10 and modular shielding inserts provide a number of advantages compared to conventional containers. The modular container 10 with modular shielding inserts is shown in FIGS. 5-9. The modular nature of both the container 10 and the shielding inserts simplifies the supply chain, shortens the delivery schedule, and allows more efficient parallel manufacturing. For example, the modular container 10 can be manufactured using higher precision nuclear-grade manufacturing processes that use, for example, fixturing to achieve low-defect production and repeatable mass production of consistently high quality product. This mitigates the high cost and delays due to non-conforming product. The shielding inserts can be manufactured in parallel with the modular container 10 using lower precision manufacturing processes. The shielding inserts can be delivered for assembly in near final form. The shielding inserts can be placed in the modular container 10 near where the modular container 10 will be used. The modular nature of both the container 10 and the shielding inserts avoids serial manufacturing that conventional integral welded containers require. The use of the modular container 10 and the modular shielding inserts allows multiple container variants to be assembled and delivered in response to varying project demands and batch-specific waste streams. It also enables market driven costing and a robust supply chain. The modular design of the components makes it well suited for local sourcing of supply and production allowing for maximum diversity, flexibility, and localization. It also allows for multiple material options that facilitate competitive sourcing and allows for reduced lead time for material, production, and delivery. The shielding inserts are self-locking and self-supporting. Once in place, the shielding inserts do not need any additional structure or joining to support them. The shielding inserts are positioned so that the seams between the inserts do not provide a direct path for radiation shine to the enclosure envelope 12. Referring to FIGS. 5-9, the modular container 10 includes a lid shielding insert 50 (alternatively referred to as a top shielding insert, lid shielding slab, or top shielding slab), a base shielding insert or slab 52 (alternative referred to as a bottom shielding insert, base shielding slab, or bottom shielding slab), wall shielding inserts 54 (alternatively referred to as wall shielding slabs), and corner shielding inserts 56 (alternatively referred to as corner post shielding inserts or corner posts). In one embodiment, the lid shielding insert 50 and/or the base shielding insert 52 have stepped edges 58 that register with the wall shielding inserts 54 as shown best in FIGS. 6 and 9. In another embodiment, the base shielding insert 52 is flush and an additional plate is inserted on top of the base shielding insert 52 to secure the wall shielding inserts 54 in place. The wall shielding inserts 54 have relaxed tolerances to make it easy to assemble the inserts. The corner shielding inserts 56 have tighter tolerances to secure wall shielding inserts 54 in place. The wall shielding insert 54 on the left side of the modular container 10 in FIGS. 6 and 7 includes a hole 60 for the vent 40. The joints between the wall shielding inserts 54 and the lid and base shielding inserts 50, 52 have a stepped geometry. The joints between the wall shielding inserts and the corner shielding inserts 56 have an oblique geometry. In one embodiment, the joints between the shielding inserts 50, 52, 54, 56 can be caulked or otherwise filled to provide an additional barrier to prevent migration of fines and loose particulates in those applications that require it. An example of one type of suitable filler material is inorganic silicone sealant. Alternatively, a drop-in liner can be used for waste that contains a significant amount of loose material. It should be appreciated that the shielding inserts can have any suitable configuration that allows them to securely fit inside the cavity 14 of the modular container 10. Also, the modular container can include more or less than four shielding inserts. For example, the corner shielding inserts can be integrated into the wall shielding inserts 54 or the base shielding insert can be provided in multiple pieces. Numerous variations are possible. In one embodiment, the modular container 10 has a cavity 14 with the dimensions shown in the table below. The size of the cavity 14 changes depending on the thickness of the shielding inserts. In general, the shielding inserts can range in thickness from 1 inch to 12 inches (25 mm to 305 mm). The size of the cavity 14 is shown in the table for a given shielding insert thickness. TABLE 3Shielding Thickness and Cavity Size ofOne Embodiment of the Modular ContainerShielding InsertCavityCavityCavityCavityThicknessLengthWidthHeightVolume(in)(mm)(in)(mm)(in)(mm)(in)(mm)(ft3)(m3)NoneNone80.00203263.00160060.0015241755.04.010271.75182254.75139151.7513141183.36.015267.75172150.75128947.751213952.79.022961.75156844.75113742.751086681.912.030555.75141638.7598436.75933461.3 The shielding inserts can be made from any suitable type of shielding material such as metallic or cementitious materials. The modular container 10 can include shielding inserts made of the same material or different materials. For example, the base shielding insert 52 can be made of one material and the corner shielding insert 56 can be made of another material. In general, the shielding inserts are non-structural, low precision simple shapes. Suitable materials for the shielding inserts include metal material such as steel (wrought, cast, or rolled), cast iron, lead, and depleted uranium metal. The metal material can be virgin or recycled. Another suitable material can be high density concrete such as: (1) heavy aggregates per ACI-211.1, ACI-304, and ASTM C637 & C638 (the ASTM standards describe radiation shielding concrete), (2) depleted uranium aggregate per ASTM C289 & C295 and BS 6073. The following table provides some additional examples of suitable shielding materials. TABLE 4Examples of Shielding Insert MaterialsShielding Unit Weightlb/ft3 (g/cm3)Cast High Density Concrete withWelded Wire Fabric Reinforcement(cast with various commerciallyavailable virgin aggregate types)Limonite/goethite hydrous iron ores180-195(2.9-3.1)Barite/baryte barium sulfate205-225(3.3-3.6)Ilmenite/hematite/magnetite iron ores215-240(3.4-3.8)Steel/iron shot, pellets or punchings310-350(5.0-5.6)Cast from Recycled Low SpecificActivity Contaminated Steel (up to 100%by volume)Ductile/nodular cast iron slabs450(7.2)Cast ferritic steel slabs470(7.5)Cast austenitic steel slabs490(7.6)Manufactured from Virgin Materials(procured from commercialMetals16 mills/foundries)Ductile/nodular cast iron slabs450(7.2)Steel casting slabs470(7.5)Hot rolled carbon steel plate490(7.6)Cast from Recycled Other Low Specific ActivityMetals (up to 100% by volume)Depleted uranium concrete (Ducrete)559(8.9)Cast lead705(11.3)Cast depleted uranium metal1,192(19.1) In one embodiment, the shielding inserts are made of 100% recycled metal. The use of 100% recycled metal provides very efficient shielding for better ALARA (as low as reasonably achievable). ALARA refers the radiation safety principle for minimizing radiation doses and releases of radioactive materials by employing all reasonable methods. Examples of recycled metals that can be used to make the shielding inserts include LSA contaminated steel, lead, DU metal, and DU aggregate in DU concrete. Recycling of radioactive waste metals reduces the total volume of contaminated metal that needs to be disposed. Recycling also eliminates the need to package and dispose of such waste metal separately. The unit costs for recycling may initially be higher compared with virgin material, but lifecycle costs are lower considering avoided disposal costs. The structural lid 18 and the lid shielding insert 50 together form a lid assembly that can be installed in the field as a single assembled unit. For example, the structural lid 18 and the lid shielding insert 50 are coupled together and the entire lid assembly is coupled to the main body 30. Alternatively, the structural lid 18 and lid shielding insert 50 can be installed separately. For example, the lid shielding insert 50 is placed over the cavity 14 then the structural lid 18 is coupled to the main body 30. Likewise, the lid shielding insert 50 can be removed with the structural lid 18 or separate from the structural lid 18. For example, it may be desirable to remove them both together to access sub-container process fittings for stabilizing waste. It may be desirable to remove them separately to enable replacement of the seal member 44 of the modular container 10 while still keeping the shielding in place. In this situation, the structural lid 18 is removed while the lid shielding insert 50 remains in place to allow the seal member 44 to be replaced. Referring to FIG. 10, the lid shielding insert 50 can be coupled to the structural lid 18 in a variety of ways. In general, the lid shielding insert 50 should be coupled to the structural lid 18 so that any holes formed in the structural lid 18 are sealed. This can be accomplished in a variety of ways. In one embodiment, the lid shielding insert 50 is coupled to the underside of the structural lid 18 with fasteners 62 that extend through holes 64 in the structural lid 18. In one embodiment, fasteners 62 are shoulder bolts having a threaded shoulder that seals the hole 64 in the structural lid 18. In another embodiment, the fasteners 62 are self-sealing cap screws. In yet another embodiment, the fasteners 62 are threaded wedge anchors drilled into the lid shielding insert 50. Numerous variations are possible. Grade A Container Variant FIGS. 11-13 show one embodiment of a Grade A modular container variant 70 (alternatively referred to as the Grade A container or Grade A modular container). The Grade A container variant 70 is configured to be used with higher activity radioactive waste including some types of SNF, HLW, and other wastes with the longest lived isotopes. It should be appreciated, however, that the Grade A container variant 70 can be used with any class or type of radioactive waste. In one embodiment, the Grade A container variant 70 is especially suited for certain types of SNF and HLW such as those that: (a) have small profiles and unique configurations that make them less suitable for storage in a large LWR spent fuel cask—e.g., fuel having a relative small cross-section and short length, (b) generate low container internal pressures—e.g., rod/cladding pressures are low compared to typical LWR fuels, (c) have low decay heats compared to typical LWR fuels—e.g., sealed/inerted container is adequate for heat removal, (d) are from smaller/older facilities with physical constraints and limited capability to handle large conventional LWR casks. The Grade A container variant 70 which is the most robust container variant is similar in many ways to the modular container 10 shown in FIGS. 1-4 except that a number of the features of the Grade A container variant 70 have been upgraded to accommodate the higher activity of the radioactive waste. For example, upgraded features can include the materials and design details such as weld joints used to make the container 70 and especially the enclosure envelope 12, the structural lid 18 and seal members 44, additional features for draining, drying, leak testing, and inerting the interior of the cavity 14. In one embodiment, the enclosure envelope 12 and/or the entire Grade A container variant 70 is made in compliance with ASME Section III (materials, fabrication, and testing). For example, the enclosure envelope 12 can be made of ½ inch to 1 inch (12 mm to 26 mm) thick stainless steel plate with complete penetration weld joints with geometric transitions to provide better impact toughness, weld strength, pressure rating, and corrosion resistance. In one embodiment, the enclosure envelope 12 and/or the entire Grade A container variant 70 can be made of SA240 Type 316/316L austenitic stainless steel for increased long term corrosion performance. This configuration is suitable for a normal operating pressure of 20-25 psig. In another embodiment, the enclosure envelope 12 and/or the entire Grade A container variant 70 can be made of high strength material such as SA-240 Type XM-19 (nitronic 50 alloy) austenitic stainless steel or SA-693 Type 630 (17-4PH alloy) martensitic stainless steel to accommodate higher internal pressures and more severe postulated drop accident conditions. Referring to FIGS. 13-15, the walls 16 of the enclosure envelope 12 for the Grade A container variant 70 can be made of two U-shaped or four L-shaped side plates 72 to provide increased strength and load capacity instead of the four plates with corner joints shown in FIG. 4. In this embodiment, the plates 72 can be formed with 2t radius inside corners and coupled together using mid-wall weld joints to provide a transition for bending stresses. In one embodiment, the plates 72 are butt welded using full penetration welds (FIGS. 16-17 show examples of full penetration welds). The welds can be examined using volumetric, non-destructive, radiographic testing. The base plate 26 of the enclosure envelope 12 for the Grade A container variant 70 can be machined from a 1.5 inch to 2.5 inch (38 mm to 64 mm) thick plate of steel. Referring to FIGS. 16-17, the base plate 26 can include raised weld necks transition joints 74. The walls 16 are coupled to the weld necks 74 using full penetration butt welds. The welds can be examined using volumetric, non-destructive, radiographic testing. This provides superior strength and robustness compared to partial penetration bottom corner weld joints. In one embodiment, all Grade A container containment welds can be examined using full penetration, non-destructive, radiographic testing (e.g., formed corners, base plate 26 weld neck 74, and the like). Moreover, all mating and sealing surfaces can be machined and sealed with one or more sealing members 44. Referring to FIGS. 12-17, the structural lid 18 of the Grade A container variant 70 and the manner in which it is coupled to the main body 30 can be upgraded. For example, the flange 36 can be cut from a full-size steel plate that is 1 inch to 2 inches thick (25 mm to 51 mm). This eliminates the joints at the corners that result from four bars being joined together as shown in FIG. 4 for the Grade B container variant 10 (and Grade C container variant 120, see below). Referring to FIGS. 16-17, the flange 36 can be coupled to the walls 16 using full penetration welds. Also, the flange 36 can include threaded bores 76 to receive the fasteners 42. The bores 76 extend partially through the flange 36 instead of completely through the flange 36 to eliminate potential leak paths when the structural lid 18 is installed. The structural lid 18 for the Grade A container variant 70 can include a one-piece, solid, unbroken lid plate 78 machined from steel having of thickness of 2 inches to 3 inches (50 mm to 77 mm). The lid plate 78 is stepped in the manner shown in FIGS. 16-17 so that the outer edges of the lid plate 78 fit over the upper edge of the walls 16. The structural lid 18 for the Grade A container variant 70 can be coupled to the main body 30 using upgraded fasteners 42. Examples of suitable fasteners 42 include SA-320 grade L43 pressure vessel flathead bolts. In one embodiment, the fasteners 42 are recessed into the structural lid 18 to protect the fasteners 42 during handling or stacking. In another embodiment, the Grade A container variant 70 is configured to allow the fasteners 42 to be easily installed while standing at grade level using standard tools. The structural lid 18 and the top lip of the walls 16 form a shear key that resists lateral loads. The fasteners 42 are not loaded in the shear plane. Also, there are no welds in the shear plane, which serves to increase the strength of the Grade A container variant 70 and the robustness of the seal between the structural lid 18 and the main body 30. The structural lid 18 can be sealed to the main body 30 in any suitable manner. In one embodiment, the structural lid 18 and the main body 30 are sealed together using one or more sealing members 44. Referring to FIGS. 16-17, the structural lid 18 for the Grade A container variant 70 can be sealed to the flange 36 using at least two sealing members 44 that fit in corresponding grooves 82 in the structural lid 18. In one embodiment, the sealing members 44 are O-ring seals made of an elastomeric material such as butyl rubber. The lid plate 78 for the Grade A container variant 70 can also be welded to the flange 36 along its entire perimeter as shown in FIGS. 16-17 to provide a more robust seal. The Grade A container variant 70 can be sealed to prevent leaks. In one embodiment, the Grade A container variant 70 provides leak tightness to at least 10−6 cm3/see via pressure drop test on the interspace between the sealing members 44 for the rated maximum normal operating pressure (unvented). The enclosure envelope 12 of the Grade A container variant 70 is typically pressure retaining and is not vented in this instance. In one embodiment, a test port is provided to test the pressure drop on the interspace between the sealing members 44. Referring to FIG. 17, the structural lid 18 can include a top cover plate 80 to further seal the Grade A container variant 70. The top cover plate 80 is coupled to the lid plate 78 by welding or other suitable techniques to seal the top of the structural lid 18. In one embodiment, the top cover plate 80 can be welded to the lid plate 78 along the entire perimeter. The top cover plate 80 can be used to eliminate the need for continuous pressure monitoring of seals during interim storage or as an alternative to seal replacement following extended interim storage. The Grade A container variant 70 can be used to remotely load waste in a wet or dry environment. In one embodiment, the Grade A container variant 70 includes features for draining, drying, leak testing, and inerting the cavity 14. These features are useful in situations where the Grade A container variant 70 is loaded underwater, such as in a SNF pool. These features allow the Grade A container variant 70 to be drained, dried, and leak tested after being loaded with waste. They can also be used to fill the cavity 14 with an inert gas. Referring to FIGS. 18-20, the Grade A container variant 70 includes a vent port 82 and a drain 84. The vent port 82 is a dog-leg passage through the lid plate 18 and wall shielding insert 54 to the cavity 14. Excess gas from the cavity 14 can be vented through the vent port 82. The ports 82, 84 can be sealed shut with the port covers 92 after the Grade A container variant 70 has been drained, dried, leak tested, and/or inerted. In one embodiment, the port covers 92 can be welded to the lid plate 78. A drain tube 90 passes through the drain port 84 and through the shielding inserts 50, 52, 54 to the bottom of the enclosure envelope 12. The drain tube 90 can be used to remove water from the cavity 14. Fittings 86 can be coupled to the structural lid 18 to allow processing equipment to be coupled to the ports 82, 84. For example, the fittings 86 can be Swagelok type fittings that can be coupled to a vacuum drying skid or the like. Referring to FIGS. 19-20, the base shielding insert 52 can include grooves 88 that convey water to the drain tube 90. Once the cavity 14 is dry, it can be filled with an inert gas such as helium to assist with removing excess heat and maintaining the integrity of the SNF. Referring to FIG. 21, the Grade A container variant 70 can include an energy absorbing material 94 positioned inside the support members 22, 24. The energy absorbing material 94 increases the robustness and energy absorbing capability when the Grade A container variant 70 is accidentally dropped. The support members 22, 24 are crushed when the Grade A container variant 70 is dropped and the energy absorbing material 94 helps to absorb the energy of the drop. The energy absorbing material 94 increases a number of parameters associated with the container—e.g., increases the lifting and handling capacity, increases stack height for interim storage, increases range of transportation conditions, increases stack height for geological disposal, and so forth. The energy absorbing material 94 can be any suitable material. In one embodiment, the energy absorbing material 94 includes an energy absorbing foam material. The energy absorbing foam can be the kind widely used in Type B transportation package impact limiters—e.g., LAST-A-FOAM from General Plastics. Referring to FIGS. 22-25, the Grade A container variant 70 can include an impact limiter 96 that can be coupled to the top of the Grade A container variant 70. The impact limiter 96 is shaped similarly to the base 20. In one embodiment, the impact limiter 96 can be coupled to the Grade A container variant 70 using the lifting members 28. For example, fasteners such as bolts can extend through the hole in the lifting members 28 and into a threaded hole 98 on the impact limiter 96. In one embodiment, the impact limiter 96 includes a main body 100 and a cover plate 102. The main body 100 can be filled with the energy absorbing material 94 as shown in FIG. 25. The cover plate 102 is fastened over the open side of the main body 100 through any suitable means such as welding and the like. The impact limiter 96 can be used selectively as needed, for example, for lifts in excess of 15 feet (4.6 m) and for the highest container in a stack (interim storage or geological disposal). Referring to FIGS. 26-28, the Grade A container variant 70 can also include additional neutron shielding panels 104, 106, 108 (alternatively referred to as neutron shielding members). These are useful when the radioactive waste has significant neutron activity. The neutron shielding panels 104, 106, 108 can be filled with any suitable neutron shielding material including hydrogenous material such as that specified in NS-4-FR. The Grade A container variant 70 includes a top neutron shielding panel 104, side neutron shielding panels 106, and bottom neutron shielding panels 108. The neutron shielding panels 104, 106, 108 can be fabricated separately and then fastened to the outside of the Grade A container variant 70 using any suitable fastener or fastening technique, for example welding. The top and side neutron shielding panels 104, 106 include shielding inserts 110 coupled to a flat plate 112 as shown in FIG. 28. The bottom neutron shielding panels 108 just include the shielding inserts 110. The top neutron shielding panel 104 is configured so that the shielding inserts 110 correspond to the cavities or recesses in the bottom of the base 20 of the main body 30 (see FIG. 2). When the Grade A containers 70 are stacked, the shielding inserts 110 extend into the cavities in the base 20. The top neutron shielding panel 104 is also compatible with the impact limiter 96. In one embodiment, the shielding inserts 110 on the top neutron shielding panel 104 fill up one half of the cavity in the base 20 and the shielding inserts 110 that form the bottom neutron shielding panels 108 are configured to fill up the other half of the cavity. The shielding inserts 110 on the side neutron shielding panel 106 are positioned to fit in and fill up the spaces between the support members 24. Grade B Container Variant The modular container 10 shown in FIGS. 1-10 constitutes one embodiment of a Grade B modular container variant that is less robust than the Grade A container variant 70, but more robust than the Grade C container variant 120. As such, the modular container 10 is referred to as the Grade B modular container variant 10 in the following description (alternatively referred to as the Grade 13 container or Grade B modular container). The Grade B modular container variant 10 is configured to be used primarily with intermediate activity radioactive waste such as class B waste, class C waste, GTCC waste, and ILW. It should be appreciated, however, that the Grade B modular container variant 10 can be used with any class or type of radioactive waste including special form waste, TRU waste, and other waste with longer lived isotopes. The Grade B container variant 10 can be made of any suitable materials. In one embodiment, the enclosure envelope 12 and/or the entire Grade B container variant 10 is made in compliance with ASME Section VIII (materials, fabrication, and testing). For example, the enclosure envelope 12 and/or the entire Grade B container variant 10 can be made of A240 Type 304/304L stainless steel. The various components can be welded together using at least partial penetration welds inspected using non-destructive examination techniques. Referring to FIGS. 3-4, the walls 16 of the enclosure envelope 12 of the Grade B container variant 10 can be made of four side plates with corner joints. In one embodiment, the plates are welded using corner welds with partial or complete joint penetration (FIGS. 16-17 show examples of full penetration welds). The welds can be examined using non-volumetric, non-destructive, dye penetrant testing. The base plate 26 of the enclosure envelope 12 of the Grade B container variant 10 can be a steel plate. The walls 16 are coupled to the base plate 26 using corner welds with partial or complete joint penetration. The welds can be examined using non-volumetric, non-destructive, dye penetrant testing. FIG. 29 shows the manner in which the structural lid 18 is coupled to the main body 30 of the Grade B container variant 10. The flange 36 is formed by four separate pieces coupled together at the corners, preferably by welding (see FIG. 4 to see the separate pieces). The flange 36 is coupled to the outside of the walls 16 just below their top edges. The flange 36 can be coupled to the walls 16 in any suitable manner such as by welding (at least partial penetration weld). The shear key spacer 38 is positioned above the flange 36 so that the top of the spacer 38 is flush with the top of the walls 16. The spacer 38 can be coupled to the walls 16 by welding or the like (at least partial penetration welds). One or more sealing members 44 is positioned between the flange 36 and the spacer 38. The sealing member 44 can be any suitable material. In one embodiment, the sealing member 44 includes a one-piece flat elastomeric gasket seal (butyl rubber). In another embodiment, all mating and sealing surfaces are machined, including, but not limited to the surfaces that contact the sealing member 44. The structural lid 18 is coupled to the main body 30 with the fasteners 42 that extend through the spacer 38 and the flange 36. In one embodiment, the fasteners 42 include 316 stainless steel flathead bolts, lock washers, and nuts. In another embodiment, the fasteners 42 are recessed into the structural lid 18 to protect them during container handling. The fasteners 42 can easily be replaced if damaged because there are no threaded parts to repair. The fasteners 42 can also easily be installed while standing at grade level using standard tools. The structural lid 18 forms a stepped arrangement when coupled to the spacer 38 and flange 36. The stepped arrangement of the structural lid 18, spacer 38, and flange 36 combined with the upper lip of the walls 16 forms a shear key that resists lateral loads. Also, the fasteners 42 are not subject to loads in the shear plane. The structural lid 18 can be welded to the spacer 38 in the manner shown in FIG. 29 to provide a more robust seal. The Grade B container variant 10 can be configured to withstand leaks. In one embodiment, the Grade B container variant 10 provides leak tightness to 10−4 cm3/sec via pressure drop test on the cavity 14. The Grade B container variant 10 is suitable for a normal operating pressure of approximately 10 psig (unvented). The enclosure envelope 12 of the Grade B container variant 10 can be pressure retaining or non-pressure retaining using a filtered vent in the latter instance. Grade C Container Variant FIGS. 30-32 show one embodiment of a Grade C modular container variant 120 (alternatively referred to as the Grade C container or Grade C modular container) that is less robust than the Grade A and B modular container variants 70, 10. It is configured similarly to the Grade B container variant 10 so that the description of the Grade B container variant 10 applies unless noted otherwise. The Grade C container variant 120 is configured to hold lower activity radioactive waste such as Class B waste, Class C waste, and ILW. In general, it is especially suitable for storing shorter lived radioactive isotopes. It should be appreciated, however, that the Grade C container variant 120 can be used with any class or type of radioactive waste. The Grade C container variant 120 can be made of any suitable materials. In one embodiment, the enclosure envelope and/or the entire Grade C container variant 120 are made of AISC (American Institute of Steel Construction) and AWS (American Welding Society) materials, fabrication, and testing. For example, the Grade C container variant 120 can be made of A36 epoxy coated carbon steel. In general, the material used to make the Grade C container variant 120 is thinner than the material used to make the Grade B container variant 10. The walls 16 and base plate 26 for the Grade C container variant 120 can be joined in largely the same manner as the Grade B container variant 10. FIG. 33 shows that the structural lid 18 can be coupled to the main body 30 in largely the same manner as the Grade B container variant 10. It should be noted that the materials used for the structural lid 18, flange 36, and spacer 38 are thinner than that used for the Grade B container variant 10. Also, fewer fasteners 42 are used to couple the structural lid 18 to the main body 30 and the fasteners 42 can be standard flathead bolts, lock washers, and nuts. The Grade C container variant 120 can be configured to withstand leaks. In one embodiment, the Grade C container 12 provides leak tightness to 10−4 cm3/sec via pressure drop test on the cavity 14. The Grade C container can also be designed for a comparatively lower normal operating pressure or is non-pressure retaining using a filtered vent in the latter instance. Solid and Granular Radioactive Waste The modular container 10 can be configured to hold a variety of physical manifestations of radioactive waste including solid and granular radioactive waste. It should be appreciated that the remainder of the description refers to the modular container 10 but applies to all of the grade variants of the modular container 10, 70, 120. Solid radioactive waste can generally be considered waste that maintains its physical form or, in other words, waste that is not in a liquid form (e.g., no more than 1% liquid by volume) or a gaseous form. It can optionally be considered to exclude certain materials specified by regulation. For example, in one embodiment, solid radioactive waste excludes material that exceeds strategic quantities of special nuclear material as referenced in 10 CFR 70.4. Solid radioactive waste includes both processed and unprocessed bulk waste. In one embodiment, solid radioactive waste can be processed using any of a variety of process. Examples of suitable processes include segregation, decontamination, size reduction, volume reduction, and the like. Solid radioactive waste is distinguished conceptually from granular radioactive waste in that solid radioactive waste has a relatively low volume of fines and loose particulate. The following table describes different waste classes and waste types along with typical configurations of the modular container that accommodate them. TABLE 5General Modular Container Solid Radioactive Waste Acceptance CriteriaModular ContainerWaste ClassWaste TypeConfigurationClass B and C low level wasteSurface contaminated objectsShielding inserts to suit waste(LLW)(SCO) with low levels of long-livedactivityWaste with concentrations of shortradionuclidesSolid waste with minimal fines andand long lived radionuclides thatActivated materials with lowloose particulate, e.g., non-fuelexceed the Class A limits specifiedspecific activity (LSA)bearing fuel components or primaryin 10 CFR 61.55, but do not exceedContact dose rates may exceed 2.0 mSv/hrsystem componentsthe limits for Class B and C(200 mrem/hr) and warrantSupport framework for drummedshielded packagingsolid wasteOther solid radioactive waste thatMay require use of a Type Bmeet the applicable criteriatransport overpack for transportGreater Than Class C (GTCC)SCO with higher levels ofThicker shielding inserts to suitWaste with concentrations of shortlong-lived radionuclideshigher waste activityor long lived radionuclides thatHighly activated materialsSolid waste with minimal fines andexceed the limits specified in 10Sealed sourcesloose particulate, e.g., reactorCFR 61.55 and that is not HLW,Contact dose rates typically exceedinternalsspecial form, or transuranic waste2.0 mSv/hr (200 mrem/hr) andSupport framework for drummedPrimarily from commercial sourcesrequire shielded packagingsolid waste(as opposed to government)Other solid radioactive waste thatTypically requires use of Type Bmeet the applicable criteriatransport overpack for transportTransuranic (TRU)SCO with transuranic radionuclidesMinimal or no shielding inserts forIncludes contact handled TRU (CH-CH-TRU wastes have surface doseCH-TRU wastes. SupportTRU) and remote handled TRUrates <200 mrem/hr and typicallyframework for drummed CH-TRU(RH-TRU)require only a lightly shieldedwasteWaste with alpha-emittingpackagingThicker shielding inserts to suit RH-transuranic radionuclides havingRH-TRU wastes have surface doseTRU wastes. Support frameworkhalf-lives and concentrations thatrates ≧200 mrem/hr and require afor drummed RH-TRU wasteexceed 20 years and 100 nCi/g.more heavily shielded packagingTypically requires use of Type BPrimarily from government sourcestransport overpack for transport(Department of Energy)Special form wasteIndispersible radioisotope materialShielding inserts to suit wasteSealed sources with high activitySealed capsules containingactivitybut low contaminationradioisotope materialContents-specific insert for sealedContact dose rates typically exceedsources2.0 mSv/hr (200 mrem/hr) andMay require use of Type Brequire shielded packagingtransport overpack for transportIntermediate level waste (ILW)SCO with higher levels ofThicker shielding inserts to suitLong-lived alpha emitterslong-lived radionuclideshigher waste activity>4 GBq/tonne (108 mCi/tonne)Highly activated metalsSolid waste with minimal fines andLong-lived beta and gammaSealed sourcesloose particulate, e.g., reactoremitters >10 GBq/tonneContact dose rates may exceedinternals(270 mCi/tonne)2.0 mSv/hr (200 mrem/hr) andSupport framework for drummedThermal power <2 kW/m3warrant shielded packagingsolid wasteOther solid wastes that meet theTypically requires use of Type Bapplicable criteriatransport overpack for transportHigh level waste (HLW)Spent nuclear fuel elementsUpgraded modular container envelopLarge concentrations of bothFission product and actinide wastewith contents-specific insert for;short- and long-lived radionuclidesfrom reprocessingVitrified HLW reprocessing wasteTypically has activityPackaging requires criticalityMetallic uranium fuels such asconcentrations in range of 104 tocontrol features, heavy shielding,research reactor fuel106 TBq/m3more robust containment function,Natural uranium fuelTypically has thermal powerand greater heat removal capabilityOther defense and research reactor>2 kW/m3 ≦ 20 kW/m3fuelsTypically requires decay in interimstorage prior to disposalSolid radioactive waste with significant fines and loose particulate may be considered granular waste that should be placed in a liner in the modular container FIG. 34 shows one embodiment of the modular container 10 filled with solid waste 122. In this embodiment, the solid waste 122 is in the form of canisters or tubes. The solid waste 122 is placed directly in the cavity 14 and surrounded by the shielding inserts 50, 52, 54, 56. It should be appreciated that the solid waste 122 is merely an example of one form of solid waste that can be put in the modular container 10 and that other forms can be put in as well. FIGS. 35-36 show another embodiment of the modular container 10 filled with granular waste. It should be appreciated that granular waste refers to solid waste that has a relatively high volume of fines and/or loose particulates. A liner 124 is positioned in the cavity 14 to hold the granular waste. The liner 124 is a single, integral piece that has all of its seams sealed to prevent the granular material from escaping. The liner 124 is sized to fit in the cavity 14 and flanged on the top to provide a better fit and seal. The liner 124 can be made of any suitable material such as formed polyethylene (e.g., 1/16″ to 3/16″ thick, etc.), fabricated stainless steel (e.g., 11-16 gauge, etc.), or fabricated galvanized steel (e.g., 11-16 gauge, etc.). An additional gasket can be added between the liner 124 and the structural lid 18 (or shielding insert 50 on the bottom of the structural lid 18) to provide additional loose contamination barrier. As previously mentioned, the design of the structural lid 18 allows the sealing member 44 that forms the outer confinement boundary to be replaced without removing shielding insert 50 that covers the liner 124. Wet Radioactive Waste The modular container 10 can be configured to hold a wet radioactive waste. Wet radioactive waste can generally be considered waste that does not maintain its physical form. It includes liquids, slurries (liquid plus suspended solids), sludge (wet solids), or dry solid particles. Wet radioactive waste includes bulk low level waste and mixed radioactive and hazardous waste. Wet radioactive waste is typically processed to stabilize and/or solidify the waste. The waste is stabilized by dewatering to remove excess water. It is solidified to chemically bind the waste into a monolithic solid and/or encapsulate the waste by surrounding it with a binder or coating. Wet radioactive waste can be processed in the container (e.g., a 55 gal (208 liter) drum) or separately to chemically and physically stabilize the waste and prevent it from dispersing waste fines or liquids. The following table describes different waste classes and waste types along with typical configurations of the modular container 10 that can accommodate them. TABLE 6General Modular Container Wet Radioactive Waste Acceptance CriteriaModular ContainerWaste ClassWaste TypeConfigurationClass B and C low level wasteWet solids including spent ionProcessed waste in sub-containers(LLW)exchange media such as powderedsuch as drums or high integrityWaste with concentrations of shortresins, bead resins and Zeolitescontainers (HICs).and long lived radionuclides thatSludges such as carbon andDrums or other sub-containers ofexceed the Class A limits specifiedcellulose filter media, andvarious sizes with encapsulatedin 10 CFR 61.55, but do not exceeddiatomaceous earthwastesthe limits for Class B and CLiquids and slurries includingDewatered resins with radiologicalLiquid wastes or solid wastesevaporator concentrates such asactivity >1 μCi/cc in HICcontaining liquid should besodium sulfate and boric acid,sub-containers.converted to a stable form thatreverse osmosis concentrate,Support framework for processedcontains ≦1% free liquid by volumedecontamination liquids, andwaste in drums and HICsand otherwise meets 10 CFR 61.56contaminated oilsModular container with shieldingOther wet or dry solids such asinserts to suit waste activitycalcine, incinerator ash, and otherModular container may require usemiscellaneous wasteof Type B transport overpack fortransportModular container and/or supportframework may be disposed withsub-container The modular container 10 can accommodate a variety of forms of wet radioactive waste. In particular, the modular container can accommodate stabilized wet radioactive waste in industry standard sub-containers. The modular container 10 can be used to process wet radioactive waste. For example the modular container 10 can be used for in-container filling and waste dewatering and/or stabilization using existing systems. The modular container 10 can also be remotely loaded with externally processed and legacy sub-containers. FIG. 37 shows one embodiment of the modular container 10 loaded with sub-containers 126 filled with wet radioactive waste. The sub-containers 126 can be positioned in various support frameworks or stillages 128. The support frameworks 128 can include a support base 132, support posts 134, and a top insert 136. Referring to FIGS. 37-40, the top shielding insert 136 can include holes 140 through which the mechanical fittings of processing equipment can be coupled to the sub-containers 126 for in-container stabilization of liquid waste. The lid shielding insert 50 can also have holes 142 that correspond to the holes 140. This allows the sub-containers 126 to be processed while the shielding insert 50 in place. When processing is completed, the holes 142 can be filled with shielding plugs 144. The support framework 128 can be used to facilitate a number of functions. For example, the support framework 128 can be used to accurately position the sub-containers 126 in the cavity 14 of the modular container 10. The support framework 128 also provides a way to handle multiple sub-containers 126 simultaneously. The support framework 128 can also mitigate the consequences of accidentally dropping the modular container 10. The support framework 128 can confine and protect the sub-containers 126 in the drop. It can also provide additional crushable material to absorb the energy of the drop. The support framework 128 can be lifted using a forklift, crane, or the like. In one embodiment, the support framework 128 includes swivel hoist rings coupled to the top of the support posts 134 to facilitate lifting. In another embodiment, a strap or sling can be positioned in the base pockets of the support framework 128 to facilitate lifting. The sub-containers 126 can be loaded into modular container 10 in a couple of ways. One way is to put the sub-containers 126 in the support framework 128 and then load the support framework into the modular container 10. Another way is to put the support framework 128 in the modular container 10 and then load the sub-containers 126 into the support framework 128. The sub-containers 126 can be filled with waste before or after they are loaded into the support framework 128 and/or the modular container 10. In one embodiment, the support framework 128 includes additional auxiliary bins 130 positioned in the interstitial space between the sub-containers 126. The auxiliary bins 130 are generally tubular and can be filled through one or more openings at the top. The auxiliary bins 130 can be closed shut by the top insert 136. They can be used to hold additional solid radioactive waste together with the wet radioactive wastes in the one or more sub-containers 126. In one embodiment, the auxiliary bins 130 include bottoms and are removable from the support framework 128. This may make it easier to fill and load the auxiliary bins 130. In another embodiment, the auxiliary bins 130 have no bottoms and are coupled to the support base 132. Examples of suitable solid radioactive waste that can be placed in the auxiliary bins 130 include irradiated and/or contaminated hardware items, dewatered/stabilized filters, and/or granular waste. The auxiliary bins 130 can also be filled with shielding material to provide extra shielding should the activity of the waste require it. Suitable shielding material includes steel shot, concrete, and the like. The auxiliary bins 130 can also be left empty. The modular container 10 can have a variety of shielding configurations when loaded with wet radioactive waste as shown in the table above. In general, the shielding inserts 50, 52, 54, 56 range in thickness from 2 inches to 8 inches (50 mm to 204 mm) with the most common being 4 inches and 6 inches (101 mm to 153 mm). The enclosure envelope 12 of the modular container 10 provides additional shielding with a typical configuration providing approximately 0.5 inches to 1.5 inches (12 mm to 39 mm) of steel shielding (e.g., 0.75 inches (19 mm) of steel shielding). Referring to FIGS. 40-41, the support framework 128 can also include supplemental shielding members 138. In one embodiment, the shielding members 138 are panels of steel with a typical configuration providing an additional 0.5 to 1.5 inches (12 mm to 39 mm) of steel shielding (e.g., 1 inch (25 mm) of steel shielding). It should be appreciated that the shielding members 138 can be made of any suitable shielding material and have any suitable shape or configuration. The modular container 10 can accommodate a variety of standard sub-containers as shown in the table below. Each configuration is described below. It should be appreciated that the above discussion related to wet radioactive waste and the various components of the modular container 10, support framework 128, and the like apply to the following configurations unless noted otherwise. TABLE 7Typical Modular Container and Sub-Container Configurations for Wet WasteSub-ContainerShieldingSupport Framework ConfigurationArrangementInsertUsableAux. BinOptionalVolumein ModularThicknessCavity Vol.Auxiliary BinUsable Vol.ShieldLiftingTypegal (liter)Containerinches (mm)ft3 (m3)Configurationft3 (m3)PlatesModesDrum55 (210)2 × 3 = 64 (100)118 (3.3)2 center + 4 side = 612.3 (0.35)1 in.Forklift orthickCraneDrum55 (210)2 + 1 + 2 = 56 (150) 95 (2.7)2 side 7.8 (0.22)1 in.Forklift orthickCraneDrum85 (322)2 + 1 + 2 = 54 (100)118 (3.3)2 side15.0 (0.42)⅝ in.Forklift orthickCraneDrum85 (322)1 × 2 + 1 × 2 = 46 (150) 95 (2.7)2 corner + 2 side = 417.3 (0.49)⅝ inForklift orthickCraneDrum132 (500) 1 × 2 = 24 (100)118 (3.3)2 side + 2 side = 428.1 (0.80)1 in.Crane or(special)thickIndividualDrum132 (500) 1 × 2 = 26 (150) 95 (2.7)4 side + 4 side = 841.3 (1.17)NoneIndividual(special)Cylindrical400 (1500)1 × 1 = 14 (100)118 (3.3)4 corner + 2 end = 625.4 (0.72)1 in.Crane orHICthickIndividualCylindrical300 (1125)1 × 1 = 16 (150) 95 (2.7)4 corner + 2 end = 619.0 (0.54)NoneCrane orHICIndividualCuboidal1200 (4540) 1 × 1 = 1None175 (5.0)NoneNoneNoneIndividualHIC or LinerCuboidal800 (3030)1 × 1 = 14 (100)118 (3.3)NoneNoneNoneIndividualHIC or LinerCuboidal600 (2270)1 × 1 = 16 (150) 95 (2.7)NoneNoneNoneIndividualHIC or LinerExcludes additional shielding provided by container enclosure envelope - e.g., 0.75 inches (19 mm) additional shielding Referring to FIGS. 37-41, one embodiment of the modular container 10 having 4 inch (102 mm) shielding inserts 50, 52, 54, 56 loaded with sub-containers 126 that are 55 gallon (208 liter) drums filled with radioactive waste. The sub-containers 126 are held in position by the support framework 128. In this configuration, the modular container 10 can hold six of the 55 gallon (208 liter) drums. The support framework 128 includes two auxiliary bins 130 positioned in the center of the support framework 128. These bins 130 can be used to hold higher activity waste that requires additional shielding and the bins 130 near the edges of the support framework can be used to hold lower activity waste. FIGS. 42-45 show another embodiment of the support framework 128 loaded with 55 gallon (208 liter) drum sub-containers 126 filled with radioactive waste. The support framework 128 is sized to fit in the cavity 14 of the modular container 10 with 6 inch (152 mm) shielding inserts 50, 52, 54, 56. In this configuration, the modular container 10 can hold five 55 gallon (208 liter) drums with two on each end and one in the middle. The support framework 128 also includes two relatively large wedge shaped auxiliary bins 130 positioned on opposite sides of the center sub-container 126. FIGS. 46-49 show another embodiment of the support framework 128 loaded with 85 gallon (322 liter) drum sub-containers 126 filled with radioactive waste. The support framework 128 is sized to fit in the cavity 14 of the modular container 10 with 4 inch (102 mm) shielding inserts 50, 52, 54, 56. In this configuration, the modular container 10 can hold five 85 gallon (322 liter) drums with two on each end and one in the middle. The support framework 128 also includes two relatively large wedge shaped auxiliary bins 130 positioned on opposite sides of the center sub-container 126. FIGS. 50-53 show another embodiment of the support framework 128 loaded with 85 gallon (322 liter) drum sub-containers 126 filled with radioactive waste. The support framework 128 is sized to fit in the cavity 14 of the modular container 10 with 6 inch (152) shielding inserts 50, 52, 54, 56. In this configuration, the modular container 10 can hold four 85 gallon (322 liter) drums in a diamond shape (one in opposite corners and two spread between so the sub-containers 126 form a diamond shape). The support framework 128 also includes two relatively large rectangular shaped auxiliary bins 130 positioned opposite each other on the ends of the support framework 128 and two smaller triangular shaped auxiliary bins 130 positioned opposite each other on the sides of the support framework 128. FIGS. 54-55 show another embodiment of the modular container 10 having 4 inch (102 mm) shielding inserts 50, 52, 54, 56 loaded with sub-containers 126 that are 500 liter drums filled with radioactive waste. The sub-containers 126 are held in position by the support framework 128. In this configuration, the modular container 10 can hold two of the 500 liter drums. FIGS. 56-59 show another embodiment of the support framework 128 loaded with 500 liter drum sub-containers 126 filled with radioactive waste. The support framework 128 is sized to fit in the cavity 14 of the modular container 10 with 4 inch (102 mm) shielding inserts 50, 52, 54, 56. In this configuration, the modular container 10 can hold two 500 liter drums side by side. The support framework 128 also includes four relatively large rectangular shaped auxiliary bins 130 (two on one side two on the other side). FIGS. 60-61 show another embodiment of the modular container 10 having 6 inch (152 mm) shielding inserts 50, 52, 54, 56 loaded with sub-containers 126 that are 500 liter drums filled with radioactive waste. The sub-containers 126 are held in position by the support framework 128. In this configuration, the modular container 10 can hold two of the 500 liter drums. FIGS. 62-63 show another embodiment of the support framework 128 enclosing 500 liter drum sub-containers 126 filled with radioactive waste. In this embodiment, the support framework 128 includes a plurality of auxiliary bins 130 positioned between the sub-containers 126. The bins 130 may or may not be coupled together. Four bins 130 are provided that surround each sub-container 126. It should be appreciated that shielding members 138 can be coupled to the support framework 128 shown in FIGS. 62-63 even though this is not shown in the figures. FIGS. 64-65 show another embodiment of the modular container 10 having 4 inch (102 mm) shielding inserts 50, 52, 54, 56 loaded with a cylindrical HIC (high-integrity-container) sub-container 126 filled with radioactive waste. The sub-container 126 is held in position by the support framework 128. HIC's are widely used in the U.S. and elsewhere to hold stabilized wet waste. In one embodiment, the HIC sub-container 126 can be fitted with internal components to facilitate dewatering of the radioactive waste, cementation of the radioactive waste, or other stabilization processes. Examples of internal components that can be included in or with the HIC sub-container 126 include process piping/tubing, screens, strainers, and process filters, expanded metal grating and foam, mixing paddles, process equipment connectors, and the like. FIGS. 66-67 show the support framework 128 in FIGS. 64-65 enclosing the HIC sub-container 126 filled with radioactive waste. The support framework 128 includes a plurality of auxiliary bins 130 positioned on the ends of the support framework 128. The bins 130 are coupled to the support base 132 and top insert 136. In one embodiment, the top insert 136 includes multiple pieces which close the tops of the auxiliary bins 130 and/or fit over the HIC sub-container 126. FIGS. 68-69 show the support framework 128 in FIGS. 66-67 with shielding members 138 coupled to the exterior of the support framework 128. The shielding members 138 include two U-shaped shielding members that face each other and almost enclose the entire HIC sub-container 126. FIGS. 70-71 show another embodiment of the modular container 10 having 4 inch (102 mm) shielding inserts 50, 52, 54, 56 loaded with a cuboidal HIC (high-integrity-container) sub-container 126 filled with radioactive waste. The sub-container 126 is held in position by the support framework 128. The cuboidal shape of the HIC sub-container 126 makes better use of the available space in the cavity 14 of the modular container 10. It should be appreciated that the HIC sub-containers 126 can be placed in the modular container 10 without the use of any shielding inserts depending on the activity level of the waste. Also, other sub-containers 126 such as bulk liquid transfer sub-containers can be put in the modular container 10 as well. Advanced Gas-Cooled Reactor (AGR) Fuel An AGR fuel bundle includes uranium oxide fuel pellets positioned in stainless steel clad pins separated by spacers and enclosed in a graphite sleeve. The spent AGR fuel bundles are dismantled by removing the graphite sleeve, extracting the fuel pins from the spacers, and consolidating the fuel pins into a slotted can that is about the same size as the original fuel bundle. The slotted cans are then stored in a pool. The modular container 10 can be configured to hold the spent AGR fuel. The Grade A modular container variant 70 is preferable due to the high activity of the AGR spent fuel. It should be appreciated that other grades of the modular container 10 can also be used if the activity of the AGR spent fuel is low enough. FIG. 72 shows one embodiment of the Grade A modular container variant 70 filled with a loading basket 150 full of spent AGR fuel 152, which can be in the form of intact fuel bundles or slotted cans with consolidated fuel pins. The Grade A modular container variant 70 can be configured to hold any amount spent AGR fuel 152. For example, it can be configured to hold approximately 20 to approximately 28 intact fuel bundles or slotted cans. The Grade A modular container variant 70 can include any amount of suitable shielding. In one embodiment, the Grade A modular container variant 70 includes 9-12 inch (228 mm to 305 mm) thick shielding inserts 50, 52, 54, 56. The loading basket 150 is removable from the Grade A modular container variant 70. The loading basket 150 includes a box 154 and dividers 156 (alternatively referred to as divider plates). The box 154 includes lifting members 162 that facilitate handling the loading basket 150. The box 154 can also include holes 158 in the bottom to allow it to drain if it is loaded in a pool. Likewise, the dividers 156 can include recesses, holes or indentations 160 that allow water to flow underneath the divider plates 156 and out the holes 158 in the box 154. FIGS. 73-74 show one embodiment of the loading basket 150 that is configured to be used with less reactive spent AGR fuel 152, for example low enriched consolidated fuel pins. In this embodiment, the box 154 and the dividers 156 are made of stainless steel (any of the stainless steels mentioned in this document) and fit together in the manner shown in FIG. 74. The box 154 and dividers 156 can be secured together using any suitable fastening technique such as welding or bolting. FIGS. 75-84 show another embodiment of the loading basket 150 that is configured to be used with more reactive spent AGR fuel 152. In this embodiment, the dividers 156 are made of one or more outer layers 164 (e.g., a first outer layer and a second outer layer) of structural material such as stainless steel and an inner layer 166 (a first inner layer or neutron absorbing layer) of boron containing material such as borated aluminum. The structural material provides structural support to the dividers 156 and the borated material provides neutron capture capability for criticality safety. As shown in FIG. 77, the dividers 156 can include one, two, or more total layers of material (structural material plus borated material). Referring to FIG. 76, the box 154 includes channels 168 coupled to the interior walls of the box 154 and sized to receive the edges of the dividers 156. FIG. 79 shows the manner in which the dividers 156 fit into the channels 168 in the box 154. The box 154 also includes collars 170 positioned in the bottom of the box 154. The collars 170 are sized to receive the bottom ends of slotted tubular support members 172 (alternatively referred to as spines) in the manner shown in FIGS. 81-83. The channels 168 and the tubular support members 172 hold the dividers 156 and, consequently, the spent AGR fuel 152 in place in the loading basket 150. This method of securing the dividers 156 in place avoids fastening or welding any of the borated materials in the dividers 156. The spent AGR fuel 152 can be loaded into the Grade A modular container variant 70 using any suitable method and in any suitable environment (wet or dry loading). In one embodiment, the spent AGR fuel 152 is loaded into the loading basket 150, which is then loaded into the Grade A modular container variant 70 as shown in FIG. 84. In another embodiment, the loading basket 150 is first loaded into the Grade A modular container variant 70 and then the spent AGR fuel 152 is loaded into the loading basket 150 as shown in FIG. 85. It should be appreciated that the loading basket 150 and/or the Grade A modular container variant 70 can be loaded with the spent AGR fuel 152 in a pool or out of a pool. Referring back to FIG. 72, one embodiment of a procedure to close and seal the Grade A modular container variant 70 includes remotely placing the lid shielding insert 50 over the cavity 14, remotely placing the lid plate 78 on the main body 30 (sealing member 44 is preinstalled in lid grooves 83) and bolting it in place, draining, drying and inerting the cavity using the ports 82, 84, installing and welding the covers 92 over the ports 82, 84, installing the cover plate 80 over the structural lid 78, and welding the mating surfaces together as shown in FIGS. 16-17. Bare Magnox Fuel Magnox fuel elements have metallic uranium fuel rods in graphite blocks positioned in a magnesium outer casing closed at the ends by magnesium end fittings. The spent uranium fuel rods are currently extracted from the casings and consolidated for reprocessing but this may not continue. The modular container 10, and especially the Grade A modular container variant 70, can be configured to hold the spent bare Magnox fuel. It should be appreciated that other grades of the modular container 10 can also be used if the activity of the spent bare Magnox fuel is low enough. It should also be appreciated that much of the description provided above in connection with the spent AGR fuel applies to the bare Magnox fuel unless noted otherwise. For example, the description of the holes 158, 160 and their function should be considered to apply to both situations. FIG. 86 shows one embodiment of the Grade A modular container variant 70 filled with a loading basket 180 full of spent bare Magnox fuel (not shown). The Grade A modular container variant 70 can be configured to hold any amount of spent bare Magnox fuel. The Grade A modular container variant 70 can include any amount of suitable shielding. In one embodiment, the Grade A modular container variant 70 includes 9-12 inch (228 mm to 305 mm) thick shielding inserts 50, 52, 54, 56. The loading basket 180 is removable from the Grade A modular container variant 70. The loading basket 180 includes a box 184 and dividers 186 (alternatively referred to as divider plates). The box 184 includes lifting members 192 that facilitate handling the loading basket 180. The box 184 can also include holes 158 in the bottom to allow it to drain if it is loaded in a pool. Likewise, the dividers 186 can include recesses, holes or indentations 160 that allow water to flow underneath the divider plates 186 and out the holes 158 in the box 184. FIGS. 87-90 show one embodiment of the loading basket 180 that is configured to be used with spent bare Magnox fuel. The box 184 and the dividers 186 can be made of stainless steel (any of the stainless steels mentioned in this document) and fit together in the manner shown in FIG. 87. The box 184 and dividers 186 can be secured together using any suitable fastening technique such as welding or bolting. The loading basket 180 includes tubes 188 that fit inside the cavities formed by the intersecting dividers 186. The tubes 188 are configured to hold the spent bare Magnox fuel. The bare Magnox fuel can be loaded into the tubes 188 before or after the tubes are put in the loading basket 180. The spent bare Magnox fuel can be loaded into the Grade A modular container variant 70 using any suitable method. In one embodiment, the spent bare Magnox fuel is loaded into the loading basket 180, which is then loaded into the Grade A modular container variant 70 as shown in FIG. 91. In another embodiment, the loading basket 180 is first loaded into the Grade A modular container variant 70 and then the spent bare Magnox fuel is loaded into the loading basket 180 as shown in FIG. 92. Canned Magnox Fuel and Other Canned High Level Waste Canned spent Magnox fuel 200 and other forms of canned HLW can be put in the Grade A modular container variant 70. Canned Magnox fuel 200 is prepared by removing the fuel rods from the casing and enclosing the fuel rods in an overpack can. Other forms of canned waste include vitrified HLW such as non-fuel bearing components, fission product, and actinide waste. Before describing the canned Magnox fuel 200 in greater detail, it should be appreciated that the above descriptions relating to AGR waste and bare Magnox fuel apply to this section unless noted otherwise. The Grade A modular container variant 70 can be configured to hold the canned spent Magnox fuel 200, although other grades of the modular container 10 can also be used if the activity is low enough. FIG. 93 shows one embodiment of the Grade A modular container variant 70 filled with spent canned Magnox fuel 200. The Grade A modular container variant 70 can be configured to hold any amount of canned Magnox fuel 200. The Grade A modular container variant 70 can include any amount of suitable shielding. In one embodiment, the Grade A modular container variant 70 includes 9-12 inch (228 mm to 305 mm) thick shielding inserts 50, 52, 54, 56. The canned Magnox fuel 200 can be positioned in the modular container 70 in a variety of different ways. In one embodiment, intersecting dividers 216 (alternatively referred to as divider plates) are positioned directly in the cavity 14 as shown in FIGS. 94-95. The canned Magnox fuel 200 is placed in the individual cavities formed by the dividers 216 as shown in FIG. 96. In this embodiment, a removable loading basket is not used. In another embodiment, a removable loading basket 210 is used to hold the canned Magnox fuel 200 in the Grade A modular container 70. The loading basket 210 includes a box 214 and intersecting dividers 216 as shown in FIG. 97. The box 214 and dividers 216 can be configured similarly to the boxes described previously (e.g., with channels 168, made of same materials, and the like). The spent canned Magnox fuel 200 can be loaded into the Grade A modular container variant 70 using any suitable method. In one embodiment, the canned Magnox fuel 200 is loaded into the loading basket 210, which is then loaded into the Grade A modular container variant 70 as shown in FIG. 98. In another embodiment, the loading basket 210 is first loaded into the Grade A modular container variant 70 and then the canned Magnox fuel 200 is loaded into the loading basket 210 as shown in FIG. 99. CANDU Fuel CANDU fuel is used in a CANDU (CANada Deuterium Uranium) reactor which is a Canadian invented pressurized, heavy water reactor. CANDU fuel 220 is in the form of bundles grouped together in magazines as shown in FIG. 104. Each magazine includes a 4×3 array of bundles. It should be appreciated that the above descriptions of handling and containerizing other types of spent fuel apply equally to the CANDU fuel 220. FIG. 100 shows one embodiment of the Grade A modular container variant 70 filled with a loading basket 230 full of spent CANDU fuel 220. The Grade A modular container variant 70 can be configured to hold any amount spent CANDU fuel 220. For example, it can be configured to hold approximately 5 to 10 magazines of CANDU fuel 220, which each hold 12 CANDU fuel bundles. The Grade A modular container variant 70 can include any amount of suitable shielding. In one embodiment, the Grade A modular container variant 70 includes 9-12 inch (228 mm to 305 mm) thick shielding inserts 50, 52, 54, 56. The loading basket 230 is removable from the Grade A modular container variant 70. The loading basket 230 includes a box 234 and dividers 236 (alternatively referred to as divider plates). The dividers 236 are positioned parallel to each other in the box 234 and do not intersect. The dividers 236 fit in the channels 168 on the interior walls of the box 234. The spent CANDU fuel 220 can be loaded into the Grade A modular container variant 70 using any suitable method. In one embodiment, the spent CANDU fuel 220 is loaded into the loading basket 230, which is then loaded into the Grade A modular container variant 70 as shown in FIG. 103. In another embodiment, the loading basket 230 is first loaded into the Grade A modular container variant 70 and then the spent CANDU fuel 220 is loaded into the loading basket 230 as shown in FIG. 104. Research Reactor Fuel The modular container 10 can also be used to package spent research reactor fuel. Examples of suitable research reactor fuel include MTR fuel and TRIGA fuel. MTR fuel includes metallic uranium fuel plates clad in aluminum. TRIGA fuel is similar in that it is metallic uranium fuel plates clad in aluminum or stainless steel. It should be appreciated that the above descriptions of handling and containerizing other types of spent fuel apply equally to research reactor fuel unless noted otherwise. FIG. 105 shows one embodiment of the Grade A modular container variant 70 filled with multiple loading baskets 250 full of spent MTR fuel 240. The Grade A modular container variant 70 can be configured to hold any amount spent MTR fuel 240. For example, it can be configured to hold 2-10 loading baskets 250 full of spent MTR fuel 240. The Grade A modular container variant 70 can include any amount of suitable shielding. In one embodiment, the Grade A modular container variant 70 includes 9-12 inch (228 mm to 305 mm) thick shielding inserts 50, 52, 54, 56. The lid shielding insert 50 can be a monolithic slab as shown in FIG. 105 or it can be segmented into different sections as shown in FIG. 112. The lid shielding insert 50 in FIG. 112 is divided into two sections with one section 242 extending around the circumference of the cavity 14 adjacent to the walls 16 and another section 244 covers the spent MTR fuel 240. The loading basket 250 is removable from the Grade A modular container variant 70. The loading basket 250 includes a box 254 and dividers 256 (alternatively referred to as divider plates) as shown in FIG. 106. The dividers 256 intersect each other in the box 254 to form a number of cavities or compartments. The loading basket 250 includes tubes 258 that fit inside the cavities formed by the intersecting dividers 256 as shown in FIG. 107. The tubes 258 are configured to hold the spent MTR fuel 240. The MTR fuel 240 can be loaded into the tubes 258 before or after the tubes are put in the loading basket 250. The spent MTR fuel 240 can be loaded into the Grade A modular container variant 70 using any suitable method. In one embodiment, the spent MTR fuel 240 is loaded into the loading basket 250, which is then loaded into the Grade A modular container variant 70 as shown in FIG. 109. In another embodiment, the loading basket 250 is first loaded into the Grade A modular container variant 70 and then the spent MTR fuel 240 is loaded into the loading basket 250 as shown in FIG. 110. In one embodiment, a separate shielding insert 246 (alternatively referred to as a shield plug) is provided to cap or cover each loading basket 250 as shown in FIG. 111. The use of separate shielding inserts 246 helps reduce exposure to the MTR fuel 240 as it is loaded and/or unloaded from the cavity of the Grade A modular container variant 70. Once the Grade A modular container variant 70 has been loaded, it can be closed in the manner shown in FIG. 112. FIGS. 113-118 show one embodiment for packaging TRIGA fuel 252. This embodiment is similar to the embodiment shown in FIGS. 105-112 for MTR fuel 240 except that the TRIGA fuel 252 uses smaller tubes 260, which means the loading basket 250 can be configured to hold more of the tubes 260. Storage The modular container 10 includes numerous features that make it easy to move and efficiently stack. For example, the modular container 10 includes lifting members 28 that can be remotely engaged by a crane. The modular container 10 also includes openings 34 (alternatively referred to as forklift pockets) that capture the forks of a forklift to reduce the likelihood of the modular container 10 tipping over when it is moved. The lifting members 28 are positioned to facilitate stacking alignment and capture. The fasteners 42 for the structural lid 18 are flush with the structural lid 18 to reduce interference and/or damage that may occur during stacking. The modular container 10 also has a self-supporting configuration. Referring to FIG. 119, the modular container 10 is configured to be stackable. In one embodiment, the modular container 10 is sufficiently strong to allow at least two modular containers 10 to be stacked on top of each other or, desirably, at least three modular containers 10 to be stacked on top of each other or, suitably, at least four modular containers 10 to be stacked on top of each other. The modular container 10 has a comparatively low center of gravity and a large aspect ratio that enhance stability for lateral loadings such as seismic loads. The modular container 10 is also designed to mitigate radiological consequences of an accidental drop or tip over. The modular container 10 can be used for interim storage of radioactive waste at the same site where it is generated or at a distant site. The modular container 10 can be stored outdoors on a concrete pad or indoors in a suitable facility. Referring to FIG. 120, the cuboidal embodiment of the modular container 10 substantially reduces the costs for interim storage compared to conventional cylindrical containers. The cuboidal container provides greater volume efficiency which reduces the total number of containers required by 23% or more for the same waste inventory volume. It also increases the stacking/packing density by up to 27% because the interstitial spaces between the modular containers 10 can be fully utilized. The increased stacking/packing density allows the pad/building footprint for the storage array to be reduced by a factor of two or more. Also, closely packing the modular containers 10 increases self-shielding, which reduces the radiation exposure. The following table shows the interim storage criteria that the modular container 10 is compliant with. TABLE 8Typical Interim Storage Criteria for the Modular ContainerDesignConditionInterim Storage Criteria DescriptionInterim Storage CriteriaDesign LifeMaintain structural, confinement and shielding integrity without25 years minimumdegradationEnvironmentalNormal ambient temperature variations0° F. to 100° F.ConditionsOff-normal ambient temperature variations−40° F. to 125° F.Solar insolation per 10CFR71.71(c)(1)Averaged over 24-hour dayGasMaximum internal pressure for container with filtered ventAtmosphericGenerationMaximum internal pressure for unvented container (Grade B,70 to 175 kPa (10 to 25Grade A)psig)Dead LoadsMaximum container weight35 tonnes (38.6 tons)Live LoadsStacking to maximum of three containers high; weight of two2X maximum containercontainers aboveweight + 15%Wind LoadsPressure due to off-normal design basis wind speed (ASCE 7)150 mphHandlingNot exceed minimum yield strength of container structural3X maximum containerLoadsmaterialweightNot exceed ultimate tensile strength of container structural5X maximum containermaterialweightDrop LoadsSingle accidental drop from height of 15 feet (4.6 meters) onto flatFree drop in the most12 inch (300 mm) thick reinforced concrete slab on compacteddamaging orientationfillFire AccidentHydrocarbon fuel/air fire with average emissivity of 0.95 minutes @ 800° C.(1475° F.)FloodingExternal hydrostatic pressure due to head of water50 feet (15 meters)Kinematic stability for flood water flow (NRC Reg. Guide 1.59)21 feet/sec (6.4 m/s)TornadoTornado wind kinematic stabilityPer NRC Reg. Guide 1.76Tornado missile impact effectsPer NUREG-0800EarthquakeContainer stack to remain kinematically stable when subjected to0.25 g horizontalDBE horizontal ground motion accelerations in both orthogonalVertical 2/3 of horizontaldirections (NRC Reg. Guide 1.60)External DoseExternal surfaces at time of loading2.0 MSv/hr (200 mrem/hr)Rate1 meter from external surfaces post accident conditions10 mSv/hr (1 Rem/hr)SurfaceMaximum removable contamination on container exterior4 Bq/cm2 beta-gammaContaminationsurfaces0.4 Bq/cm2 alpha averagedover 300 cm2Heat OutputMaximum heat load per unit volume of waste2 kW/m3ResidualSolid waste - maximum free liquid by volume1%LiquidsWet waste - maximum free water by volume5%ConfinementLeak tightness (unvented) (Grade B, Grade A)10−4 cm3/sec to 10−6 cm3/sec Transport The modular container 10 can be used to transport radioactive waste in a variety of ways depending on the characteristics of the waste. The modular container 10 allows the determination of the transport packaging to be made at the time of shipment rather than at the time of loading. This simplifies front-end loading and closure operations for placement in interim storage. Future regulatory changes won't impact the already fabricated and loaded containers because they can packaged for transport according to the then applicable regulations. The modular container 10 can be used without modification to transport radioactive waste that meets Type A or IP-2 requirements as shown in FIG. 121. The modular container 10 can be transported immediately after it is loaded or following sufficient decay in interim storage to be within applicable limits. The modular container 10 can include ports for leak testing the enclosure prior to transport. The configuration of the structural lid 18 allows the one or more sealing members 44 to be replaced prior to transport if needed. In general, Type A packaging is used to transport radioactive material with higher concentrations of radioactivity than those shipped in industrial packaging. Type A packaging is typically constructed of steel, wood, or fiberboard, and has an inner containment vessel made of glass, plastic, or metal surrounded with packing material made of polyethylene, rubber, or vermiculite. The modular container 10 in all of its grades satisfies Type A packaging requirements, which are out lined in 49 CFR 173.412. As a result, it also meets the requirements of 49 CFR 173.411 for an IP-2 packaging. Examples of material typically shipped in Type A Packages include nuclear medicines (radiopharmaceuticals), radioactive waste, and radioactive sources used in industrial applications. Type A packaging and its radioactive contents must meet standard testing requirements designed to ensure that the package retains its containment integrity and shielding under normal transport conditions. Type A packaging should withstand moderate amounts of heat, cold, reduced air pressure, vibration, impact, water spray, drop, penetration, and stacking tests. Type A packagings are not, however, designed to withstand the forces of an accident such as those defined for a more robust Type B transport packaging. The consequences of a release of the material in one of the Type A packages would not be significant since the quantity of material in this package is limited. Type A packaging is only used to transport amounts of radioactive material that are not life threatening or life-endangering. In general, Type B packaging is used to transport material with the highest levels of radioactivity. Examples of material transported in Type B packaging include SNF, HLW, and high concentrations of other radioactive material such as cesium and cobalt. These package designs must withstand all Type A tests and a series of tests that simulate severe or worst-case accident conditions. Accident conditions are simulated by performance testing and engineering analysis. Life-endangering amounts of radioactive material are required to be transported in Type B Packages. Requirements for Type B packaging can be found in 49 CFR 173.411, 49 CFR 73.413, and 10 CFR 71 (which also comply with IAEA T-SR-1). The modular container 10 can be placed in a purpose built transport overpack 280 that is configured to accommodate one modular shielding container and to comply with Type B packaging requirements for waste that requires it. FIG. 121 shows the modular container 10 positioned in the transport overpack 280. This configuration has a number of advantages. One is that the demanding 10 CFR 71 and T-SR-1 requirements are satisfied primarily by the transport overpack 280 thereby separating the transport licensing from the design, production and deployment of the modular container 10. This makes it possible to change the design of the modular container 10 without requiring amendments to the transport license certificate of compliance. Also, container fabrication deviations can be dispositioned without requiring amendments to the transport license. The modular container 10 in its various configurations can be included in the Type B overpack license as contents specifications. This facilitates transport license renewal every five years. The transport overpack 280 includes a transport container 282 shown in FIGS. 122-123 and top and bottom impact limiters 298. The transport container 282 is configured to receive and enclose the sealed and closed modular container 10. The transport container 282 includes a main body 284 that holds that the modular container 10 and a lid 286 that closes the transport container 282. The main body 284 and the lid 286 form an enclosure envelope 288. The main body 284 includes a base plate 290 and walls 292 that extend upward from the base plate 290. Lifting members 294 are coupled to the outside of the walls 292 to make it easy to handle and move the transport container 282. The lid 286 is coupled to a flange 296 that extends around the top edge of the main body 284 in a manner similar to how the structural lid 18 is coupled to the main body 30 (e.g., bolts and nuts, bolts in threaded holes, and so forth). Likewise, the lid 286 and the main body 284 can fit together and sealed in any of the ways disclosed in connection with the modular container 10. The impact limiters 298 can be made of any of the same materials described above in connection with the impact limiters that can be added to the modular container 10. The impact limiters 298 are coupled together using heavy duty turnbuckles 300 that extend between the top and bottom impact limiters 298. The transport overpack 280 can be configured to maintain containment function during transport. It does this by satisfying the applicable requirements, for example, a 10 m drop onto an unyielding surface as well as puncture, fire, and deep immersion requirements. The modular container 10 is designed to maintain shielding during transport. The transport overpack 280 can be sealed and leak tested at the time of shipment. In one embodiment, the transport overpack provides leak tightness to 10−6 cm3/sec via pressure drop test on interspace between the sealing members (e.g., O-rings made of butyl rubber) at a maximum normal operating pressure of approximately 50 psig. The transport overpack 280 reduces waste packaging and transport costs. The transport overpack 280 can be reused and not every modular container 10 requires it. It is only necessary to maintain a small number of transport overpacks 280 at any given time. There is no need to incorporate costly Type B packaging on every modular container 10 because the entire waste stream may not be Type B or the modular container 10 may have decayed below Type B levels. The decision to use the transport overpack 280 can be made on a case-by-case basis at the time of shipment rather than at the time of loading the modular container. The two-part lid for the modular container enables in-situ replacement of the one or more sealing members 44 prior to transport if it is needed to restore Type A capability following extended storage. Indeterminate or degraded modular containers 10 can be transported as Type B using the transport overpack 280. Future regulatory changes won't impact the modular containers 10 that are fabricated and loaded now. The modular container 10 and transport overpack 280 are designed to meet legal weight and size requirements for truck shipment in all configurations. The heaviest Type A configuration (i.e., the modular container 10 alone) is 31.5 tonnes and can be moved by a truck with six axles. The heaviest Type B configuration (i.e., the modular container 10 and transport overpack 280) is 41.6 tonnes and can be moved by a truck with seven axles. The modular container 10 and/or transport overpack 280 can be secured directly to the vehicle without the need for a shipping skid as shown in FIG. 121. Also, there is generally no need for oversize or overweight permits. The modular container 10 and/or transport overpack 280 complies with all of the Type A and Type B transport requirements shown in the following table. TABLE 9Type A and Type B Transport Requirements the Packaging System Complies WithType A TransportType B TransportDesignCriteria (49 CFR 178.350Criteria (10 CFR 71ConditionTransport Criteria Descriptionand IAEA TS-R-1)and IAEA TS-R-1)Maximum SizeFor highway to avoid oversize permittingContainer aloneContainer + overpackFor secondary railway tunnel2.65 m W × 2.40 m H2.65 m W × 2.55 m HMaximumRoad to avoid overweight permitting35 tonnes (38.6 tons)35 tonnes (38.6 tons)WeightRail for secondary railways70 tonnes (77.2 tons)70 tonnes (77.2 tons)TemperatureMinimum to maximum ambient−29° C. to 38° C. normal−40° C. to 70° C. accidenttemperature range50° C. nonexclusive85° C. exclusiveMaximum container surface temperatureHandling≦ yield strength of container structural3X max. container weight3X max. containerLoadsmaterial5X max. container weightweight≦ tensile strength of container structural5X max. containermaterialweightVibrationAccelerations in three orthogonalPer 49 CFR 178.608Per NUREG/CR-0128directions simultaneouslyTie-downTie-down attachments that are integral to0.6 g vert., 0.3 g axial, 0.2 g2 g vertical, 10 g axial,containertrans.5 g trans.Dose RateMaximum container external surfacesType A quantities ≦ A22.0 mSv/hr (200 mrem/hr)2 meters from transport vehicleLSA and SCO with0.1 mSv/hr (10 mrem/hr)unshielded dose rate ≦1 R/hr at 3 mSurfaceMaximum removable contamination on49 CFR 173.403 limits for4 Bq/cm2 beta-gammaContaminationcontainer exterior surfacesSCOs0.4 Bq/cm2 alpha ave.49 CFR 173.443over 300 cm2PressureMaximum internal pressure for containerAtmosphericAtmosphericwith filtered ventMNOP 70 kPa (10 psig)MNOP 700 kPa (100Maximum internal pressure for unventedpsig)containerFree DropSingle accidental drop from the specified1.2 m (4 foot) <11,000 kg9.0 m (30 ft)height onto flat unyielding surface0.3 m (1 foot) >15,000 kgCrushDrop of mass onto containerNone500 kg (1,100 lbs) steelplate from 9.0 m (30 ft)Penetration/Drop of cylindrical bar onto weakest part3.2 cm (1.25″) Ø15 cm (6″) ØPunctureof container (Type A) or drop ofMass of 6 kg (13.2 lbs)Mild steelcontainer on puncture pin (Type B)Drop height 1 m (3.3 ft)Drop height 1 m (3.3 ft)StackingContainer stack to remain kinematically5X mass of package orNonestable when subjected to stacking loads13 kPa (1.9 psi)Fire AccidentHydrocarbon fuel/air fire with average10 minutes @ 800° C.30 minutes @ 800° C.emissivity of 0.9(1475° F.)(1475° F.)ImmersionImmersed under 15 m (50 ft) head ofWater spray per 49 CFR150 kPa (21.8 psig)water for 8 hrs (Type B)173.465ContainmentLeak tightness (unvented)10−4 cm3/sec10−6 cm3/sec Disposal The modular container 10 is configured to be ready for disposal. Disposal of radioactive waste takes primarily two forms: (1) shallow land burial (see 10 CFR 61 or similar) and (2) deep geological repository (DGR) (per 10 CFR 63 or similar). Shallow land burial is suitable primarily for Class B and C low level waste and/or waste with short lived isotopes. The radioactive waste can be disposed of by shallow land burial in one of the following ways. First, the sub-container(s) 126 and/or the support framework 128 can be removed from the modular container 10 and buried. Second, the sub-container(s) 126, the support framework 128, and/or the liner 124 can be placed in a vault 302 (e.g., concrete vault) and buried as shown in FIGS. 127-128. The vault 302 can include a main body 304 and a lid 306 as well as lifting members 308. Third, the modular container 10 is buried directly with or without modification, which is more prevalent for the Grade C modular container variant 120. The first two options allow the modular container 10 to be reused. It should be appreciated that the modular container 10 or the packaged waste container therein can have any number of configurations for disposal by shallow land burial. Disposal in a deep geological repository usually requires disposal of the modular container 10. This type of disposal is suitable for spent nuclear fuel (SNF), greater than class C (GTCC) waste, intermediate level waste (ILW), high level waste (HLW), remote handled transuranic waste (RH-TRUW), and/or waste with long-lived isotopes. The modular container 10 alone or combined with an engineered barrier overpack satisfies the requirements for deep geological repository burial as set forth in the following table. Repository emplacement of the modular container 10 alone typically occurs for ILW, GTCC, and RH-TRUW packaged in the Grade B modular container variant 10. Repository emplacement of the modular container 10 in an engineered barrier overpack typically occurs for SNF and HLW packaged in the Grade A modular container variant 70. The engineered barrier overpack can be made of long term performance materials such as copper or Alloy-22 to provide a second confinement barrier to that of the modular container to mitigate the dispersion of radionuclides over the very long term. It should be appreciated that the modular container 10 or the packaged waste contained therein can have any number of configurations for disposal by repository emplacement. TABLE 10Typical Deep Geological Repository (DGR) Container RequirementsDesignConditionTypical DGR Criteria DescriptionTypical DGR CriteriaDesign LifeMaintain structural, confinement and shielding integrity50 years minimum (DGR prewithout degradationclosure period)EnvironmentalAmbient temperature variations (during DGR pre closure)15° C. to 25° C.ConditionsApproximate relative humidity (during DGR pre closure)75%Maximum SizeMaximum container length, width and height2.65 m W × 5.2 m L × 2.55 m HMaximumMaximum container weight35 tonnes (38.6 tons)WeightHandlingNot exceed minimum yield strength of container structural3X maximum container weightLoadsmaterial5X maximum container weightNot exceed ultimate tensile strength of container structuralmaterialDrift SizeContainer stacks fit within finished dimensions of7.4 m wide (min.) × 6.3 m highemplacement rooms(min.)Minimum clearances from walls, between containers and from300 mm, 50 mm, and 1.2 mceilingStackabilityStack to maximum height in stable, self supporting2X maximum container weightconfiguration up to 6 m+15%External DoseExternal surfaces at time of emplacement2.0 mSv/hr (200 mrem/hr)Rate1 meter from external surfaces post accident conditions10 mSv/hr (1 Rem/hr)SurfaceMaximum removable contamination on container exterior4 Bq/cm2 beta-gammaContaminationsurfaces0.4 Bq/cm2 alpha averaged over300 cm2Heat OutputMaximum heat load per unit volume of waste - no restrictions0.1 W/m3Maximum heat load per unit volume of waste - with10 W/m3restrictionsResidualSolid waste - maximum free liquid by volume1%LiquidsResin waste - maximum free water by volume5%Gas GenerationMaximum internal pressure for container with filtered ventAtmosphericMaximum internal pressure for unvented container (Grade B,70 to 175 kPa (10 to 25 psig)Grade A)Impact LoadsSingle accidental drop from the highest handling height (≦Free drop in the most damaging4.6 m (15 ft.)) onto flat 300 mm (12 inch) thick reinforcedorientationconcrete floor on compacted fillSeismic EventContainer stack to remain kinematically stable when subjected0.25 g horizontalto horizontal ground motion acceleration in both orthogonalVertical 2/3 of horizontaldirectionsFire AccidentHydrocarbon fuel/air fire with average emissivity of 0.9 (limit10 minutes @ 800° C. (1475° F.)combustibles)ImmersionImmersed under 15 meter head of water for 8 hours (during150 kPa (21.8 psig)DGR pre closure)ConfinementLeak tightness (unvented) (Grade B, Grade A)10−4 cm3/sec to 10−6 cm3/sec In one embodiment, the materials used to make the modular container 10 can be tailored for disposal conditions. For example, the Grade C modular container variant 120 can be made of coated carbon steel that is adequate for near surface land burial. The Grade A and B modular containers variants 10, 70 can be made of stainless steel that is suitable for DGR disposal. The modular container 10 can be a component of the DGR's engineered barrier system. The robust enclosure envelope of the modular container 10 provides integrity through DGR pre-closure. The Grade A and B modular containers 10, 70 can provide a stable enclosure envelope for at least 25 years of interim storage followed by at least 50 years of pre-closure periods for the DGR. The stainless steel enclosure envelope extends the performance of the modular container. Also, there are no coatings that can degrade. The configuration of the structural lid 18 enables the structural lid 18 and/or sealing members 44 to be repaired or replaced prior to or following placement of the modular container 10 at the disposal site. The composition of the shielding inserts 50, 52, 54, 56 in the modular container 10 can be adjusted according to the requirements of the disposal site. For example, it may be desirable to reduce the total volume of metal in the DGR to reduce post closure gas generation due to long term corrosion of metals. This could be an important factor for DGRs with low natural pH and low rock permeability. One way to reduce the volume of metal in the DGR is to use recycled waste metals as the shielding inserts 50, 52, 54, 56. This reduces the amount of metal because it avoids the need to package and dispose of such recycled waste metals in additional, separate containers. Another way to reduce the volume of metal in the DGR is to make the shielding inserts 50, 52, 54, 56 out of high density concrete. The enclosure envelope 12 can be made of low carbon stainless steel of various compositions to slow the corrosion rate. The modular container 10 can also be placed in an engineered barrier overpack made of long term performance material. The cuboidal shaped modular container 10 reduces the required disposal space by a factor of two or more compared to conventional cylindrical containers. The cuboidal shaped modular container 10 has greater volume efficiency, which reduces the total number of containers by 23% or more for the same waste volume. Also, the stacking/packing density of the cuboidal shaped modular containers 10 is increased by 27% relative to conventional cylindrical containers because the interstitial spaces between the modular containers 10 are fully utilized. Referring to FIGS. 129-130, the modular container 70 with HLW and/or SNF can be placed in a disposal overpack 310 for repository disposal. The disposal overpack 310 can be an integrated component of the engineered barrier system. The modular container 10 can be positioned in the deep geological repository in a variety of different ways. In one embodiment, the modular containers 10 are positioned horizontally and can be stacked or unstacked by itself or in an engineered barrier overpack. In another embodiment, the modular containers 10 are positioned in vertical bore holes and can be stacked or unstacked by itself or in an engineered barrier overpack. The bore hole voids can be filled with bentonite plugs and rings. FIG. 131 shows the typical uses and disposal options for the different grades of the modular container 10. Container Configurations Moreover, the following tables show exemplary configurations of the modular container 10 and various storage, transport, and disposal options. Such configurations can be pre-qualified and cataloged for a range of applications to simplify adaptation and deployment by the end-user. It should be appreciated that the modular container 10 can have any number of configurations and applications. TABLE 11Exemplary Modular Container ConfigurationsModularShielding InsertCavity FeaturesWasteWasteContainerThicknessSupportShieldFormTypeGradeNone4 in6 in9 in12 inLinerFrameworkPlatesBinsBasketTubesSolid wasteActivatedB, C●●●●hardwareContaminatedB, C●●●hardwareGranularB, C●●●●materialGTCCA●●●●●Stabilized55 gal drumsB, C●●●●●wet waste85 gal drumsB, C●●●●●500 L drumsB, C●●●●●Cylindrical HICA, B, C●●●●●Cuboidal HICA, B, C●●●Bulk liquidA, B, C●●●●●transferSpentAGR FuelA●●nuclear fuelMagnox fuel -A●●●and HLWbare(alsoMagnox fuel -Stretched A●●incorporatescannedexternalUK HLWStretched A●●neutroncannistersshieldingMTR fuelA●●●panels)TRIGA fuelA●●●CANDU fuelA●●DepartmentSealed sourcesA, B●●●of EnergyIsotope Rx fuelA●●●●wasteProduction RxA●●●●fuelU.S. HLWStretched A●●cannistersRH-TRUA●●●CH-TRUB●●●● TABLE 12Modular Container Loading, Storage, Transport, and Disposal ConfigurationsOperational Mode ConfigurationWasteWasteWet/DryLoadingLoadingStorage onStorage inTransportDisposal byRepositoryFormTypeLoadingOption 1Option 2PadBuildingTypeBurialDisposalSolid wasteActivatedWet or DryICICICICA or BICIChardwareContaminatedDryICICICICA or IP2ICN/AhardwareGranular materialDryILICICICA or IP2ILVICGTCCWet or DryICIBICICBN/AICStabilized55 gal drumsN/AIDSFICICIDSFA or IP2IDSFICwet waste85 gal drumsN/AIDSFICICIDSFA or IP2IDSFIC500 L drumsN/AIDSFICICIDSFA or BIDSFICCylindrical HICN/AIHSFICICIHSFA or BIVICCuboidal HICN/AIHICICICA or BIVICBulk liquidN/AILICICILSFA or BN/AN/AtransferSpentAGR FuelWet or DryIBICICICBN/AICnuclear fuelMagnox fuel -Wet or DryIBTICICICBN/AICand HLWbareMagnox fuel -DryIBTICICICBN/AICcannedUK HLWDryIBICICICBN/AICcanistersMTR fuelWet or DryIBTICICICBN/AICTRIGA fuelWet or DryIBTICICICBN/AICCANDU fuelWet or DryIBICICICBN/AICDepartmentSealed sourcesDryIBTICICICBN/AICof EnergyIsotope Rx fuelWet or DryIBTICICICBN/AICwasteProduction RxWet or DryIBTICICICBN/AICfuelU.S. HLWDryIBICICICBN/AICcanistersRH-TRUDryICICICICBN/AICCH-TRUDryIDSFICICIDSFA or BN/AIDSFIC = in containerIL = in linerIV = in vaultIH = in HICIB = in basketIBT = in basket/tubesIDSF = in drum/support frameworkIHSF = in HIC/support frameworkILSF = in liner/support frameworkILV = in liner/vault It should be appreciated that some components, features, and/or configurations may be described in connection with only one particular embodiment, but these same components, features, and/or configurations can be applied or used with many other embodiments and should be considered applicable to the other embodiments, unless stated otherwise or unless such a component, feature, and/or configuration is technically impossible to use with the other embodiment. Thus, the components, features, and/or configurations of the various embodiments can be combined together in any manner and such combinations are expressly contemplated and disclosed by this statement. The terms recited in the claims should be given their ordinary and customary meaning as determined by reference to relevant entries in widely used general dictionaries and/or relevant technical dictionaries, commonly understood meanings by those in the art, etc., with the understanding that the broadest meaning imparted by any one or combination of these sources should be given to the claim terms (e.g., two or more relevant dictionary entries should be combined to provide the broadest meaning of the combination of entries, etc.) subject only to the following exceptions: (a) if a term is used in a manner that is more expansive than its ordinary and customary meaning, the term should be given its ordinary and customary meaning plus the additional expansive meaning, or (b) if a term has been explicitly defined to have a different meaning by reciting the term followed by the phrase “as used herein shall mean” or similar language (e.g., “herein this term means,” “as defined herein,” “for the purposes of this disclosure the term shall mean,” etc.). References to specific examples, use of “i.e.,” use of the word “invention,” etc., are not meant to invoke exception (b) or otherwise restrict the scope of the recited claim terms. Other than situations where exception (b) applies, nothing contained herein should be considered a disclaimer or disavowal of claim scope. The subject matter recited in the claims is not coextensive with and should not be interpreted to be coextensive with any particular embodiment, feature, or combination of features shown herein. This is true even if only a single embodiment of the particular feature or combination of features is illustrated and described herein. Thus, the appended claims should be given their broadest interpretation in view of the prior art and the meaning of the claim terms. As used herein, spatial or directional terms, such as “left,” “right,” “front,” “back,” and the like, relate to the subject matter as it is shown in the drawings. However, it is to be understood that the described subject matter may assume various alternative orientations and, accordingly, such terms are not to be considered as limiting. Articles such as “the,” “a,” and “an” can connote the singular or plural. Also, the word “or” when used without a preceding “either” (or other similar language indicating that “or” is unequivocally meant to be exclusive—e.g., only one of x or y, etc.) shall be interpreted to be inclusive (e.g., “x or y” means one or both x or y). The term “and/or” shall also be interpreted to be inclusive (e.g., “x and/or y” means one or both x or y). In situations where “and/or” or “or” are used as a conjunction for a group of three or more items, the group should be interpreted to include one item alone, all of the items together, or any combination or number of the items. Moreover, terms used in the specification and claims such as have, having, include, and including should be construed to be synonymous with the terms comprise and comprising. Unless otherwise indicated, all numbers or expressions, such as those expressing dimensions, physical characteristics, etc. used in the specification (other than the claims) are understood as modified in all instances by the term “approximately.” At the very least, and not as an attempt to limit the application of the doctrine of equivalents to the claims, each numerical parameter recited in the specification or claims which is modified by the term “approximately” should at least be construed in light of the number of recited significant digits and by applying ordinary rounding techniques. All disclosed ranges are to be understood to encompass and provide support for claims that recite any and all subranges or any and all individual values subsumed therein. For example, a stated range of 1 to 10 should be considered to include and provide support for claims that recite any and all subranges or individual values that are between and/or inclusive of the minimum value of 1 and the maximum value of 10; that is, all subranges beginning with a minimum value of 1 or more and ending with a maximum value of 10 or less (e.g., 5.5 to 10, 2.34 to 3.56, and so forth) or any values from 1 to 10 (e.g., 3, 5.8, 9.9994, and so forth). All disclosed numerical values are to be understood as being variable from 0-100% in either direction and thus provide support for claims that recite such values or any and all ranges or subranges that can be formed by such values. For example, a stated numerical value of 8 should be understood to vary from 0 to 16 (100% in either direction) and provide support for claims that recite the range itself (e.g., 0 to 16), any subrange within the range (e.g., 2 to 12.5) or any individual value within that range (e.g., 15.2).
	abstract	Methods, apparatus and systems for controlling x-ray exposure during a scan with a computed tomography system are provided. The system includes an x-ray source, a collimator, and a detector array. The collimator includes a first cam configured to shutter an x-ray fan beam, generated by the x-ray source, in a first direction, a second cam located on an opposite side of a focal point of the x-ray source from the first cam, the second cam configured to shutter the x-ray fan beam in a second direction, the second direction being opposite from the first direction, and a cam drive configured to position at least one of the first cam and the second cam during the scan.
053496140	abstract	A tool for remotely installing a plug in a steam line of a reactor vessel. The tool has a strongback assembly on which a plug is securely mounted, a collapsible structure (e.g., a scissors jack assembly) which supports the strongback assembly and is movable between an extended state and a collapsed state, a hanging bracket assembly for hanging the collapsible structure inside the reactor vessel, and a rotatable actuating screw coupled to the collapsible structure. The collapsible structure moves between the extended and collapsed states in response to rotation of the actuating screw. During this operation, the plug is carried from a retracted position outside and aligned with the steam line nozzle to an inserted position inside the nozzle.
	description	This is a continuation of application Ser. No. 14/495,137, filed Sep. 24, 2014, which claims the benefit of priority to GB 1317016.2, filed Sep. 25, 2013 and GB 1411005.0, filed Jun. 20, 2014, all of which are expressly incorporated herein by reference in their entirety. The present invention relates to a collimator, particularly but not exclusively to a collimator for radiotherapy apparatus. Radiotherapy is a form of treatment for tumours and other lesions which involves directing a beam of ionising radiation toward the lesion. The radiation harms the tumour tissue and causes its reduction or elimination. However, the radiation is also harmful to healthy tissue around the lesion; although healthy tissue is slightly less susceptible to the effects of ionising radiation, measures are taken to limit the exposure of healthy tissue to the extent possible. One such measure is to direct the beam toward the lesion from a number of radial directions by mounting the radiation source in a treatment head which is movable with respect to the lesion, such as by being mounted on a rotatable gantry. Thus, the lesion (or part of it) remains in the beam at substantially all times whereas each individual section of healthy tissue around the beam is only exposed to the beam briefly. In this way, the dose delivered to the lesion can be a multiple of the dose delivered to healthy tissue remote from the lesion. Another measure is to collimate the beam so as to limit its lateral extent and avoid the unnecessary irradiation of healthy tissue. Modern collimators for radiotherapy devices are known as “multi-leaf collimators” and comprise an array of adjacent tungsten leaves, each of which is narrow so as to provide a high resolution but deep (in the direction of the beam) so as to provide an effective attenuation of the beam. Each leaf is moveable into and out of the beam, largely independently of those around it, so that the tips of the individual leaves can define a variable shape as required. Two such “banks” of leaves will usually be provided, on opposing sides of the beam aperture, thereby allowing a field to be defined within that aperture, or substantially any shape. An example of a multi-leaf collimator (MLC) is disclosed in our earlier application published as EP-A-0314214. There are however limits to the attenuation that can be provided by a multi-leaf collimator. In particular, rules govern the minimum distance between opposing leaves so as to prevent the leaves from jamming or being damaged. Further, whilst it may be permissible for one leaf to be extended so that its tip touches or very closely approaches the tip of the exactly opposite leaf, those tips are usually round so as to provide a small penumbra at the patient, and therefore there will be leakage from the gap between them. For these reasons, there is usually a “block collimator” in series with the multi-leaf collimator, in the form of a substantial block of tungsten that can be extended or retracted in a direction transverse to the movement direction of the leaves. Thus, it can cover a region outside the defined field where the entire width of the aperture needs to be covered. Typically, there will be a pair of block collimators, one either side of the beam, the or each block collimator being substantially square or rectangular, as seen in the direction of the radiation beam to be collimated. The block does impose a substantial weight penalty. The collimators are usually accommodated in the radiation source, which is to be rotated around the patient in order to allow the beam to be directed toward the lesion from a variety of radial directions. Thus, a reduction in this weight would be beneficial. Our earlier application EP2153448A1 described one such way of doing so. The collimator blocks are required to be of the order of 8 cm thick solid tungsten material. This imposes a significant weight burden. Correspondingly, the mechanism required to move a significantly greater mass of collimator block will be correspondingly heavier itself. Both of these increase the overall mass of the treatment head, which in turn causes the structure of the radiotherapy apparatus to deflect more, resulting in further complications for the compensating control systems. It should be borne in mind that most clinical accelerations place the treatment head at the end of long arm which is mounted on a rotatable support so that the treatment head can be rotated around the patient. Additional mass at the end of that arm causes the arm to deform in a direction which will vary (relative to the treatment head) as the treatment head traverses in an arc around the patient. The present invention therefore seeks to provide an arrangement which is able to offer the necessary blocking of the radiation beam, whilst reducing mass over conventional arrangements. The present invention provides a collimator for a radiotherapy apparatus, comprising a block of radiation-attenuating material for moving into and out of a beam of therapeutic radiation having a depth and a front face forming the leading edge of the block when, in use, it is moved into the beam and one, two or more main rear faces opposite the front face which main rear face(s) together substantially define(s) the trailing edge of the block when. In use, it is moved into the beam, the or each main rear face being substantially planar in the direction of the depth of the block and non-parallel to the front face. Such an arrangement enables the size and therefore weight of the collimator to be reduced compared to conventional block collimators, as will be further described below. There may be a single rear face, or there may be two rear faces which together form a concave V-shape in the block opposite the front face. The or each rear face is substantially planar and non-parallel to the front face. These arrangements are both simple and enable significant savings in block material and hence weight to be made; single rear face, or “wedge” embodiments, enable greater weight savings than similar implementations having two rear faces (or concave V-shaped embodiment), as will be explained below. Embodiments having three or more rear faces in a concave shape are technically feasible, but these are more complex to manufacture and provide little additional benefit in weight saving. The angle between the or each rear face and the front face is suitably determined so as to match the trajectory the leaves of an MLC take move between extremes of positioning with the trajectory of the collimator block, which is dependent on the speeds of movement of the MLC leaves and the collimator block; accordingly, this angle may be between 10 and 80 degrees, and may be between 30 and 60 degrees. There may be two side faces leading from the front face to the rear face (where there is only one rear face); this provides a safety margin when movement of the block collimator and MLC leaves is initiated. The side faces may be substantially parallel. The collimator may have a top and bottom face (as seen in the direction of the radiation beam which is collimated), and these faces may be planar; to the extent that they are planar these faces may be substantially parallel. Additionally or alternatively, these faces may be shaped as described in our EP2153448A1, so as to provide additional weight saving; in particular the edge of the collimator block at its front face (the front edge) may be of greater thickness than at least one region behind the front edge (i.e. towards the rear face(s), between the leading and trailing edges of the block. It is envisaged that the collimator blocks may be mountable in a radiotherapy apparatus so that they may be moved back and forth in a direction transverse to the front face. The collimator blocks may be shaped and configured so as to be moveable through an arc centred on the nominal point source of the radiation beam, as is known in the art. In other aspects, the invention also provides radiotherapy apparatus including such collimators, and methods of operating such radiotherapy apparatus. The present invention is predicated upon the movement of the leaves of an MLC. In one particular arrangement, where the MLC leaves are capable of travelling across the entire width of the aperture formed by the primary collimator edge, we have recognised that it is not necessary to have a collimator block which extends across the aperture for the entirety of its length (i.e. in the direction parallel to the direction of movement of the collimator block—“the movement direction”), because the main source of radiation leakage (the attenuation of which is a main objective of the collimator block) is not between the sides of the leaves but rather through the “gap” between opposing leaf tips. When the leaf tips are outside of the beam of radiation (i.e. extended fully across the aperture, so that the “gap” between the leaf tips is within the penumbra of the primary collimator edge and hence shielded from the radiation source) there is no radiation leakage as such, and the collimator block is not required to attenuate any radiation leakage. Accordingly, the collimator block need only be deep enough in the movement direction to cover the leaf tips while they move across the beam towards the primary collimator edge. Assuming that the leaves move from the centre of the aperture, between the two primary collimator edges, a concave V-shape can be provided (or cut) into the rear edge of the collimator block in such a way that the edge of this V-shape matches the trajectory the MLC leaves will take as the MLC leaves move and as the collimator block moves transversely thereto, and taking account of the speeds of movement of the MLC leaves and of the collimator block—which may be their maximum speeds, and/or may take account of their acceleration and/or deceleration. Depending on which side of the V the MLC leaves were last used, and on which side they are next to be used, a control system can determine which primary collimator edge the leaves next travel to. FIG. 1 is an isometric schematic view of one embodiment of a collimator block 2 which has a leading or front face 4, a main rear face 6 in the form of a V-shaped concave cut out formed by two substantially planar faces (only one, 6a being visible in the drawing). The collimator has two substantially planar side faces (only one 8a being visible in the drawing), and substantially planar faces to the top 10 and bottom (not visible). The collimator is moved in use in the direction of the arrows M. The distance S in the direction between the front face 4 and the apex of the V-shape provides a safety margin, as will be described below. Note that FIG. 1 shows two small rear surfaces either side of the V-shaped cut-out; such an arrangement, provided the width (parallel to the front face 4 and transverse to the M axis) of these surfaces forms only a minor proportion of the entire width of the block 2, is within the scope of the claims, and the word “main” should be interpreted accordingly. FIG. 2 is a side view, showing the collimator block 2 beneath the opposed leaves 14a, 14b of a multi-leaf collimator, with a gap 16 between the tips of the opposed leaves; with reference to direction M in FIG. 1, in FIG. 2 the direction M is perpendicular to the plane of the drawing. The MLC leaves are movable to left or right in FIG. 2. Radiation beam B (produced by a linear accelerator, for example) is shaped by a primary collimator (not shown) so as to have the outline shown by the dotted lines. As shown in FIG. 2, the MLC leaves are fully withdrawn to one side of the radiation beam B, so that the gap is outside the beam B and therefore the MLC and collimator 2 in combination provide full shielding of the radiation. Referring now to FIG. 3, this top view shows successive positions of the MLC leaves 14a′, 14a″, 14a′″, 14b′, 14b″, 14b′″ as the MLC leaves move from the centre line 20 towards one side of the radiation aperture 20 created by the primary collimator (the aperture being illustrated between areas 18a, 18b, in which areas the radiation is fully shielded by the primary collimator (not shown)). As the MLC leaves move, so does the gap 16 between their tips; combining this movement with simultaneous movement of the collimator block 2, line 22a shows as a single vector line the trajectory of the gap 16 relative to the rear face 6 of the collimator block 2, where the angle of the rear face 6 is sufficient to ensure full shielding (i.e. to ensure that the collimator block 2 shields radiation which might pass through the gap 16). Line 22a is drawn in the case where the speed of movement of the MLC leaves 14a, 14b is approximately the same as that of the collimator block 2, so that line 22a is at an angle of at about 45 degrees to centre line 20 (which corresponds to the angle of the rear face 6 to the direction M). Line 22b illustrates the case where the MLC speed is somewhat slower, so that the angle to the centre line 20, and thus the angle of the rear face 6 to direction M, is decreased. Those skilled in the art will readily understand the geometries suitable for different types of MLC and collimator block movement apparatus. In most commercially available MLC/collimator arrangements, angles between 10 and 80 degrees would be feasible, and angles between 30 and 60 degrees represent a good compromise between the movement capabilities of the apparatus and enabling a significant reduction in weight of the collimator block. It will be appreciated that the above arrangements assume a constant speed of movement of the MLC leaves and of the collimator block. Of course, in practice these elements are normally made of a dense material such as tungsten, which have considerable inertia, and therefore in practice the apparatus must accommodate the necessary acceleration and deceleration of the elements. This could be provided by a suitably programmed controller to control movement of the MLC leaves and/or collimator block appropriately, however the simple approach of providing a length S of collimator block 2 between the leading, front face 4 and the rear face 6 provides a suitable safety margin, thus ensuring that inertial effects do not allow the gap 16 to be unshielded whilst it moves towards a “parked” position, behind the penumbra 18b of the primary collimator. The same applies in the case where the MLC leaves are moving in the opposite direction, towards the penumbra 18a. Referring now to FIG. 4, FIG. 4a shows in plan view the collimator block 2 of FIGS. 1 to 3 (but without two small rear surfaces either side of the V-shaped cut-out in FIG. 1); as explained above, this V-shaped collimator block corresponds to the case where the MLC leaves move from the centre line 20 of the radiation beam to one side of the aperture 20. It is also possible for the MLC leaves to move from one side of the aperture 20 to the other. In this case it will be appreciated that the shape of the collimator block 2″ can be simpler, namely a “wedge” as shown in FIG. 4b. The collimator block 2″ of FIG. 4b is provided with the same safety margin S as in the previous embodiment, however the movement speeds are adjusted to provide the same overall length of collimator block 2″ (in the M direction) as was the case with the previous embodiment—meaning that there is a significant saving of material, and hence weight, compared to the first embodiment, as indicated by the shaded area A in FIG. 4b. It will of course be understood that many variations may be made to the above-described embodiment without departing from the scope of the present invention. For example, although shown as symmetrical in FIG. 4a, the two main rear faces could be of different lengths so as to accommodate different speeds of movement of MLC leaves in different directions, and the apex of the V-shape could be central, as shown, or it could be offset to one or other side. Where there are more than two main rear faces, these before define a concave depression in the rear surface of the block, these rear faces may be arranged symmetrically or asymmetrically. As previously mentioned, it may be advantageous in some applications for the depth of the collimator block (i.e. into the plane of FIG. 4) to vary, such as by making the leading edge (adjacent the front face 4) thicker, or by profiling the block between its leading and trailing edges as described in EP2153448A1. Additionally or alternatively, the block may have a web of material, of lesser thickness than the remainder of the block, which is disposed against the rear face(s) so as to fit into the V-shaped space in the first embodiment or to render its appearance in FIG. 4a, or the appearance of the second embodiment in FIG. 4b, substantially rectangular. This web may be of the same material as the remainder of the block, and may be integral with it, and is useful for capturing any stray or scattered radiation; it may be disposed at any position vertically or the block (i.e. as shown in the vertical direction in FIG. 1. Where different variations or alternative arrangements are described above, it should be understood that embodiments of the invention may incorporate such variations and/or alternatives in any suitable combination.
059206020	claims	1. An underground storage facility for the interim storage of waste transportable in a container (34, 36), in particular radioactive waste such as spent fuel elements, with transport gallery (10) giving access to a storage gallery (12) for interim storage of the waste, the storage gallery (12) runs underneath the transport gallery (10) and is separated from the latter by a floor (14) designed as a transport level having closable openings (18) intended for introduction and removal of the waste and closable with locking covers (16) separating the storage gallery from the transport gallery in respect of the effects of radiation, wherein the storage gallery has an approximately rectangular cross-section and the transport gallery an approximately semi-elliptical or semi-oval cross-section, and wherein the width of the storage gallery is less than that of the transport gallery. the floor of the transport gallery (10) is a floor made of concrete (14) whose openings (18) are closable with cylindrical closing covers (16) preferably made of concrete. the openings (18) in the floor (14) of the transport gallery (10) are arranged one behind the other in the longitudinal direction of the transport gallery and along which run guides such as rails (28, 30). the storage gallery (12) of approximately rectangular cross-section is supported at the sides by earthquake protection devices such as struts (22). the storage gallery (12) is a section of a convection air channel. the storage gallery (12) has at its end filtering devices. the waste is arranged in an inner container (36) of the container (34) having an opening closable at the top, wherein the inner container is provides stronger radioactive shielding at the closing side than at the circumference and bottom sides, and wherein the inner container is surrounded inside the transport gallery (10) by a conveying container (32) closing at the bottom and using which the inner container can be placed inside the storage gallery (12) through one of the floor openings (18) of the transport gallery. the conveying container (32) has a lifting/lowering or holding device (50) for the inner container (36) on the interior side. the covers (16) closing the floor openings (18) can be lifted/lowered and transferred by a transport carriage (24) movable along the floor (14). the storage gallery (12) runs underneath the transport gallery (10) and is separated from the latter by a concrete transport floor (14), wherein closable openings (18) are provided in the transport floor and have covers (16) separating the storage gallery from the transport gallery in respect of the effects of radiation and via which inner containers (36) containing the waste can be introduced, removed and/or transferred in the storage gallery, wherein the transport gallery has a semi-elliptical or semi-oval cross-section with a width less than that of the storage gallery of approximately rectangular cross-section, and wherein the inner container is picked up inside the transport gallery by a conveying container (32) closable at the bottom and having on the inside a conveying device (50) for the inner container. the inner container containing the waste is picked up in the facility or in its immediate vicinity by a movable conveying container on its floor side that moves inside the transport gallery along a floor closing the storage gallery at the top, wherein after introduction of the inner container into the conveying container the latter is closed on the floor side and moved to an opening closed by a lid in the floor of the transport gallery, and wherein the cover is removed, the conveying container then aligned with the opening and its floor opened, and the inner container is passed by conveying elements provided in the conveying container through the opening and into the storage gallery, then the opening is closed and the conveying container is moved to a transport container to pick up an inner container or to a further opening in the transport gallery floor to remove or transfer the inner container placed in the storage gallery. the inner container is provided with corrosion protection after removal from the storage facility and is transported to a final storage facility. once the afterheat has receded the storage gallery is filled up in order to form a final storage facility. the storage gallery is filled with bentonite. 2. A storage facility according to claim 1, 3. A storage facility according to claim 1, 4. A storage facility according to claim 1, 5. A storage facility according to claim 1, 6. A storage facility according to claim 1, 7. A storage facility according to claim 1, 8. A storage facility according to claim 1, 9. A storage facility according claim 1, 10. An underground storage facility for the interim storage of waste stored in an inner container (36) such as the inner container of a transport container (34), in particular radioactive waste such as spent fuel elements, with transport gallery (10) giving access to a storage gallery (12) for interim storage of the waste, 11. A method for interim storage of waste, in particular of spent fuel elements, in an underground storage facility with transport gallery giving access to a storage gallery, the waste being transported to the storage facility in an inner container of a transport container, 12. A method according to claim 11, 13. A method according to claim 11, 14. A method according to claim 13,
043550002	summary	BACKGROUND AND SUMMARY OF THE INVENTION The present invention is related to a gate seal, and more particularly to a lightweight, removable sealing means operable to provide a fluid-tight seal between a gate and the wall of a reactor well or storage pool. In certain designs of nuclear-reactor generating plants, for example, there are large reactor well and storage pool structures wherein nuclear fuels are either stored or arranged in a reactor apparatus. These pools or wells are typically filled with a fluid, such as water, to isolate the fuels from the surrounding environment and serve as a protective outer structure for the nuclear fuel material. In order to provide access to the interior of a single well or pool or to provide connecting passageways between the various structures, the liners and walls of the structures are provided with openings including heavy, thick-walled gates to close off the openings. It is, of course, necessary that the gates be in a fluid-tight relation with the pool liner to insure the leakproof integrity of the pool structure when the gate is closed. Due to the critical nature of the desired sealing relation, it is essential that the sealing means be highly reliable and effective in providing the leak-tight seal and at the same time be easily accessible and removable for routine repair, maintenance and eventual replacement. An initial problem encountered in providing a sealing means with the above-described features is the size of the storage pool and reactor well structures. These structures are typically very large and are housed within a single building. The gates and sealing means are also of rather heavy construction and of large dimensions. To afford economical construction of the nuclear power plant, and to comply with applicable safety regulations, the outer building is designed to offer sufficient protection for the various apparatuses of the plant, but does not possess the structural strength or space required to support heavy duty hoisting and hauling machinery. Thus, it is not feasible to provide equipment capable of bodily removing an entire gate structure, in order to secure or replace the seals. It is accordingly one of the primary objectives of the present invention to provide a new form of gate sealing means which is ideally suitable for use in connection with access gates for reactor wells and fuel storage pools of a nuclear reactor power plant, for example. In accordance with the broadest aspects of the invention, there is provided a sealing means comprising a lightweight, frame structure removably mounted between the gate and portions of the pool wall when the gate is in a closed position. The frame structure is arranged to support continuous yieldable, expansible sealing members which are connectable to a source of fluid under pressure. After the gate has been closed and locked, the pressure fluid is admitted to inflate the sealing members and expand them in a manner to form a fluid-tight seal between the inner surface of the gate and the wall of the pool or well. In one advantageous form of the invention, the beam structure comprises a generally U-shaped, I-beam with transverse members extending between the spaced vertically disposed leg portions. The gate is provided with support elements engageable with the transverse members of the beam structure to support the beam on the gate. A plurality of inflatable sealing elements are arranged along the entire length of the U-shaped I-beam and are inflated after the gate is closed to form the desired leak-tight seal. The beam structure and inflatable seals provide a lightweight, easily removable sealing means to facilitate maintenance, repair and/or replacement of the inflatable sealing members. At the same time, the pneumatic sealing means is highly effective in providing a reliable leak-tight seal between the gate and the pool wall. For a better understanding of the above and other features and advantages of the invention, reference should be made to the following detailed description of preferred embodiments of the invention and to the accompanying drawings.
	description	This application is a continuation application of copending U.S. patent application Ser. No. 11/779,498, filed on Jul. 18, 2007. This application claims the benefit of Japanese Patent Application No. 2006-197747, filed Jul. 20, 2006, which is hereby incorporated by reference herein in its entirety. 1. Field of the Invention The present invention relates to, e.g., a deflector array suitable as a component of a charged particle beam exposure apparatus, such as an electron beam exposure apparatus and an ion beam exposure apparatus, used to manufacture a device, such as a semiconductor integrated device, an exposure apparatus having the deflector array, and a device manufacturing method using the exposure apparatus. 2. Description of the Related Art Conventionally, as shown in FIG. 10, a deflector 200 is used as a component of a charged particle beam exposure apparatus, such as an electron beam exposure apparatus and an ion beam exposure apparatus, used to manufacture a device, such as a semiconductor integrated device. As shown in FIG. 10, the deflector 200 has a substrate 211, an opening 212 formed to pass a charged particle beam to the substrate 211, and a pair of electrodes 213 opposing each other through the opening 212. As shown in FIG. 11, the deflectors 200 are arrayed such that a direction 221 of a line connecting the centers of two deflectors 200 located nearest to each other become perpendicular to a direction 300 in which a pair of opposing electrodes 213 deflect a charged particle beam. For this reason, a length L of the electrode 213 of the deflector 200 cannot be longer than a distance D between the centers of two deflectors located nearest to each other in the direction 221. The deflector 200 of the deflector array used for a charged particle beam exposure apparatus for drawing a pattern with a plurality of charged particle beams must be arranged at the pitch of the charged particle beams, e.g., a pitch of several tens to several hundreds of microns. This makes it impossible to sufficiently ensure the length of the electrode 213 of the deflector 200 to result in an increase in deflection aberration. When, however, the charged particle beam exposure apparatus uses a charged particle beam deflected by the deflector 200 of the deflector array, deflection aberration must be decreased to attain high drawing accuracy. To decrease the deflection aberration of the deflector 200 of the deflector array, it is effective to maximize the length of the opposing electrodes 213. Japanese Patent Laid-Open No. 7-297107 discloses deflectors arrayed to make uniform the signal delay amount. It is an object of the present invention to provide a deflector array having an arrangement advantageous to decrease deflection aberration, and an application example of the deflector array. A first aspect of the present invention relates to a deflector array in which a plurality of deflectors, which deflect charged particle beams, are arrayed on a substrate. In the array, each deflector includes an opening formed on the substrate, and a pair of electrodes opposing each other through the opening. The length of the pair of electrodes in the longitudinal direction is not less than the distance between the centers of two deflectors, which are located nearest to each other. According to a preferred embodiment of the present invention, the direction of a line connecting the centers of two deflectors, which are located nearest to each other, form an angle of 45° with respect to the direction in which the deflector deflects the charged particle beam. Alternatively, according to another preferred embodiment of the present invention, the direction of a line connecting the centers of two deflectors, which are located nearest to each other, form an angle of 63.4° with respect to the direction in which the deflector deflects the charged particle beam. According to a preferred embodiment of the present invention, the direction in which the deflector deflects the charged particle beam is perpendicular to the longitudinal direction of the pair of electrodes. According to a preferred embodiment of the present invention, the pair of electrodes are parallel to each other. Alternatively, according to another preferred embodiment of the present invention, the distance between the pair of electrodes shortens toward the end portions of the pair of electrodes. A second aspect of the present invention relates to an exposure apparatus which exposes a wafer with a charged particle beam. The exposure apparatus includes a charged particle source, which emits the charged particle beam, a first electron optical system which forms a plurality of intermediate images of the charged particle source, a second electron optical system which projects the plurality of intermediate images formed by the first electron optical system onto the wafer, and a positioning apparatus which holds, drives, and positions the wafer. The first electron optical system includes the above-described deflector array. A third aspect of the present invention relates to a device manufacturing method. The manufacturing method includes the steps of exposing a wafer using the above-described exposure apparatus, and developing the wafer. According to the present invention, a deflector array having an arrangement advantageous to decrease deflection aberration, and an application example of the deflector array are provided. Further features of the present invention will become apparent from the following description of exemplary embodiments, with reference to the attached drawings. Embodiments of the present invention will be described below with reference to the accompanying drawings. [First Embodiment] A deflector array according to the first embodiment of the present invention will be explained with reference to the plan view shown in FIG. 1. The deflector array is formed by arraying a plurality of deflectors 200a and 200j for deflecting charged particle beams on a substrate 211. Each deflector has an opening 212a formed on the substrate 211, and a pair of electrodes 213a opposing each other through the opening 212a. A length L of the pair of electrodes 213a in the longitudinal direction is not less than a distance D between the centers of the two deflectors 200a and 200j located nearest to each other. In the deflector array according to the first embodiment, therefore, the length L of the pair of electrodes 213a in the longitudinal direction can be relatively long. This makes it possible to decrease the deflection aberration of charged particle beams deflected by the deflectors 200a and 200j. In the first embodiment, a direction 300 in which a deflector deflects a charged particle beam is perpendicular to the longitudinal direction of the electrode 213a. Also, in the first embodiment, the distance between each pair of electrodes 213a shortens toward their end portions. Applying the deflector array according to the first embodiment to an electron beam exposure apparatus makes it possible to attain high drawing accuracy. [Second Embodiment] A deflector array according to the second embodiment of the present invention will be explained with reference to the plan view shown in FIG. 2. Directions 246 and 247 form an angle of about 45° with respect to a direction 300 in which a deflector deflects a charged particle beam. The directions 246 and 247 are of lines connecting the center of a deflector 200b to the centers of deflectors 200d and 200e and to the centers of deflectors 200c and 200f, respectively, all of which are located nearest to the deflector 200b. In the deflector array according to the second embodiment, a length L of electrodes 213a of the deflectors 200b, 200c, 200d, 200e, and 200f can be about √{square root over (2)} times as long as a distance D between the center of the deflector 200b and the centers of the deflectors 200c, 200d, 200e, and 200f, all of which are located nearest to the deflector 200b. According to the second embodiment, it is possible to obtain a deflector array, which minimizes deflection aberration. In addition, applying the deflector array according to the second embodiment to an electron beam exposure apparatus makes it possible to attain high drawing accuracy. [Third Embodiment] A deflector array according to the third embodiment of the present invention will be explained with reference to the plan view shown in FIG. 3. One of directions 291 and 292 forms an angle of about 63.4° with respect to a direction 300 in which a deflector deflects a charged particle beam. The directions 291 and 292 are of lines connecting the center of deflector 200g to the centers of deflectors 200h located nearest to it. In the deflector array according to the third embodiment, a length L of electrodes 213b of the deflectors 200g and 200h can be about √{square root over (5)} times as long as a distance D between the center of the deflector 200g and the centers of the deflectors 200h located nearest to it. According to the third embodiment, it is possible to obtain a deflector array which minimizes deflection aberration. In addition, applying the deflector array according to the third embodiment to an electron beam exposure apparatus makes it possible to attain high drawing accuracy. An electron beam exposure apparatus (drawing apparatus) using a deflector array, according to an embodiment of the present invention, will be explained. The following description will exemplify an exposure apparatus which adopts an electron beam as the charged particle beam. However, the present invention is also applicable to an exposure apparatus using a charged particle beam of another type, such as an ion beam. An electron beam exposure apparatus using a deflector array according to the present invention will be explained with reference to the schematic views of the main part shown in FIGS. 4A and 4B. A multi-source module 1 forms a plurality of electron source images by emitting electron beams from its electron source (charge particle source). In this example, 3×3 multi-source modules 1 are arrayed, and details thereof will be described later. In this example, each of magnetic field lens arrays 21, 22, 23, and 24 forms an electron optical system and has magnetic disks MD. The magnetic disks MD have 3×3 openings with the same shape, and are vertically arranged with spacings between them. A common coil CC excites the magnetic disks MD. As a consequence, each opening serves as a magnetic pole of a magnetic field lens ML to generate a lens magnetic field as designed. Four magnetic field lenses ML1, ML2, ML3, and ML4 corresponding to the magnetic field lens arrays 21, 22, 23, and 24 project the plurality of electron source images of each multi-source module 1 onto a wafer 4. An optical system, which acts on electron beams from one multi-source module 1 until they strike the wafer, is defined as a column. That is, in this example, the exposure apparatus includes nine columns, column 1 to column 9. The two corresponding magnetic field lenses of the magnetic field lens arrays 21 and 22 once form an image. Then, the two corresponding magnetic field lenses of the magnetic field lens arrays 23 and 24 project the resultant image onto the wafer 4. The common coils individually control the respective excitation conditions of the magnetic field lens arrays 21, 22, 23, and 24. This makes it possible to adjust the optical characteristics (focal position, image rotation, and magnification) of each column uniformly, i.e., by the same amount. A main deflector 3 is a positioning apparatus for deflecting a plurality of electron beams from the multi-source module 1 and displacing a plurality of electron source images in the X and Y directions on the wafer 4. A stage 5 is a positioning apparatus which supports the wafer 4 to be movable in the X and Y directions perpendicular to an optical axis AX (Z-axis) and the rotation direction about the Z-axis. A stage reference plate 6 is fixed on the stage 5. A reflected electron detector 7 detects electrons reflected when an electron beam strikes a mark on the stage reference plate 6. FIG. 5 is a view showing details of one column. The multi-source module 1 and its function of adjusting the optical characteristics of an electron beam applied from the multi-source module 1 to the held wafer 4 will be explained. An electron source 101 (charged particle source) formed by an electron gun emits an electron beam to form a crossover image. A condenser lens 102, which forms an electron optical system, collimates the electron beam emitted by the electron source 101 into a collimated electron beam 101a. The condenser lens 102 in this example is an electro-static lens including three opening electrodes. An aperture array 103 is an electron optical system having a plurality of two-dimensionally arranged openings. A lens array 104 is an electron optical system having a plurality of two-dimensionally arrayed electro-static lenses. The plurality of electro-static lenses have the same optical power. Deflector arrays 105 and 106 are electron optical systems, each of which is formed by two-dimensionally arraying electro-static deflectors that can be driven individually. A blanker array 107 is an electron optical system formed by two-dimensionally arrayed electro-static blankers that can be driven individually. A deflector array represented by those according to the above-described first to third embodiments is suitable as the deflector arrays 105 and 106 and blanker array 107. The functions of units of the exposure apparatus will be explained with reference to FIG. 6. The aperture array 103 divides a collimated electron beam from the condenser lens 102 into a plurality of electron beams. Each divided electron beam forms an intermediate image of the electron source (charged particle source) on a corresponding blanker of the blanker array 107 via a corresponding electro-static lens of the lens array 104. At this time, the deflector arrays 105 and 106 individually adjust the positions (positions within a plane perpendicular to the optical axis) of the intermediate images of the electron source formed on the blanker array 107. An electron beam deflected by the blanker array 107 is shielded by a blanking aperture AP shown in FIG. 5 and, therefore, does not reach the wafer 4. On the other hand, an electron beam, which is not deflected by the blanker array 107, is not shielded by the blanking aperture AP shown in FIG. 5 and, therefore, reaches the wafer 4. Referring back to FIG. 6, each of a plurality of intermediate images 101b of the electron source (charged particle source) formed by the multi-source module 1 is projected onto the wafer 4 via the two corresponding magnetic field lenses of the magnetic field lens arrays 21 and 22. Of the optical characteristics, when the plurality of intermediate images are projected onto the wafer 4, the image rotation and magnification can be adjusted by the deflector arrays 105 and 106 capable of adjusting the position of each intermediate image on the blanker array 107. The focal position can be adjusted by dynamic focus lenses (electro-static or magnetic field lenses) 108 and 109 arranged for each column. A system configuration of the exposure apparatus will be explained with reference to the system configuration shown in FIG. 7. A blanker array control circuit 41 individually controls the plurality of blankers of the blanker array 107. A deflector array control circuit 42 individually controls the deflectors of the deflection arrays 104 and 105. A D_FOCUS control circuit 43 individually controls the dynamic focus lenses 108 and 109. A main deflector control circuit 44 controls the main deflector 3. A reflected electron detection circuit 45 processes a signal from the reflected electron detector 7. The blanker array control circuit 41, deflector array control unit 42, D_FOCUS control circuit 43, main deflector control circuit 44, and reflected electron detection circuit 45 are prepared for each of the columns, column 1 to column 9. A magnetic field lens array control circuit 46 controls the common coils of the magnetic field lens arrays 21, 22, 23, and 24. A stage driving control circuit 47 controls the driving of the stage 5 in cooperation with a laser interferometer (not shown) for detecting its position. A main control system 48 controls the above-described plurality of control circuits to manage the overall electron beam exposure apparatus. An embodiment of a device manufacturing method using the above-described exposure apparatus will be explained with reference to FIGS. 8 and 9. FIG. 8 is a flowchart for explaining the manufacture of a device (e.g., a semiconductor chip, such as an IC or LSI, an LCD, or a CCD). A semiconductor chip manufacturing method will be exemplified here. In step 1 (circuit design), the circuit of a semiconductor device is designed. In step 2 (exposure control data preparation), exposure control data is prepared on the basis of the designed circuit pattern. In step 3 (wafer manufacture), a wafer is manufactured using a material such as silicon. In step 4 (wafer process), called a pre-process, circuit patterns are formed on the wafer by using the above-described exposure apparatus. The exposure apparatus is controlled by the exposure control data. In step 5 (assembly), called a post-process, a semiconductor chip is formed using the wafer manufactured in step 4. This step includes an assembly step (dicing and bonding) and a packaging step (chip encapsulation). In step 6 (inspection), the semiconductor device manufactured in step 5 undergoes inspections, such as an operation confirmation test and a durability test. After these steps, the semiconductor device is completed and shipped, in step 7. FIG. 9 is a flowchart showing details of the wafer process in step 4. In step 11 (oxidation), the wafer surface is oxidized. In step 12 (CVD), an insulating film is formed on the wafer surface. In step 13 (electrode formation), an electrode is formed on the wafer by vapor deposition. In step 14 (ion implantation), ions are implanted in the wafer. In step 15 (resist process), a photosensitive agent is applied to the wafer. In step 16 (exposure), the exposure apparatus draws the circuit pattern on the wafer in accordance with the exposure control data. In step 17 (development), the exposed wafer is developed. In step 18 (etching), portions other than the developed resist image are etched. In step 19 (resist removal), any unnecessary resist remaining after etching is removed. These steps are repeated to form multiple circuit patterns on the wafer. While the present invention has been described with reference to exemplary embodiments, it is to be understood that the invention is not limited to the disclosed exemplary embodiments. The scope of the following claims is to be accorded the broadest interpretation so as to encompass all such modifications and equivalent structures and functions.
	abstract	Illustrative embodiments provide a nuclear fission reactor, a vented nuclear fission fuel module, methods therefor and a vented nuclear fission fuel module system.
	summary
	abstract	A unitary support device for a polycapillary optic is provided wherein a housing has a central opening therethrough and at least two locating structures, such as positioning shoulders, formed therein. Each locating structure is sized and positioned to accommodate a different polycapillary positioning component within the housing. Each polycapillary positioning component has at least one opening for holding at least one polycapillary of the polycapillary optic. One or more coaxial bores can define the central opening of the housing and the locating shoulders in one continuous fabrication operation. Depending upon the polycapillary positioning components employed, i.e., location of the openings therein for accommodating the polycapillaries, the positioning components can be oriented within the housing such that radiation from one of a divergent beam, a focused beam, or a parallel beam is collected by the optic, and such that the optic can output one of a collimated beam, a focused beam or a divergent beam.
	summary
	description	This application claims Paris convention priority from EP 14 185 226.9 filed Sep. 17, 2014 the entire disclosure of which is hereby incorporated by reference. The invention relates to an X-ray computed tomography (=CT) apparatus, comprising an X-ray source for emitting an X-ray beam, in particular a divergent X-ray beam, along a beam axis (z), a filter element for attenuating the X-ray beam, a stage for an object to be investigated with the attenuated X-ray beam, a 2D X-ray detector with a detection area, and a gantry system capable of rotating either the entirety of the X-ray source, the filter element and the 2D X-ray detector, or the stage for the object with respect to a rotation axis (x) which is perpendicular to the beam axis (z), wherein the filter element has a spatially varying X-ray absorption capability along a cross direction (y) which is perpendicular to both the beam axis (z) and the rotation axis (x). Such a CT apparatus is known from US 2012/0002782 A1. The use of X-rays in imaging is a well-established technology, in particular in medicine and science. The basic principle is that the X-rays produced by an X-ray source are partially attenuated by an object to be investigated, and that the X-rays which have enough energy to pass through the object are detected by a two dimensional (2D) X-ray detector (camera), resulting in a two-dimensional (2D) image. This image is also referred to as a projection image. Different types of X-ray sources are used in combination with different cameras, however, the basic principle remains. In order to generate a three-dimensional (3D) image of the object, projection images have to be acquired from different angles. This can be done by rotating the object, or by rotating the X-ray source together with the camera around the object. In both cases a dataset of 2D projection images is generated. These projection images are then used to generate a new dataset of cross-sections by a process called back-projection. It is from this second cross-sectional dataset from which the 3D information can be obtained. This method is also referred to as computed tomography (CT). A particular field of interest in preclinical research is the investigation of small living animals, such as mice or rats. When dealing with live animals, rotating the animal is often not advisable. Therefore special scanners have been developed where the animal is placed on a bed, which, just like in a clinical scanner, moves inside the scanner while both the X-ray source and the camera rotate around the bed acquiring projection images at multiple angles. Imaging live animals using X-rays always results in a certain radiation dose; without X-rays there would not be an image. An important point to consider when scanning live animals is to reduce the radiation dose in order to minimize or even avoid any biological effects. In the past, filters have been placed between the X-ray source and the object; the filters were designed to preferentially absorb the low energy X-rays. Typically, these filters were made of sheets of metal (e.g. aluminum) with a well-defined and uniform thickness. Using such a filter, the low energy X-rays are predominantly attenuated by the filter and no longer by the animal, and the total radiation dose is reduced. Increasing the thickness of the filter or changing the filter type (e.g. from aluminum to copper) will absorb more X-rays, thereby further reducing the radiation dose. However, if the X-ray intensity becomes too low, or the X-ray energy becomes too high, the image quality will deteriorate. Therefore, optimal combinations of X-ray energies and filters are looked for in order to minimize the total radiation dose while still obtaining an image with enough information allowing image analysis. US 2012/0002782 A1 proposes a CT apparatus wherein an X-ray source and a 2D X-ray detector are rotated around a bed for a human patient. A compensation filter with a varying thickness, and therefore a varying X-ray absorption capability, is located in front of the X-ray source. The center of the compensation filter, which shadows X-rays intersecting an axis of rotation, exhibits a minimum thickness, and thus a minimum of X-ray absorption. Another CT apparatus is known from WO 2013/182928 A1, wherein a filter of basically triangular shape is introduced into an X-ray beam from the side, such that the X-ray beam is attenuated to the largest degree at its edge, and less at its center. The latter filters, or rather beam shaping inserts or absorbers, are intended for equalizing intensity of radiation which reaches the detector array. The absorber provides a thin layer of material in the central part of the beam and thicker layers in periphery areas. The shape of such absorbers is based on the assumption that a human or animal body shows more absorption of X-ray radiation in the central area and less in the periphery. The shape of the absorbing layer applied to the primary beam is inverted to the expected local thicknesses of the scanning object, which finally equalizes the signals on the detector array across illuminated area. In this way, the required dynamic range of detectors used in tomographical and microtomographical set-ups can be reduced. It is the object of present invention to provide a CT apparatus which allows to further reduce the radiation dose for an object to be investigated, while retaining a high image quality at the same time. This object is achieved, in accordance with the invention, by a CT apparatus as introduced in the beginning, characterized in that the spatially varying X-ray absorption capability exhibits a maximum absorption along the cross direction (y) at a zero position (y0), wherein X-rays passing through the filter element at the zero position (y0) intersect the rotation axis (x). The basic idea of the present invention is to deliberately reduce the X-ray intensity close to the rotation axis of the gantry system, whereas farther away from the rotation axis, the X-ray intensity is kept higher. Further, in accordance with the invention, the average X-ray energy close to the rotation axis may be increased, whereas farther away from the rotation axis, the average X-ray energy may be kept lower, and preferably nearly unchanged. This is done by means of a corresponding filter element with a spatially varying X-ray absorption capability. The filter has a maximum absorption, and may create a maximum shift to higher average X-ray energies, in the region corresponding to (i.e. shadowing) the rotation axis, i.e. at the zero position y0, typically by means of a maximum thickness there. In general, the filter element attenuates low energy X-rays to a higher degree, and preferably much higher degree, than high energy X-rays. Reducing the X-ray intensity close to the rotation axis will reduce the signal to noise ratio in the corresponding region of the detection area of the 2D X-ray detector (typically at the center of its detection area) and the corresponding projection images, respectively, which also deteriorates the image quality of cross sections. However, when calculating cross-sections in the course of the reconstruction algorithm (by back projection, in particular filtered back projection), more information is available from the various projection images the closer you come to the rotation axis, increasing the image quality of cross-sections there. Therefore, the loss in signal to noise ratio will be compensated for by the information increase closer to the rotation axis. Preferably, the intensity reduction approximately matches the information increase such that a uniform image quality within a cross-section (from its center to its edges) is obtained. As a result, the reduction in X-ray intensity and the increase of the average X-ray energy close to the rotation axis will decrease the overall radiation dose for the object to be investigated (such as a living human or animal body or a part thereof), in particular in a region which is often very sensitive to radiation damage (such as the bone marrow or intestines), without reducing the overall image quality of cross-sections. Note that the elements of the CT apparatus are arranged such that an X-ray beam emitted by the X-ray source first passes through the filter element, then passes through the object on the stage, and then reaches the 2D X-ray detector. Note that typically, the X-ray absorption capability of the filter element is constant in the direction of the rotation axis (x). In a preferred embodiment of the inventive CT apparatus, the filter element is made of a material exhibiting a higher absorption for low energy X-rays than for high energy X-rays. By this means, the energy spectrum of the X-ray source is modulated; the fraction of high energy X-rays increases as compared to low energy X-rays. Low energy X-rays are more absorbed in human or animal tissue than high energy X-rays, and therefore tend towards causing more accumulated damage to body tissue and/or DNA. High energy X-rays, in contrast, are well suited for transmission through relatively thick body parts, such as the head or the chest, yielding good contrast still. In a preferred further development of the above embodiment, the material exhibits an absorption coefficient μ for 5 keV X-rays which is at least 10 times larger, preferably at least 50 times larger, than an absorption coefficient μ for 40 keV X-rays. This reduces damage to the object and at the same time provides good depth contrast. Note that the absorption behavior of a material can be adjusted in particular via the atomic number of elements contained. Note that the material of the filter element can also be chosen with μ (5 keV) being at least 20 times larger, and preferably at least 100 times larger, than μ (40 keV), in accordance with the invention. In another preferred embodiment, the filter element is made of aluminum or a material containing aluminum. Aluminum (Al) exhibits an absorption strongly dependent on the X-ray energy. Preferably, the material contains at least 30 wt % (percent per weight) of Al, more preferably at least 50 wt % of Al, and most preferably at least 80 wt % of Al. Particularly preferred is an embodiment wherein the filter element has a spatially varying thickness along the cross direction (y), with a maximum thickness at the zero position (y0). This is the simplest way to establish a spatially varying X-ray energy distribution and absorption capability of the filter element, typically with the material of the filter element having a homogeneous (spatially constant) absorption coefficient. The thickness is measured in parallel to the beam axis (z). Note that typically, the thickness does not vary along the direction of the rotation axis (x). Further note that alternatively or in addition, the density and/or the composition of the material of the filter element may vary along the cross direction (y) in order to establish a varying X-ray absorption capability. In a further development of this embodiment, the filter element exhibits a thickness variation along the cross direction (y) by a factor of 2 or more, preferably by a factor of 5 or more, in the area of the filter element which shadows the detection area of the 2D x-ray detector. Such a variation can easily be manufactured and establishes an intensity variation of the transmitted high energy X-rays at the projection images which can well be compensated for under typical measurement conditions. Note that thickness variation typically does not exceed a factor of 20, though. In an advantageous embodiment, the spatially varying X-ray absorption capability exhibits an absorption variation along the cross direction (y) by a factor of 5 or more, preferably by a factor of 20 or more, most preferably by a factor of 50 or more, in the area of the filter element which shadows the detection area of the 2D X-ray detector. The factors apply at least to X-rays of 5 keV (and preferably to all low energy X-rays of about 1 keV-10 keV), and possibly also to high energy X-rays (with an X-ray energy of about 10 keV-100 keV); however note that the absolute absorption is typically much less for high energy X-rays than for low energy X-rays. The factors compare the strongest filtered X-ray intensity with the weakest filtered X-ray intensity behind the filter element (assuming a uniform illumination of the filter element). This embodiment has shown good radiation dose reduction and still high image quality in practice. An intensity factor of about 50 (and often more) for the projection images can, under typical measurement conditions, still be compensated for by back projection when calculating cross-sections. Also preferred is a an embodiment wherein the X-ray absorption capability exhibits a monotonic decreasing absorption, in particular strict monotonic decreasing absorption, on both sides away from the zero position (y0) along the cross direction (y), in the area of the filter element which shadows the detection area of the 2D X-ray detector. With increasing distance from the rotation axis, the information from the entirety of the projection images becomes less, and therefore less absorption of the filter element can be compensated for. With this embodiment, gradually more X-ray intensity is provided at the object when departing from the rotation axis, so uniform cross-section image quality may be obtained in summary. With a strict monotonic decreasing absorption, sharp changes in the absorption capability of the filter element are excluded, so artefacts in image reconstruction are avoided. Note that this embodiment is typically realized by a thickness variation of the filter element exhibiting a monotonic decreasing thickness, in particular strict monotonic decreasing thickness, on both sides away from the zero position (y0) along the cross direction (y), in the area of the filter element which shadows the detection area of the 2D X-ray detector. Another preferred embodiment provides that the spatially varying X-ray absorption capability exhibits at least approximately a Gaussian type distribution of the absorption with respect to the cross direction (y), centered around the zero position (y0), in the area of the filter element which shadows the detection area of the 2D X-ray detector. With the Gaussian type distribution, the intensity reduction towards the rotation axis approximately matches the information increase such that a uniform image quality within a cross-section (from its center to its edges) is obtained. The absorption capability is considered to be approximately of Gaussian type if the actual absorption deviates from an exact Gaussian distribution by at most 10% (with the basis being the exact Gaussian distribution), in the area of the filter element which shadows the detection area of the 2D X-ray detector. Note that equally preferred is a thickness variation of the filter element exhibiting at least approximately a Gaussian type distribution of the thickness with respect to the cross direction (y), centered around the zero position (y0), in the area of the filter element which shadows the detection area of the 2D X-ray detector. The thickness variation is considered to be approximately of Gaussian type if the actual thickness deviates from an exact Gaussian distribution by at most 10% (with the basis being the exact Gaussian distribution), in the area of the filter element which shadows the detection area of the 2D X-ray detector. Again, a basically uniform image quality within a cross-section (from its center to its edges) may be obtained. Further preferred is an embodiment wherein the X-ray source is a microfocus X-ray tube, in particular with a tungsten anode. By this means, a high intensity of polychromatic radiation with high energy X-rays of up to about 65 keV, or even up to about 100 keV, may be obtained. In another preferred embodiment, an object to be investigated is located on the stage, in particular with longitudinal object orientation along the rotation axis (x). By this means, the object can be investigated with a minimum radiation dose. The object may be located such that parts (e.g. organs of a live human or animal) particularly sensitive to radiation damage are intersected by the rotation axis or located close to the rotation axis. Further, the object may be scanned easily by translation along the direction of the rotation axis by means of the longitudinal object orientation. In a highly advantageous embodiment, the stage is moveable along a longitudinal axis (LA) which is parallel to the rotation axis (x). By this means, larger volumes of the object may be investigated by scanning along the longitudinal axis. Note that the translation along the longitudinal axis is typically motor driven and automated. Also within the scope of the present invention is the use of an inventive CT apparatus described above, characterized in that an object to be investigated is located on the stage, that a plurality of projection images of the object are recorded with the 2D X-ray detector at different rotation positions of the gantry system, and that from the dataset of projection images, a dataset of cross-sections is generated by a back projection reconstruction algorithm. By this means, good quality 3D images of the object may be obtained at a reduced radiation dose for the object, in particular a living object. The back projection (in particular filtered back projection) reconstruction algorithm compensates for the signal to noise reduction due to reduced intensity in the center region near the rotation axis. The closer a volume pixel is to the rotation axis, the more projection images contain information from this volume pixel which is available for compensation purposes. Particularly preferred is a variant of the inventive use, in correlation with the inventive CT apparatus comprising a movable stage, wherein the plurality of projection images of the object are recorded also at different movement positions of the stage along the longitudinal axis (LA). By this means, a volume of the object extending arbitrarily far along the longitudinal direction may be investigated. Further preferred is a variant of the inventive use wherein the object is a living animal of up to the size of a rat or a non-living object of up to the same size. With objects of a relatively small size, such as up to a rat (comparable to a cylindrical volume of about 8 cm in diameter and 20 cm in length), a relatively large fraction of the object can benefit from the higher information density close to the rotation axis, so the achievable radiation reduction is particularly significant here. However note that the invention can also be used for CT apparatus dimensioned for scanning complete human beings. Further advantages can be extracted from the description and the enclosed drawing. The features mentioned above and below can be used in accordance with the invention either individually or collectively in any combination. The embodiments mentioned are not to be understood as exhaustive enumeration, rather have exemplary character for the description of the invention. The invention is shown in the drawing. In X-ray computed tomography, a large number of X-ray images (projection images) are recorded, with each recording coming along with an exposure of the object to be investigated with X-rays. X-rays may damage the material exposed, in particular by ionization. In human or animal tissue, X-rays may in particular induce carcinogenesis. The present invention therefore seeks to reduce the radiation dose to objects investigated in X-ray computed tomography. The absorbed dose depends on absorbing coefficients of materials in the object, thickness of the object materials, and the intensity and energy distribution in the primary beam. The first two factors cannot be changed by the scanning setup, but the energy spectrum and intensity distribution in the primary beam can be adjusted in such a way that the absorbed dose will be reduced without significant impact to the quality of results of tomographical reconstruction, in accordance with the invention. The central, in general thickest part of the object to be investigated, such as an animal or human body, absorbs much more radiation than peripheral parts with shorter passages of the X-ray beam through the body. At the same time, the central part of the body contains organs most sensitive to radiation damages. The invention is based on the idea of a spatial modulation of the energy spectrum and the intensity of the primary beam by such a way that the central part of the body will receive less intensity and higher energy of X-ray radiation. Both intensity reduction and spectral shift to higher energies are reducing dose rate absorbed by this most sensitive part of the body. Such a reduction of intensity in the primary beam may reduce the signal on the central part of the detector array behind the object, which in general will at least partially, and in most cases practically completely be compensated during tomographical reconstruction, since the central part of the reconstructed area is covered by a more-consistence ray pattern during the back-projection procedure. The core of the present invention is a beam-shaping filter element, an embodiment of which is shown in FIG. 1. FIG. 1 shows a primary X-ray beam 1 (here of parallel type), directed onto a beam-shaping filter element 2. The filter element 2 shadows an object 3 to be investigated, i.e. the X-ray beam 1 is attenuated (but not completely blocked) when passing through the filter element 2. The filter element 2 has a variable local thickness d (measured z direction, along which the X-ray beam 1 propagates), with its maximum thickness at its center 4, which is aligned here with the center 5 of the object 3, both with respect to the y direction. The thickness d decreases symmetrically towards the sides, and minimum thicknesses are reached at the edges 12, 13. The size (width) of the beam-shaping filter element 2, both in x and y direction, is similar to the size of the object 3 in the case that the filter element 2 is placed close to the object and far from the X-ray source (not shown); this situation is illustrated in FIG. 1. In the case of a short distance between such a beam shaping filter element 2 and an emission point inside the X-ray source, the size of the filter element 2 should be reduced proportionally. Note that it is generally preferred that no unfiltered X-rays reach the detector; if necessary, a too small filter element may be complemented with an aperture to match the x-ray beam cross-section. The selection of the filter material is dependent on the energy distribution of emission from the X-ray source. For the most typical range of energies between 40 and 100 keV for in-vivo microtomography of small laboratory animals, the material can be Al or Al alloys. The peak thickness and the thickness distribution of the filter element 2 across the beam (in y direction) is chosen such that a significant cut of the low-energy part of X-ray spectrum in the central part of the beam is achieved, while keeping enough intensity of the primary beam in this part for obtaining a reasonable signal-to-noise ratio on the detector behind the object. With the object in place, the cut (attenuation) in the low-energy part of X-ray spectrum in the central part of the object does not significantly change the signal distribution (with respect to both intensity end energy) arriving at the detector, because the object in the central part is also working as a strong filter for low energy radiation. The absorption of the low-energy photons by the beam-shaping filter element 2 before they reach the object makes a big improvement (i.e. reduction) in the dose absorbed inside the body, though, especially in the parts facing towards primary beam. The shape of the filter element surface should be smooth enough to avoid sharp intensity and spectrum variations, which can create reconstruction artifacts. In a practical implementation of such a beam-shaping filter element in accordance with the invention, for an in-vivo small animal microtomography system, Al was used as a filter material with thickness variation from 0.5-0.8 mm in the side parts of the filter element up to 5-6 mm in the central part. The polychromatic radiation from a microfocus X-ray tube with peak energy up to 65 keV and a Tungsten anode was used as a primary beam. A measured value of the dose rate, absorbed by an animal, has been estimated using a rate meter with a probe surrounded by a “mouse-size” plastic phantom. Using the described above beam-shaping filter element allows reducing the absorption dose 2-5 times compared to scanning with standard flat Al filters 0.5-1 mm thick and scanning with a non-filtered primary beam. Such reduction of the absorbed dose produces only very small influence on the quality of results, obtained after tomographical reconstruction. FIG. 2 shows an embodiment of an X-ray CT apparatus 10 in accordance with the invention. An X-ray source 11 emits at an X-ray beam 1, here of divergent type, along a (central) beam axis z. The X-ray beam 1 then reaches a filter element 2 made of aluminum here, where the X-ray beam 1 is attenuated as a function of the location in a cross direction y, which extends perpendicular to the beam axis z. At a center 4, i.e. at a zero position y0 in y direction, the thickness (measured in z) of the filter element 2 is maximum, and towards both sides (in FIG. 2 up and down) or edges 12, 13, respectively, the thickness decreases in a strict monotonic way. The thickness in z as a function of y is approximately of Gaussian type here, centered around y0. In a longitudinal direction which is perpendicular to both z and y, compare x, the thickness of the filter element 2 is not changing (note that in case of an uneven intensity distribution in x direction of the X-ray source 11, or in the X-ray beam 1, respectively, the thickness of the filter element 2 may be varied in x direction correspondingly, though). The shape of the filter element 2 leads to a strong attenuation of the X-ray beam 1, in particular for its low energy part, near the center 4, and to a decreasing attenuation towards the edges 12, 13. Since the thickness at the center is here about 5 times larger than the thickness at the edges 12, 13, the variation of absorption should be at least about a factor of 5, and typically is much higher due to the exponential absorption behavior. The attenuated X-ray beam 1a then approaches a stage 14, here designed as a bed for a living animal of up to the size of a rat as an object (not shown) to be investigated. Close to the stage 14, in an area where the object is to be placed, lies a rotation axis x of the CT apparatus 10. The rotation axis x runs perpendicular to both the beam direction z and the cross direction y. The part of the X-ray beam 1 which passes the filter element 2 at the center 4 runs through the rotation axis x; in other words, the emission point 11a of the X-ray source 11, the center (or rather center line) 4 and the rotation axis x lie in the same plane. The X-ray beam 1b after passing the stage 14 (and the object not shown) finally reaches a 2D X-ray detector 15 (also called camera) for recording projection images of the object. If no object is on the stage 14, the X-ray intensity distribution on the detection area 15a will exhibit a minimum level (marked dark) in a central plane corresponding to the zero position y0 at the center 4 of the filter element 2 or the rotation axis x, respectively, and increasing levels (marked lighter) when going up or down, with respect to the y direction. Accordingly, when placing an object at the stage 14, parts of the object intersected by the rotation axis x exhibit a minimum radiation dose, and the radiation dose increases with larger distance from the rotation axis x in y direction. More specifically, central part X-rays 18 of the X-ray source 11 propagating along the beam axis z (or in the corresponding xz plane) pass the filter element 2 at zero position y0 in y direction, intersect the rotation axis x and reach the detection area 15a here at its center 19, wherein said center part X-rays 18 experience a maximum absorption at the filter element 2. In contrast, X-rays with a higher or lower y position at the filter element 2 as compared to y0 will experience less absorption at the filter element 2. It should be noted that preferably, the filter element 2 shadows the complete detection area 15a of the 2D X-ray detector 15. For obtaining projection images from different angles with respect to the object, a gantry system 16 is provided. In the embodiment shown, the gantry system 16 couples the X-ray source 11, the filter element 2 and the 2D X-ray detector 15 and allows the rotation of the entirety of said items with respect to the rotation axis x, i.e. around the non-rotating stage 14, compare rotation arrow 17. Note that it is also possible to have a gantry system 16 which rotates the stage 14 relative to the fixed entirety of the X-ray source 11, the filter element 2 and the 2D X-ray detector 15, in accordance with the invention. The stage 14 is further movable along a longitudinal axis LA (which is in parallel with the rotation axis x), so projection images can be obtained for different rotation positions of the gantry system 16 and for different translation positions along the longitudinal axis LA of the stage 14. In summary, the invention presents a particular design of the filter element 2 which is placed in between the X-ray source 10 and the object on the stage 14 of a CT apparatus 10. As the X-rays 1 are attenuated by the filter element 2, the 3D shape of the filter element 2 will result in an uneven distribution in such a way that the center 4 of the filter element 2 along the cross direction y attenuates and cuts out low X-ray energies more as compared to the peripheral part. The 3D shape of the filter element 2 is to maximize the attenuation of the X-rays in the center 4 along the Y-axis. As X-rays pass through the filter element 2, there is an exponential decay in the X-rays that reach the camera the closer to the center 4 the X-rays pass through the filter element 2, where it has the maximum thickness (along the Y-axis). This is based on the Beer-Lambert law: I=I0e−μd where d is the thickness of the material, μ is the absorption coefficient of the material, I is the transmitted intensity and I0 is the initial intensity. Because of the exponential decrease in X-rays reaching the camera in the central position (along the Y-axis) the result is a decrease in the radiation dose. This uneven distribution also results in a reduced signal to noise ratio in the center of the camera on the projection images, but is overcome by the reconstruction algorithm (back projection) when calculating the cross-sections. This is based on the fact that during the back projection there is more information as you reconstruct closer to center of the rotation axis (comparable to the distance between the spokes on wheel, where the closer you get to the center, the smaller the distance becomes). The combination of the 3D shape of the filter and the back projection algorithm used results in a reduced radiation dose without compromising the image quality.
	description	Referring to FIG. 1, a typical container 1 for receiving nuclear fuel elements (not shown) comprises an elongate cylindrical casing 2 within which is housed a plurality of compartments 3. The compartments 3 are located within the casing 2 by a plurality of baffle plates 4 positioned at intervals along the container 1. The compartments 3 are of rectangular cross section and are formed by two sets of elongate plates 5, 6. A first set of plates 5 extends perpendicularly with respect to the second set of plates 6 so as to form opposing walls of the compartments 3. Although four compartments have been shown in FIG. 1, the number of compartments will be chosen to suit a particular application. Each compartment 3 is of a size that is sufficient to accommodate a nuclear fuel element. Referring now to FIGS. 2, 3 and 4, the first set of plates 5 comprises a plurality of interconnected individual plates, two of which plates 7, 8 are shown. Similarly, the second set of plates 6 comprises a plurality of individual plates, two of which plates 9, 10 are shown. A joint at a typical cross junction of the two sets of plates if shown in FIGS. 2, 3 and 4. The, plates 7, 8 of the first set are formed with a series of alternating projections 11 and recesses 12 along the longitudinal edge, the projections and recesses being arranged at a regular pitch or spacing. Each projection 11 is generally T-shaped defined by a first portion 13 extending from the longitudinal edge and two arm portions 14. The arm portions 14 extend laterally from an end of the first portion 13 at each side thereof. Each recess 12 is formed to correspond in shape to the T-shaped projections 11. An initial passage 15 of the recess 12 extends from the longitudinal edge. At an end of the passage 15 are two laterally extending side portions 16 which are formed so as to define two shoulders 17 behind the arm portions 14 of a projection 11. Along the longitudinal edges of the plates 8 of the first set is a series of alternating projections 11 and recesses 12 which are formed in a manner similar to those on plates 7. When the plates 7 and 8 are joined together the projections 11 on one of the plates interlock with the recesses 12 on the other plate. The shoulders 17 are engaged by the arm portions 14 which thereby serve as retaining means. Separation of the plates 7 and 8 in a direction normal to the longitudinal edges along the plane containing the first set is not possible. Formed along the longitudinal edges of the plate 9 of the second set 6 is a series of projections in the form of tenons 18. The tenons 18 project from the edge by a distance substantially equal to the thickness of the plates 7, 8 of the first set. Similarly, a series of tenons 19 is formed along the longitudinal edges of the plates 10 of the second set 6. As seen in FIG. 4, at a mid-position of the upper surface of the tenon 18 is stepped to form a latch 20. An upwardly and rearwardly inclined portion 21 of the upper surface extends to the latch 20 from a leading end of the tenon 18. A lower surface of the tenon 19 is also stepped at a mid-portion thereof to form a latch 22. A downwardly and rearwardly inclined portion 23 of the lower surface extends from a leading end of the tenon 19 to the latch 22. The tenons 18, 19 are received in rectangular slots 24 formed in the plates 7, 8 of the first set, each slot being located within the periphery defining a projection 11. To assemble the compartments 3, the plates 7, 8 of the first set 5 are joined together by interlocking the projections 11 and recesses 12 along mutually adjacent longitudinal edges of the plates. Several plates 7, 8 may be joined together in this manner to obtain the desired width of plates of the first set 5. An end plate 25 (see FIG. 1) is provided at each end of the plates 7, 8. Along a longitudinal edge of each end plate 25 is a series of projections 11 and recesses 12 in the form as provided on the end plates 7, 8. These projections and recesses interlock with complementary projections and recesses in an adjacent plate 7, 8 of the first set. Rectangular slots 24 are provided within the peripheries defining the projections 11. After assembling the plates 6, 7 of the first set 5, the plates 8, 9 of the second set 6 are interlocked therewith. This is effected by inserting the tenons 18 of a plate 9, and tenons of a plate 8 into the slots 24. The leading end of the tenons 18, 19 are inserted from opposite sides of the plates 7, 8 so that the inclined surfaces 21, 23 slide along one another. The plates 7, 8 are forced together so that latches 20, 22 become interengaged. The latches 20, 22 are positioned so that the plates 8, 9 are clamped against the plates 6, 7 thereby forming a strong interlocking joint at the cross junction of the four plates. An end plate 26 (see FIG. 1) is provided at each end of the plates 9, 10 of the second set 6. Along the longitudinal edge of the plate 26 is a series of tenons 18 which interengage with tenons 19 on a mating plate 10 within a slot 24. A container 31 incorporating an alternative type of interlocking joint according to the present invention is illustrated in FIGS. 5 to 8. The container 31 comprises an elongate cylindrical casing 32 within which is housed a plurality of compartments 33. The compartments 33 are located within the casing 32 by a plurality of baffle plates 34 positioned at intervals along the container. The compartments 33 are of rectangular cross section and are formed by two sets of elongate plates 35, 36. A first set of plates 35 extends perpendicularly with respect to a second set of plates 36 so as to form opposing walls of the compartments 33. Seven compartments 33 have been illustrated in FIG. 5 by way of example, but the number of compartments will be selected to suit a particular application. FIG. 6 illustrates a typical interlocking joint for connecting individual plates in the first set of plates 35 to obtain a wider plate to extend across the container. A plate 35a of the first set has a series of projections 37 arranged at a regular pitch or spacing along a longitudinal edge of the plate. Each projection 37 comprises a first portion 38 extending forwardly from the longitudinal edge of the plate. Two arm portions in the form of resilient tabs 39 extends rearwardly and outwardly from an end of the first portion 38. A series of recesses 40 is provided on the longitudinal edge of the mating plate 35b. The recesses 40 are arranged at spaced intervals so as to receive the projections 37. Each recess 40 has a first passage region 41 extending into an enlarged region 42. A shoulder 43 is formed at the junction of the first passage region 41 and the enlarged region 42. When assembling the plates 35a, 35b, the projections 37 are forced through the first passage region 41. The dimensions of the passage region 41 are such as to cause the two resilient tabs 39 to be deflected inwardly. When the projections 37 are completely located in the recesses 40, the enlarged region 42 allows the resilient tabs 39 to move towards their undeflected positions. The free ends of the tabs 39 locate behind the shoulders 43 which act to retain the projections 37 thereby interlocking the plates 35a, 35b. The strength of the interlocking joint is enhanced by the provision of a series of rectangular extensions 44 along the longitudinal edge of plate 35a which mate with a series of correspondingly shaped rectangular cavities 45 formed in the longitudinal edge of the plate 35b. A typical interlocking joint for interconnecting plates at a T-junction of the first and second set of plates 35, 36 is shown in FIGS. 7 and 8. A series of projections 37 is arranged at a regular pitch along the longitudinal edge of the plate 36. These projections 37 correspond in form to the projections 37 described with reference to the joint shown in FIG. 6. A series of recesses 40 is formed in the plate 35 which extends perpendicularly with respect to the plate 36. Again, the recesses 40 are formed in a similar manner to those described with reference to the joint shown in FIG. 6. In this embodiment, however, the enlarged region 46 is in the form of a slot having semi-circular ends. A series of rectangular extensions 44 is formed along the longitudinal edge of the plate 36. These extensions are received in rectangular slots 47 formed in the plate 36. The plates 35, 36 are interlocked by forcing the projections 37 into the recesses 40 so that the resilient tabs 39 locate behind the shoulders 43 in the manner previously described.
	description	Referring to FIG. 1, a prior art radiation modulating device 100 is shown. The radiation modulating device 100 has a housing 110 and a core area 106. The housing 110 includes a top 102, a bottom 104, and side walls 112. The core area 106 has a plurality of segments 108 generally configured to partially or completely block one or more portions of a radiation beam. As stated above, a problem exist in that one has to frequently change out the radiation modulating device 100 between the radiation source and the target area. This generally is fairly time consuming and causes too much treatment down time. Thus, referring now to FIGS. 2-4, an apparatus 10 which is coupled to a radiation device 12 which will allow the radiation device 12 to alter the amount of radiation delivered to a treatment area is shown. The apparatus 10 has a plate member 14. In the Figures, the plate member 14 is circular in shape. However, this is just given as an example and should not be seen as to limit the scope of the present invention. The plate member 14 will have a plurality of openings 16 located thereon. Each opening 16 is used to house and hold a radiation modulating device 100. As may be seen most clearly in FIG. 3, the openings 16 are located around the periphery of the plate member 14. The plate member 14 is of a sufficient size to have a plurality of radiation modulating device 100 positioned around the periphery thereof. The plate member 14 may have one or more handles 17. The handles 17 may be positioned anywhere on the plate member 14. In the embodiment depicted in FIG. 3, the handles 17 are located on the outer periphery of the plate member 14. However, this should not be seen as to limit the scope of the present invention. The handles 17 are provided to allow one to hold and move the plate member 14. The apparatus 10 has a frame member 18. The frame member 18 is similar in shape to the plate member 14. As shown in FIG. 3, the frame member is semi-circular in shape. However, this should not be seen as to limit the scope of the present invention. The frame member 14 just has to be able to support the plate member 18 in a horizontal position. In the embodiment depicted in the Figures, the frame member 18 fits around approximately half of the outer perimeter of the plate member 14. A channeling 20 runs along an inner circumference of the frame member 18. When the frame member 18 is used to support and hold the plate member 14, the plate member 14 will slide into the channeling 20. Thus, the channeling 20 is also used to support and hold the plate member 14 within the frame member 18. Located within the channeling 20 is a plurality of rollers 21. As may be seen more clearly in FIG. 4, the rollers 21 may be located on a top surface, a bottom surface, and/or a side surface of the channeling 20. The rollers 21 are used to rotate the plate member 14 within the channeling 20. In general, the rollers 21 will have a rubberized surface in order to grip and rotate the plate member 14. However, this should not be seen as to limit the scope of the present invention. Any type of rollers 21 may be used as long as the rollers 21 are able to rotate the plate member 14 within the channeling 20. The rollers 21 may be coupled to a control mechanism. The control mechanism may be a computer or the like. The control mechanism is used to control the movement of the rollers 21. By controlling the movement of the rollers 21, one can rotate the plate member 14 within the channeling 20. Alternatively, one can manually rotate the plate member 14 within the channeling 20. The frame member 18 has a cross bar 22. The cross bar 22 runs from a first end of the frame member 18 to a second end of the frame member 18. The cross bar 22 is used to provide additional support for the frame member 18 so that the frame member 18 can hold and support the plate member 14. A rotatable arm 24 is coupled to the cross bar 22. The rotatable arm 24 is rotatably coupled to a cental area of the cross bar 22 by a peg member 25. The rotatable arm 24 will rotate about the peg member 25 in either a clockwise or counter-clockwise fashion. The rotatable arm 24 is also used to support and hold the plate member 14 in the frame member 18. Located on one end of the rotatable arm 24 is a raised member 26. The raised member 26 is perpendicular to the rotatable arm 24. The raised member 26 is used to secure the plate member 14 within the frame member 18 so that the plate member 14 is free to rotate within the frame member 18 but cannot be removed without rotating the rotatable arm 24 to an open position. A pair of arms 30 extend up from each end of the frame member 18. The pair of arms 30 are used to secure the apparatus 10 to a radiation source 12. In operation, the pair of arms 30 are used to secure the apparatus 10 to a radiation source 40. The rotatable arm 24 will be moved so that a plate member 14 may be inserted into the channeling 20 of the frame member 18. In order to proper align the plate member 14 within the frame member 18, an opening 32 may be formed within the center of the plate member 14. The opening 32 is then positioned over the peg member 25 to ensure that the plate member 14 is proper aligned within the frame member 18. Once the plate member 14 is positioned in the channeling 20 of the frame member 18, the rotatable arm 24 is moved so that the rotatable arm 24 is perpendicular to the cross bar 22. This will lock the plate member 14 within the frame member 18. The plate member 14 is then rotated either by a control mechanism or by hand so that the proper radiation modulating device 100 is in position directly below the radiation beam. The plate member 14 can be rotated so that different radiation modulating device 100 may be used to provide different levels of radiation treatment. While the invention has been particularly shown and described with reference to preferred embodiments thereof, it will be understood by those skilled in the art that the foregoing and other changes in form and details may be made therein without departing from the spirit and scope of the invention.
044951469	abstract	A system and method for the loading of spherical nuclear fuel in vertical fuel rods. The system includes a fuel rod support, a glovebox, a weighing station system, a means for feeding spherical fuel to the fuel rod and transportation means for moving the fuel between parts of the system.
	description	The present invention relates to an X-ray tube and a conditioning method therefor. Conventional X-ray tubes use a filament as a cathode and uses thermoelectrons emitted from the filament as an electron source. On the other hand, there are proposed some X-ray tubes that use a cold cathode source as an electron emission element. Such an X-ray tube is disclosed in, e.g., U.S. Pat. Nos. 7,778,391, 7,809,114, and 7,826,595. However, when a cold cathode source is used as an electron emission source, there is a problem that electron emission is easily affected by the degree of vacuum of an X-ray tube during its operation because the electron emission is sensitive to a surface state of the cathode compared to a hot cathode. Particularly, it is known that in a Spindt-type cold cathode array using a molybdenum (Mo) material, a current decrease occurs due to generation of oxidizing gas in a vacuum tube being in an operating state (see J. Vac. Sci. Technol. B16, 2859 (1998), Effect of O2 on the electron emission characteristics of active molybdenum field emission cathode arrays (B. Chalamala, et al)). Thus, for some situations, there is a problem that decrease in anode current occurs by that the operation of the X-ray tube is conducted continuously. In order to prevent such a problem, a method of gradually increasing extraction voltage is also adopted (see IVNC2013 P15, Stable, High Current Density Carbon Nanotube Field Emission Devices (D. Smith et al), Proc Of SPIE Vol. 7622 76225M-1, Distributed Source X-ray technology for Tomosynthesis imaging (F. Sprender, et al)); in this case, however, a problem such as discharge may occur when the extraction voltage exceeds a predetermined value. The object of the present invention is to provide an X-ray tube and a conditioning method therefor capable of avoiding the above problems. An X-ray tube according to the present invention includes: an electron emission unit including an electron emission element using a cold cathode; an anode unit disposed opposite to the electron emission unit, with which electrons emitted from the electron emission unit collide; and a focus structure disposed between the electron emission unit and a target unit disposed on a surface of the anode unit that is opposed to the electron emission unit. The electron emission unit is divided into first and second regions which can independently be turned ON/OFF. The X-ray tube is focus-designed such that collision regions of electron beams emitted from the respective first and second regions substantially coincide with each other. A conditioning method according to the present invention is a conditioning method for an X-ray tube. The X-ray tube includes: an electron emission unit including an electron emission element using a cold cathode; an anode unit disposed opposite to the electron emission unit, with which electrons emitted from the electron emission unit collide; and a focus structure disposed between the electron emission unit and a target unit disposed on an opposing surface of the anode unit to the electron emission unit. The electron emission unit is divided into first and second regions which can independently be turned ON/OFF. The X-ray tube is focus-designed such that collision regions, at the anode unit, of electron beams emitted from the respective first and second regions substantially coincide with each other. In the conditioning method, one of the first and second regions is used for conditioning and other one of them for actual operation. Preferred embodiments of the present invention will be explained below in detail with reference to the accompanying drawings. The present invention controls and stabilizes a vacuum state in an X-ray tube so as to prevent current variation which occurs in the conventional cold cathode electron tubes during operation. Specifically, there is provided an emitter structure including a plurality of electron beam emission regions, and focus design is made such that an electron beam collides with the same region of the anode while independently controlling the plurality of electron beam emission regions. This allows at least one first electron beam emission region to be used for conditioning to make an electron beam collide with the anode, making it possible to degas the electron beam collision region. At this time, a fixed potential is applied between the gate and the cathode of the second electron beam emission region not used for conditioning so as to turn OFF the second electron beam emission region. The emitter in an OFF state is inactive, so that even when degassing occurs during conditioning, there is a low probability that the surface condition of the emitter varies. After the surface of the anode is sufficiently degassed by conditioning, the second electron beam emission unit which is turned OFF during conditioning is used for actual operation. By making a focus design such that collision regions, at the anode, of the electron beams emitted from the respective first and second electron beam emission regions substantially coincide with each other, it is possible to suppress degassing during actual operation, thereby obtaining stable operation. Hereinafter, first and second embodiments of the present invention will be described successively. FIG. 1 is a cross-sectional view schematically illustrating an X-ray tube 1 according to a first embodiment of the present invention. As illustrated in FIG. 1, the X-ray tube 1 has a structure in which an electron emission unit 10, an anode unit 11, a target unit 12, and a focus structure 13 are disposed in a vacuum area surrounded by a glass outer wall 14. FIG. 1 also illustrates a controller 2 for the X-ray tube 1. The electron emission unit 10 has an electron emission element using a cold cathode and is configured to emit electrons from the cold cathode. While details will be described later, the electron emission unit 10 is divided into two regions A and B (first and second regions). The regions A and B are grounded through transistors TA and TB, respectively. The anode unit 11 is disposed opposite to the electron emission unit 10 and connected to a power supply P. Thus, when either of the transistors TA or TB is turned ON, current flows from the power supply P through the anode unit 11 and electron emission unit 10. At this time, a plurality of electrons are emitted from the electron emission unit 10. These electrons collide with the anode unit 11, pass therethrough, and is absorbed by the power supply P. As illustrated in FIG. 1, a surface 11a of the anode unit 11 that is opposed to the electron emission unit 10 is inclined to the electron moving direction (direction from the left to the right in FIG. 1). The target unit 12 is a member made of a material that generates an X-ray by receiving electrons and disposed on the opposing surface 11a. Since the target unit 12 is disposed on the opposing surface 11a, some or all of the plurality of electrons that collide with the anode unit 11 pass through the target unit 12, and an X-ray is generated in the target unit 12 during the passage. The thus generated X-ray is radiated downward owing to inclination of the opposing surface 11a. The focus structure 13 is a structure having a function of correcting the trajectory of the electron emitted from the electron emission unit 10 and has a window 13a as illustrated in FIG. 1. The electrons emitted from the electron emission unit 10 are directed to the target unit 12 through the window 13a. For example, the window 13a preferably has a circular shape. FIG. 2(1) is a view illustrating a method of dividing the electron emission unit 10 according to the present embodiment. As illustrated, the electron emission unit 10 according to the present embodiment is divided into two regions A and B which are line-symmetrical to each other. More specifically, the electron emission unit 10 according to the present embodiment is formed into a square shape, and the region A is formed by one of the two regions equally divided by a straight line parallel to one side of the square, and the region B is formed by the other one of the two regions. The regions A and B are connected to the controller 2 respectively through the mutually different transistors TA and TB. The controller 2 is configured to independently turn ON/OFF the transistors TA and TB by controlling the gate potentials of the respective transistors TA and TB. Thus, the regions A and B can independently be turned ON/OFF. The ON-state means that the region A or B functions as an electron emitter, that is, a state where electrons are emitted toward the anode unit 11 from the region A or B. On the other hand, the OFF-state means that the region A or B does not function as the electron emitter, that is, a state where electrons are not emitted toward the anode unit 11 from the region A or B. The X-ray tube 1 according to the present embodiment is focus-designed such that a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the region A illustrated in FIG. 2(1) and a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the region B illustrated in FIG. 2(1) substantially coincide with each other. That is, the electron emission unit 10 and the focus structure 13 are configured such that a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the region A illustrated in FIG. 2 (1) and a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the region B illustrated in FIG. 2(1) substantially coincide with each other. Such a configuration can be achieved by disposing the electron emission unit 10 and the focus structure 13 so that the center of the window 13a (having a circular shape, for example) and the center of the square-shaped electron emission unit 10 coincide with each other as viewed in the electron moving direction and by controlling adequately the gate-cathode voltage Vgc (i.e., gate-collector voltage of the respective transistors TA and TB) of the respective regions A and B. The X-ray tube 1 “focus-designed such that the two collision regions substantially coincide with each other” includes one in which the two collision regions do not coincide with each other within the range where the effect of the present invention can be obtained. FIG. 3 is a view explaining the drive state of the X-ray tube 1. As illustrated, the controller 2 performs different controls between during conditioning and during actual operation. Specifically, during conditioning, the controller 2 applies a voltage Vgc of 30 V to 40 V between the gate and the cathode of the region A (i.e., between the gate and the collector of the transistor TA) to turn ON the region A as the emitter of the electrons, while applying a voltage Vgc of 0 V to 10 V (a specific potential in an non-operating state) between the gate and the cathode of the region B (i.e., between gate and collector of the transistor TB) to turn OFF the region B as the emitter of the electrons. As a result, no electron is emitted from the region B, and only electrons emitted from the region A collide with the target unit 12. On the other hand, during actual operation, the controller 2 applies a voltage Vgc of 30 V to 40 V between the gate and the cathode of the region B (i.e., between the gate and the collector of the transistor TB) to turn ON the region B as the emitter of the electrons, while applying a voltage Vgc of 0 V to 10 V (a specific potential in an non-operating state) between the gate and the cathode of the region A (i.e., between the gate and the collector of the transistor TA) to turn OFF the region A as the emitter of the electrons. As a result, no electron is emitted from the region A, and only electrons emitted from the region B collide with the target unit 12. According to the above control method (conditioning method), the electron beam collision regions during conditioning and during actual operation substantially coincide with each other, allowing reduction in degassing amount during actual operation, which in turn reduce current variation in the region B during actual operation. Further, it is possible to reduce a possibility of causing problems due to abnormal discharge or the like during operation. As described above, according to the present embodiment, degassing from the electron beam collision region on the anode unit 11 of the X-ray tube 1 is suppressed to prevent current from varying even in long time operation, thereby allowing stable operation of the X-ray tube 1. Further, it is possible to reduce a probability of causing problems due to the degassing, such as abnormal discharge, allowing the service life of the X-ray tube 1 to be prolonged. Next, the second embodiment of the present invention will be described. The second embodiment differs from the first embodiment in the dividing method of the electron emission unit 10. Other configurations are the same as those in the first embodiment. Hereinafter, a description will be given focusing on differences from the first embodiment with the same reference numerals given to the same elements as in the first embodiment. FIG. 2(2) is a view illustrating the dividing method of the electron emission unit 10 according to the present embodiment. As illustrated, the electron emission unit 10 according to the present embodiment is divided into two or more regions including a center region B and one or more peripheral regions A surrounding the center region B. Specifically, the electron emission unit 10 is formed into a square shape as in the first embodiment, and the region obtained by concentrically overlapping another square having a site slightly smaller than the square of the electron emission unit 10 and having an inclination of 45° with respect thereto is defined as the center region B. Further, each of four regions obtained by removing the center region B from the square-shaped electron emission unit 10 is defined as the peripheral region A. The peripheral region A according to the present embodiment corresponds to each of the peripheral regions A, and the center region B corresponds to the center region B. The X-ray tube 1 according to the present embodiment is focus-designed such that a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the peripheral region A illustrated in FIG. 2(2) and a collision region, the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the center region B illustrated in FIG. 2(2) substantially coincide with each other. That is, the electron emission unit 10 and the focus structure 13 are configured such that a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the peripheral region A illustrated in FIG. 2(2) and a collision region, at the anode unit 11 (region within the opposing surface 11a), of the electron beam emitted from the center region B illustrated in FIG. 2(2) substantially coincide with each other. Such a configuration can be achieved by disposing the electron emission unit 10 and the focus structure 13 so that the center of the window 13a (having a circular shape, for example) and the center of the square-shaped electron emission unit 10 coincide with each other as viewed in the electron moving direction and by controlling adequately the gate-cathode voltage Vgc (i.e., gate-collector voltage of the respective transistors TA and TB) of the respective region A and region B. In the present embodiment as well, the X-ray tube 1 “focus-designed such that the two collision regions substantially coincide with each other” includes one in which the two collision regions do not coincide with each other within the range where the effect of the present invention can be obtained. The operation of the controller 2 in the present embodiment may be the same as the operation described in the first embodiment. That is, when the controller 2 executes the operation described in the first embodiment, the same effects as in the first embodiment can be obtained in the present embodiment. That is, degassing from the electron beam collision region on the anode unit 11 of the X-ray tube 1 is suppressed to prevent current from varying even in long time operation, thereby allowing stable operation of the X-ray tube 1. Further, it is possible to reduce a probability of occurrence of problems due to the degassing, such as abnormal discharge, allowing the service life of the X-ray tube 1 to be prolonged. While the preferred embodiments of the present invention have been described, the present invention is not limited to the above embodiments but may be variously modified within the scope thereof. For example, the specific dividing method of the electron emission unit 10 is not limited to those described in the first and second embodiments. FIG. 2(3) is a view illustrating another example of the dividing method of the electron emission unit 10. The example of FIG. 2(3) is basically the same as that illustrated in FIG. 2(2) but differs therefrom in that the center region A is smaller than that in the example of FIG. 2(2), and that the peripheral region B is a single region. Even in this example, by focus-designing the X-ray tube 1 such that a collision region, at the anode unit 11, of the electron beam emitted from the center region A and a collision region, at the anode unit 11, of the electron beam emitted from the peripheral region B substantially coincide with each other, it is possible to obtain the same effects as in the first and second embodiments. Although the areas of the regions A and B are not particularly mentioned in the second embodiment, the electron emission unit 10 may be divided so that the area of the region A (e.g., total area of one or more peripheral regions) and the area of the region B (e.g., the area of the center region) are substantially equal to each other. By doing this, current of the same amount as that during actual operation can be conveniently taken during the conditioning. Further, compatibility exists between the regions A and B, thus improving usability of the X-ray tube 1.
054405998	claims	1. A nuclear fuel rod spacer grid comprising: a first set of strips which is slottedly interlocked with a second set of strips; a plurality of integral tabs formed on each of said first strips and which are coplanar with said first strips; and side-supported vanes comprising a portion of each said tab which is bent along a predetermined bend line which extends outwardly along each said tab and which is coplanar with the strip on which each said tab is formed. a set of first strips which are each formed with a plurality of first slots which extend from an upper edge thereof; a set of second strips which are each formed with a plurality of second slots which extend from a lower edge thereof, said second strips being interleaved with said first strips so that said first and second slots respectively receive portions of said second and first strips which are contiguous with said second and first slots; a plurality of integral tabs formed exclusively on an upper edge of each of said second strips and which are coplanar with said second strips; and side-supported vanes comprising a portion of each of said tabs which is bent along a predetermined bend line which extends outwardly along each of said tabs and which is coplanar with the strip on which each said tabs is formed. 2. The nuclear fuel rod spacer grid of claim 1, in which the vanes are supported by said tabs in a manner wherein they are spaced from the strips by an amount sufficient that they become fluidly decoupled from the flow restriction caused by said first and second sets of strips. 3. The nuclear fuel rod spacer grid as set forth in claim 2, further comprising a flow control vane which comprises a portion of said integral tab which is bent at a predetermined angle with respect to the portion of said tab which is coplanar with said first strip and along a bend line which is coplanar with said first strip. 4. A nuclear fuel rod spacer grid as set forth in claim 3, wherein said first tabs each assume an essentially triangular shape after the at least one edge portion is bent to form said flow deflecting vane. 5. A nuclear fuel rod spacer grid comprising: 6. A nuclear fuel rod spacer grid as set forth in claim 5, further comprising a plurality of essentially inverted U-shaped apertures which are formed in each of said first strips, each of said essentially inverted U-shaped apertures being formed in said strips so as to proximate a base portion of a tab and so dimensioned and located as to facilitate welding of said first and second strips.
050739140	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a stereoscopic X-ray apparatus for irradiating X-rays from right and left portions in two directions to intersect at a to-be-examined object by using a stereoscopic X-ray tube or the like and using X-ray images obtained in the above two directions so as to permit the object to be stereoscopically observed. 2. Description of the Related Art As an example of a conventional stereoscopic X-ray apparatus, an X-ray apparatus is known in which X-rays are sequentially irradiated from two foci of a stereoscopic X-ray tube and two X-ray images obtained by two X-ray irradiations are photographed on two X-ray films. The X-ray films after being subjected to development are placed side by side on a film viewer and the object may be stereoscopically observed by using the parallax between the two images. In recent years, digital fluorography is widely used, and in this case, an X-ray image having passed an object to be examined is picked up by means of an image intensifier tube and TV camera and picked-up images created by X-rays from the right and left foci are stored into an image memory. Then, the picked-up images created by X-rays from the right and left foci and read out from the image memory are displayed on two display units. Also, in this case, the object can be stereoscopically observed in the same manner as in the case of using the films. The stereoscopic display method in the digital fluorography includes a dynamic stereoscopic display method which requires use of field switcher glasses with electronic shutters and in which right and left images are alternately displayed on a single display unit and the electronic shutters are alternately driven in synchronism with alternate display of the images in addition to a static stereoscopic display method in which two images are simultaneously displayed on the two display units as described above. In the above conventional cases, since the stereoscopy is determined by the parallax between the images, it can be determined by a distance between the foci of the stereoscopic X-ray tube, which is generally a fixed value, a distance between the image intensifier tube (or film) and the X-ray tube, and a distance between the image intensifier tube (or film) and the object. The latter two distances change according to the image pick-up condition (geometrical relation among the X-ray tube, object, and image intensifier tube). Therefore, the stereoscopy is determined by the image pick-up condition and it will be changed according to the image pick-up condition. Further, since the sensitivity to the stereoscopy may be different for different persons, it sometimes happen that the stereoscopy of the same stereoscopic display image is sufficiently high for one person but may be low for another. In this case, if the stereoscopy is low, the positional relation in the depth direction between two things may be erroneously recognized. SUMMARY OF THE INVENTION An object of the present invention is to provide a stereoscopic X-ray apparatus capable of adjusting the stereoscopy at the time of stereoscopic observation after imaging an object to be examined irrespective of the imaging condition (positional relation among an X-ray tube, object, and image intensifier tube) so as to provide a desired stereoscopy under any imaging condition. A stereoscopic X-ray apparatus according to the present invention comprises means for irradiating X-rays in two directions to an object to be examined; means for picking-up images created by X-rays having passed the object to provide X-ray images picked-up in the two directions; means for detecting the positional relation of the object with respect to the irradiation means and picking-up means; and means for adjusting the positional relation between two X-ray images output from the picking-up means according to the positional relation of the object detected by the detection means and then stereoscopically displaying the X-ray images. According to the present invention, a distance between display positions of the two X-ray images output from the picking-up means is adjusted to a distance corresponding to desired stereoscopy and then the X-ray images are displayed in a stereoscopic manner so that desired stereoscopy can always be attained irrespective of the imaging condition. Additional objects and advantages of the present invention will be set forth in the description which follows, and in part will be obvious from the description, or may be learned by practice of the invention. The objects and advantages of the present invention may be realized and obtained by means of the instrumentalities and combinations particularly pointed out in the appended claims.
	summary
	description	1. Field of the Invention This invention relates generally to nuclear reactors and more particularly, to assemblies and methods for coupling core spray line assemblies within such reactors in a repair. 2. Description of Related Art A reactor pressure vessel (RPV) of a boiling water reactor (BWR) typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A core shroud or shroud typically surrounds the core and is supported by a shroud support structure. Boiling water reactors have numerous piping systems, and such piping systems may be utilized, for example, to transport water throughout the RPV. For example, core spray piping may be used to deliver water from outside the RPV to core spargers inside the RPV and to cool the core. The core spray piping may be coupled to a thermal sleeve which may be slip fit into a RPV nozzle safe end. Stress corrosion cracking (SCC) is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners, and welds which are exposed to high temperature water. The reactor components may be subject to a variety of stresses. These stresses may be associated with, for example, differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other stress sources, such as residual stresses from welding, cold working and other inhomogeneous metal treatments. In addition, water chemistry, welding, heat treatment and radiation can influence the susceptibility of metal in a component to SCC. Reactor internal piping, such as thermal sleeves and core spray lines, may occasionally require replacement as a result of failure due to SCC. Replacing the core spray piping typically may include removing the core thermal sleeve from the RPV nozzle safe end. In the event a safe end requires replacement, the reactor must be shut down for maintenance and drained to an elevation below that of the safe end. The safe end is then removed and a replacement safe end is welded to the RPV nozzle. Thereafter, a replacement core spray line (external to the reactor) may be welded to the replacement safe end. Replacing a safe end is typically time consuming and costly, since such replacement generally requires a lengthy reactor outage of several days to a week or more. It would be desirable to provide an assembly which facilitates replacing core spray lines without removing the reactor pressure vessel safe end. It also would be desirable to provide such an assembly which is easily removed and installed without the necessity of welding. An exemplary embodiment of the present invention is directed to a core spray T-box attachment assembly for a core spray nozzle. The assembly may include a primary cruciform wedge and a secondary cruciform wedge in contact with the primary cruciform wedge to form a cruciform wedge subassembly adapted for insertion within a bore of the core spray nozzle to sealingly engage an interior converging portion of a safe end of the core spray nozzle. The assembly may include a spider in contact with the cruciform wedge subassembly, and a draw bolt for engaging an axial bore of a center portion of the cruciform wedge subassembly and the spider to the T-box. Another exemplary embodiment of the present invention is directed to a method of replacing a subassembly (having a T-box and thermal sleeve) within a core spray nozzle of a nuclear reactor. The method may include removing the T-box and thermal sleeve and machining the safe end. Replacement hardware may be inserted therein. The replacement hardware may be configured to create a seal against a converging inner surface of the safe end of a core spray nozzle. Another exemplary embodiment of the present invention is directed to an attachment assembly. The attachment assembly may include a hollow wedge having a plurality of parts. Each part may be configured to contact a first surface, the attachment assembly may also include a first component configured to pull each of the plurality of parts in a first direction against the first surface; and a second component configured to push against a second surface in a second direction, while pulling the first component in the first direction. FIG. 1 is a top plan view of a boiling water nuclear reactor pressure vessel (RPV) 10. RPV 10 includes a vessel wall 12 and a shroud 14 which surrounds the reactor core (not shown) of RPV 10. An annulus 16 may be formed between vessel wall 12 and shroud 14. The space inside annulus 16 may be limited, as most reactor support piping may be located within annulus 16. In the event of a reactor plant casualty, such as a loss of coolant accident, cooling water is delivered to the reactor core through core spray distribution header pipes 18 and 20, which are connected to respective downcomer pipes 22 and 24. Downcomer pipes 22 and 24 are connected to shroud 14 through respective lower T-boxes 26 and 28, which are attached to shroud 14 and internal spargers 30. FIG. 2 is a top sectional view of a T-box attachment assembly in accordance with an exemplary embodiment of the invention. Distribution header pipes 18 and 20 may diverge from an upper T-box attachment assembly 32. Particularly, T-box attachment assembly 32 may include, in one example, a T-box housing 34 having first, second, and third ends 36, 38 and 40 as shown in FIG. 2. First end 36 of T-box housing 34 is coupled to a safe end 42 of core spray nozzle 44 by a thermal sleeve 50 (shown in FIG. 2). Ends 38 and 40 are configured to be in substantial alignment and configured to couple to core spray distribution header pipes 18 and 20 respectively. Header pipes 18 and 20 are coupled to second and third ends 38 and 40 by pipe connectors 46 and 48 respectively. Pipe connectors 46 and 48 may be any pipe connectors known in the art, for example, ball flange connectors. FIG. 3 is a detailed view of a portion of the T-box attachment assembly in accordance with an exemplary embodiment of the invention. FIG. 4 is an exploded view of the T-box attachment assembly in accordance with an exemplary embodiment of the invention. Referring primarily to FIG. 3 (with occasional reference to FIG. 2), the T-box attachment assembly 32 includes in addition to T-box housing 34, a thermal sleeve 50, a spider 52, a cruciform wedge subassembly 53, a draw bolt 59, a draw bolt nut 58, and a nut keeper 60. First end 36 of T-box housing 34 is welded to a first end 62 of thermal sleeve 50. T-box housing 34 is configured to be positioned so that thermal sleeve 50 is located inside core spray nozzle 44 and is fitted to the inside of the spray nozzle safe end 42. For example, safe end 42 includes a safe end bore 68 extending through the safe end 42. Safe end bore 68 includes an inside surface 120 with a converging tapered portion 70 and a diverging tapered portion 71. A second end 72 of thermal sleeve 50 is positioned within core spray nozzle safe end 42. An inner surface of thermal sleeve 50 may include threads 114 at second end 72. T-box housing 34 may also include a cover opening (not shown for clarity) that is in substantial alignment with first end 36, and is configured to receive a T-box cover plate 82. Referring to FIG. 4, spider 52 may include a cylindrical shell that includes external threads 124 sized to threadedly engage internal threads 114 of thermal sleeve 50. Vanes extend from an inside surface of the spider shell to a spider center member. The spider 52 center member includes an axial spider bore 138 extending through the spider 52. The spider bore 138 may be sized to receive draw bolt 59. The spider 52 may include a tongue or groove to interface with a tongue or groove of the cruciform wedge subassembly 53 to form a tongue and groove joint 75, as generally shown in FIG. 3., for example. Cruciform wedge subassembly 53 may comprise a plurality of components. For example, the cruciform wedge subassembly 53 may include a primary cruciform wedge 54 and a secondary cruciform wedge 55. The primary cruciform wedge 54 may include a first support member 154a that extends between two web members 57 and a second support member 154b that also extends between two web members 57. The secondary cruciform wedge 55 may include a third support member 155a that extends between two web members 57 and a fourth support member 155b that also extends between two web members 57. The primary cruciform wedge 54 and the secondary cruciform wedge 55 may be joined to make the cruciform wedge subassembly 53. Cruciform wedge subassembly 53 includes a central member having a cruciform central member bore 136 extending, therethrough. The web members 57 of the primary cruciform wedge 54 and the secondary cruciform wedge 55 may be joined together to extend from the central member to form an “X” shaped configuration, for example. The support members 154a, 154b, 155a and 155b may be joined together to form a contiguous substantially circular support member. The support members 154a, 154b, 155a and 155b may be tapered to engage inside surface 120 of the nozzle safe end bore 68 tapered portion 70. The engagement of the support members 154a, 154b, 155a and 155b of the cruciform wedge subassembly 53 against the inside surface 120 of the nozzle safe end bore 68 tapered portion 70 may function as a mechanical seal to minimize leakage, for example. Additionally, the web members 57 are contoured to minimize flow resistance. Moreover, cruciform wedge subassembly 53 may include a tongue or groove to interface with a tongue or groove of the spider 52 to form the tongue and groove joint 75. Referring to FIG. 4, a draw bolt 59 may extend through the cruciform central member bore 136 and the spider bore 138. Draw bolt 59 may include a head portion 158 located at a first end. Head portion 158 may be larger than the diameter of the cruciform central member bore 136, and may be conical-shaped to substantially reduce or possibly minimize the conical shape is merely one example, other shapes which would reduce their resistance would be evident to those skilled in the art. A second end of draw bolt 59 is threaded to threadenly engage draw bolt nut 58 (see also FIG. 2). An adjacent section of the draw bolt 59 may have a hexagonal contour that interfaces with a mating hexagonal shaped bore of nut keeper 60 to prevent draw bolt nut 58 from loosening. The nut keeper 60 may be crimped to the outer surface of the draw bolt nut 58, for example. To replace a core spray line in a nuclear reactor pressure vessel 10, the existing T-box/thermal sleeve combination is removed from the core spray nozzle safe end 42 by any suitable method. The T-box attachment assembly 32 may also be removed from core spray distribution header pipes 18 and 20 (also referred to as “core spray liner”) for example, by roll cutting, conventional underwater plasma arc cutting, and/or electric discharge machining (EDM). A new T-box attachment assembly 32 may be used to connect 18 and 20 to safe end 42 of core spray nozzle 44 by coupling first end 36 of T-box housing 34 to safe end 42 with thermal sleeve 50 and coupling ends 38 and 40 to core spray distribution header pipes 18 and 20. Spider 52 is attached to the second end 72 of thermal sleeve 50 by threadedly engaging spider external threads 124 with thermal sleeve internal threads 114. This threaded connection may provide for ease of fabrication and a means of adjusting the total length of the T-box attachment assembly 32. Once in place, the length of the T-box attachment assembly 32 maybe maintained by installing a dowel pin 168 in the spider 52 and thermal sleeve 50 to prevent relative rotation, as seen in FIG. 3, for example. Draw bolt 59 is then inserted through the cruciform central member bore 136 of the primary and secondary cruciform wedges 54, 55 and the spider bore 138 (not shown in FIG. 4 for reasons of clarity) with threaded end of draw bolt 59 extending away from safe end 42 and towards T-box housing 34. This may be accomplished by attaching a stainless steel cable or wire rope, of about 3 to 5 millimeters in diameter, (not shown) to the threaded end of bolt 59, and threading the cable through the cruciform central member bore 136 of the primary and secondary cruciform wedges 54, 55 and the spider bore 138 before inserting the primary and secondary cruciform wedges 54, 55 and draw bolt 59 into the safe end 42. The primary and secondary cruciform wedges 54, 55 and draw bolt 59 may then be inserted into safe end bore 68. The primary and secondary cruciform wedges 54 and 55 are sequentially inserted in an orientation that positions the axis of the cruciform central member bore 136 of the primary and secondary cruciform wedges 54, 55 perpendicular to the axis of the safe end bore 68 of nozzle safe end 42. Primary and secondary cruciform wedges 54, 55 are then tilted so as to move cruciform central member bore 136 into co-axial alignment with the safe end bore 68. The primary and secondary cruciform wedges may then be assembled to form the cruciform wedge subassembly 53 and support members 154a, 154b, 155a and 155b may be pulled to engage the tapered portion 70 of the safe end bore 68. After the cruciform wedge subassembly 53 has been oriented to its operational position, the wire may be pulled through the cruciform central member bore 136 of the cruciform wedge subassembly 53, which in turn pulls the threaded end of draw bolt 59 through cruciform central member bore 136 and the spider bore 138 into position. The head portion 158 of draw bolt 59 may then engage the cruciform wedge subassembly 53. Draw bolt 59 may be tensioned to fix the cruciform wedge subassembly 53 against the spider in tongue and groove 75. Draw bolt nut 58 is then tightened and nut keeper 60 may be crimped to draw bolt nut 58 to prevent loosening. Consequently, the cruciform wedge subassembly 53 may be pulled tight against the spider 52. Keeper 60 interfaces with the hexagonal section of draw bolt 59 to prevent rotation of the draw bolt nut 58 relative to draw bolt 59. The positioning of the cruciform wedge assembly 57 and the manipulation of the draw bolt 59 may be accomplished through an access 90 in the T-box attachment assembly 32. T-box cover plate 82 is then inserted to cover the access 90. Core spray distribution header pipes 18 and 20 may then be coupled to ends 38 and 40 of the T-box housing 34. First end 36 may be welded to the first end 62 of the thermal sleeve 50, in order to couple first end 36 of the T-box housing 34 to safe end 42, Jack bolt clamp assemblies 25 may be attached to ends of the T-box attachment assembly 32 to complete the installation. The jack bolt clamp assemblies 25 may be adjusted to push against an inner surface of the vessel wall 12 in a first direction and pull, in a second direction, the T-box attachment assembly 32 toward the center of the RPV 10. While clamp assemblies 25 may be used to pull the T-box attachment assembly 32, a spreader or wedge may also be used to cause a similar pull to occur. This pulling action helps create a seal between the cruciform wedge subassembly 53 and the converging, inside surface 120 of the nozzle safe end bore 68 tapered portion 70, for example. The above described T-box attachment assembly 32 may facilitate replacing core spray distribution header pipes 18 and 20 without removing core spray nozzle safe end 42 or draining RPV 10. In addition T-box attachment assembly 32 may facilitate attaching core spray distribution header pipes 18 and 20 to safe end 42 without welding. While the invention has been described in terms of various exemplary embodiments, those skilled in the art will recognize that the exemplary embodiments of the present invention can be practiced with modification within the spirit and scope of the claims.
043476228	claims	1. A method for surveillance of nuclear fuel material in an elongated fuel material containing container, wherein said fuel material includes distributed ferromagnetic particles, comprising the steps of: (1) providing apparatus for scanning said container to produce susceptibility change signals indicative of changes in magnetic susceptibility of said material along the length of said container due to said ferromagnetic particles; (2) performing a first scan along the length of said container at a given time in its life to produce a first susceptibility change signal as a first signature of said fuel material containing container; (3) recording said first signature; (4) performing a second scan along the length of said container at a subsequent time in its life to produce a second susceptibility change signal as a second signature of said fuel material containing container; and (5) comparing said second signature to said first signature whereby similarity of the signatures indicates absence of tampering. (1) providing apparatus for scanning said container to produce susceptibility change signals indicative of changes in magnetic susceptibility of said material along the length of said container and including means for subtracting the contribution of said paramagnetic additive to said susceptibility change signals; (2) performing a first scan along the length of said container at a given time in its life to produce a first susceptibility change signal due to said ferromagnetic particles as a first signature of said fuel material containing container; (3) recording said first signature; (4) performing a second scan along the length of said container at a subsequent time in its life to produce a second susceptibility change signal due to the ferromagnetic particles in the fuel material therein as a second signature of said fuel material containing container; and (5) comparing said second signature to said first signature whereby similarity of the signatures indicates absence of tampering. (1) establishing a direct current magnetic field having a strength of at least 1000 Gauss; (2) disposing inductive means in said magnetic field for producing signals indicative of changes in direct current susceptibility of material moved adjacent thereto; (3) performing a scan along the length of said container at a given time in its life by moving said container adjacent said inductive means at a given velocity whereby said inductive means produces a signal varying in amplitude with time in accordance with the distribution of said ferromagnetic particles in said fuel material; (4) recording said signal as a first signature of said fuel material containing container; (5) providing a second signature of said fuel material containing container at a subsequent time in its life by repetition of steps (1)-(3); and (6) comparing said second signature to said first signature whereby similarity of the signatures is indicative of the absence of tampering with said fuel material. (1) establishing first and second direct current magnetic fields of different field strengths greater than about 1000 Gauss; (2) disposing in each magnetic field respective first and second inductive means for producing signals indicative of changes in susceptibility of material moved adjacent thereto; (3) performing a scan along the length of said container at a given time in its life by moving said container adjacent said first and second inductive means in sequence whereby each of said inductive means produces a signal varying in amplitude with time in accordance with changes in magnetic susceptibility of said material; (4) processing the signals from said first and second inductive means to subtract the contribution of said paramagnetic additive to said signals and to provide a resulting signal varying in amplitude with time in accordance with the distribution of said ferromagnetic particles in said material as a first signature of said fuel material containing container; (5) recording said first signature of said fuel material containing container; (6) providing a second signature of said fuel material containing container at a subsequent time in its life by repeating steps (1)-(4); and (7) comparing said second signature to said first signature whereby similarity of the signatures is indicative of the absence of tampering with said fuel material. (1) providing apparatus for scanning said container to produce susceptibility change signals indicative of changes in magnetic susceptibility of said material along the length of said container due to the distribution of said ferromagnetic particles; (2) performing a first scan along the length of said container at a given time in its life to produce a first susceptibility change signal as a signature of said fuel material containing container; (3) recording said first signature; and (4) performing at least one subsequent scan along the length of said container at subsequent times in its left to produce susceptibility change signals as subsequent signatures of said fuel material containing container whereby significant differences in a subsequent signature from said first signature are indicative of tampering with said material. (1) providing apparatus for scanning said body to produce susceptibility change signals indicative of changes in magnetic susceptibility of said material along the length of said body due to the distribution of said ferromagnetic particles; (2) performing a first scan along the length of said body at a given time in its life to produce a first susceptibility change signal as a signature of said body; (3) recording said first signature; (4) performing at least one subsequent scan along the length of said body at subsequent times in its life to produce susceptibility change signal as subsequent signatures of said body whereby significant differences in a subsequent signature from said first signature are indicative of tampering with said material. 2. The method of claim 1 wherein said fuel material containing container is a fuel bundle. 3. The method of claim 1 wherein said fuel material containing container is a fuel element. 4. A method for surveillance of nuclear fuel material in an elongated fuel material container, wherein said fuel material includes distributed ferromagnetic particles and a paramagnetic additive varying in amount along the length of the container, comprising the steps of: 5. The method of claim 4 wherein said fuel material containing container is a fuel bundle. 6. The method of claim 4 wherein said fuel material containing container is a fuel element. 7. A method for surveillance of nuclear fuel material in an elongated non-magnetic fuel material containing container, wherein said fuel material includes distributed ferromagnetic particles, comprising the steps of: 8. The method of claim 7 wherein said fuel material containing container is a fuel bundle. 9. The method of claim 7 wherein said fuel material containing container is a fuel element. 10. The method of claim 7 wherein said inductive means comprises a sensing coil and said container is moved through said sensing coil. 11. The method of claim 10 wherein the signal produced by said sensing coil is amplified and recorded with respect to time of movement of said container through said coil. 12. The method of claim 7 wherein said magnetic field has a strength in the range of 1000 to 100,000 Gauss. 13. The method of claim 12 wherein said magnetic field is established by a superconductive magnet. 14. The method of claim 7 wherein said velocity is from about 1 to about 100 feet per minute. 15. The method of claim 7 wherein said fuel material is a compound selected from the group consisting of oxides of uranium, plutonium, thorium and mixtures thereof. 16. A method for surveillance of nuclear fuel material in an elongated, non-magnetic fuel material containing container, wherein said fuel material includes distributed ferromagnetic particles and a paramagnetic additive varying in amount along the length of the container, comprising the steps of; 17. The method of claim 16 wherein said fuel material containing container is a fuel bundle. 18. The method of claim 16 wherein said fuel material containing container is a fuel element. 19. The method of claim 16 wherein said processing includes adjusting the magnitude of the signals from said first and second inductive means in accordance with respective constants K.sub.3 and K.sub.4 and determining the sum of the magnitudes of the adjusted first and second signals from the same successive incremental portions of said container to provide said resulting signal. 20. The method of claim 19 wherein K.sub.3 and K.sub.4 are constants of the scanning system determined by scanning at least two containers containing material having known different ferromagnetic content. 21. A method for surveillance of a paramagnetic material in an elongated material containing container wherein said material includes ferromagnetic particles distributed therein, comprising the steps of: 22. A method for surveillance of a body of non-ferromagnetic material which includes ferromagnetic particles distributed therein, comprising the steps of:
	abstract	According to one embodiment, a data detection system for an X-ray CT apparatus includes a data acquisition circuit and a connection structure. The data acquisition circuit includes at least one row of X-ray detection elements arrayed in a channel direction. The data acquisition circuit is configured to acquire data required for generating X-ray CT image data corresponding to the at least one row of the X-ray detection elements. The connection structure is configured to connect the data acquisition circuit with another data acquisition circuit directly or indirectly in a row direction.
050733354	claims	1. A recirculation system for driving reactor coolant water in an annular downcomer defined between a boiling water nuclear reactor vessel and a core shroud spaced radially inwardly therefrom comprising: means for supplying feedwater to said vessel; and a turbopump disposed inside said downcomer and including: a feedwater pump disposed in flow communication with said steam turbine for receiving condensed steam therefrom for forming said feedwater under pressure; a variable first control valve disposed in flow communication between said feedwater pump and said turbopump for selectively regulating said feedwater channeled from said feedwater pump to said turbopump for controlling recirculation of said coolant water being pumped by said turbopump; and a controller for controlling said first control valve. a variable second control valve disposed in flow communication between said feedwater pump and said sparger for selectively regulating said feedwater channeled from said feedwater pump to said feedwater sparger, said second control valve being disposed in parallel flow with said first control valve so that said feedwater from said feedwater pump is split between said feedwater sparger and said turbopump; and said controller is effective for controlling both said first and second control valves inversely relative to each other. 2. A recirculation system according to claim 1 wherein said impeller includes a plurality of circumferentially spaced mixed flow impeller blades, and said turbine blades are fixedly joined to said impeller blades at said impeller outlet end. 3. A recirculation system according to claim 2 wherein said turbopump further includes a plurality of circumferentially spaced stationary nozzle vanes joined in flow communication between said plenum and said turbine blades for channeling said feedwater to said turbine blades for rotating said impeller. 4. A recirculation system according to claim 3 wherein said turbine blades are disposed in flow communication with said outlet guide vanes for discharging said feedwater from said blades to mix with said discharged coolant water from said impeller. 5. A recirculation system according to claim 1 wherein said impeller further includes a radially inner cylindrical surface defining with said axle a radial bearing, and a radially extending aft surface defining with said axle an axial thrust bearing; and said impeller lubricating means is effective for channeling a portion of said feedwater from said plenum to both said radial and thrust bearings for supporting said impeller on said axle upon rotation of said impeller and for providing lubrication therebetween. 6. A recirculation system according to claim 5 wherein said impeller lubricating means includes a conduit extending in flow communication from said plenum to both said radial and thrust bearings for providing said feedwater thereto. 7. A recirculation system according to claim 1 wherein said turbopump has a longitudinal centerline axis disposed parallel to a longitudinal centerline axis of said vessel, and said inlet guide vanes and said outlet guide vanes are axially spaced from each other for driving said coolant water generally parallel to said vessel centerline axis within said downcomer. 8. A recirculation system according to claim 7 further including an annular pump deck extending radially outwardly from said vessel centerline axis and between said vessel and said core shroud for fixedly supporting said turbopump. 9. A recirculation system according to claim 8 wherein said vessel includes a reactor core disposed inside said core shroud, and said turbopump is disposed axially above said reactor core. 10. A recirculation system according to claim 1 further including a steam turbine joined in flow communication with said reactor vessel for receiving steam for driving said steam turbine; and said feedwater supplying means includes: 11. A recirculation system according to claim 10 wherein said reactor vessel further includes a feedwater sparger; and said feedwater supplying means further includes:
	abstract	Methods and apparatus for separating ions of metallic elements are provided. Preferred methods utilize a hydrophobic chelating extractant, such as an organophosphorus compound, adsorbed onto carbon or graphite fibers in the form of felt. Also described is a new thallium-201 generator that comprises a column containing an acidic organophosphorus extractant adsorbed on carbon or graphite fibers, and a yttrium-90 generator system comprised of two extraction columns designed to selectively absorb yttrium-90 at different pH, to enable the separation of yttrium-90 from strontium-90. The two columns are connected in series for stepwise separation. The yttrium-90 product is freed from residual strontium-90 and metal contaminants and can be eluted from the second column with dilute acid, acetate buffer, water or saline for labeling biological targeted molecules. The new generator system provides rapid and efficient separation of yttrium-90 and is amenable to both scale-up and automation. Also described is a new 99mTc generator that comprises a column containing an acidic organophosphorus extractant adsorbed on carbon or graphite fibers designed to selectively absorb 99Mo at a selected pH, to enable the separation of 99mTc from 99Mo.
	summary
046817280	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The apparatus shown in the drawings is a nuclear reactor 31 such as the Mechanical Moderator Controlled Reactor. This reactor includes a pressure vessel 33 consisting of a circular cylindrical body 35 having a spherical bottom and a dome-shaped top 37. The body has a plurality of inlet nozzles 38 and outlet nozzles 40. The body and the top have flanges 39 and 41 which, when the reactor is in use, are engaged and sealed. The reactor has a nuclear core 43 including a plurality of fuel assemblies 45. The fuel assemblies 45 are mounted between a lower core-support plate 47 and an upper core-support plate 49 in the lower part of the pressure vessel 33. The core 43 is encircled by a core barrel 50 which is mounted on flange 39 at the upper end and is secured to the lower core-support plate 47 at the lower end. Clusters 51 of control rods and gray rods 53 (FIGS. 2, 3, 19) and clusters 55 of water-displacement rods 57 (FIGS. 4, 19) are moveable in and out of the fuel assemblies 45 by drive rods 58 (FIG. 1) which are actuable by a control (not shown) external to the reactor. Each cluster 51 typically carries eight control rods or gray rods. Each cluster 55 typically carries forty water-displacement rods. Each cluster 51 of control rods or gray rods 53 is associated with a corresponding fuel assembly 45 and, when actuated by its associated drive rod, its control rods are moved in or out of the corresponding assembly. Each WDRC 55 is associated with a plurality of fuel assemblies positioned peripherally with respect to the axis of the cluster and symmetrical with respect to the axis except at the periphery of the reactor. When the WDRC is actuated by its drive rod, the water-displacement rods which it carries are moved in or out of these assemblies 45. Each control-rod or gray-rod cluster 51 is of cruciform shape. It includes a central sleeve 61 (FIGS. 3, 19) from whose periphery vanes 63 spaced at 90.degree. from each other, to define quadrants between them, extend. A control rod or gray rod 53 is suspended from the end of each vane 63 and a control rod or gray rod is suspended from a position intermediate the end of the vane and the sleeve 61. Typically, the distance between the center of each rod suspended from the end of a vane 63 and the center of the intermediate rod 53 on the same vane is equal to the distance between the center of this intermediate rod and the axis of the sleeve 61. The sleeve 61 is connected to a drive rod 58 (FIG. 1). The water-displacement-rod cluster 55 (FIGS. 4, 19) includes a central sleeve 71 connected to a drive rod 58. Vanes 73 and 75 radiate from the outer surface of the sleeve 71. The vanes 73 and also the vanes 75 are, respectively, spaced angularly by 90.degree., each pair of consecutive vanes 73 defining a quadrant which is bisected by a vane 75. Each vane 73 and 75 terminates in a crucifix having a cross member 77 and 79, respectively. Typically, each vane 73 carries three water-displacement rods 57 and one labelled 57a with their centers spaced equally along vane 73 between the center of the sleeve 71 and the center of the end rod 57. The rod 57a just inwardly of the end rod is centered at the intersection of the cross vane 77 and the vane 73. In addition, the cross vane 77 typically carries a pair of rods 57b at its ends. The spacing between the centers of the rods 57a and 57b on each side is typically equal to the spacing between the centers of the rods 57, 57a along the vane 73. The rods 57c and 57d on each vane 75 are in a crucifix configuration. There are rods 57c along each vane 75, one at the end of the vane and another inwardly. The centers of rods 57c are typically spaced equally from the center of the intersection of each vane 75 and each cross vane 79. The cross vane 79 typically carries rods 57d at its ends. The spacing of the centers of rods 57d from the center of the intersection of the cross vane 79 and the vane 75 is equal to the corresponding spacing for rods 57b. The centers of rods 57c and 57d thus define the corners of a square. The reactor 31 has upper internals 81 (FIG. 1) including guide means for the neutron-absorber rod clusters 51 and 55. The guide means includes columnar guides of cruciform transverse cross-section for the control-rod and gray-rod clusters 51 and plates 85 for the water-displacement-rod clusters 55. Each columnar guide is formed of a plurality of coextensive vertical columnar sections generally labelled 87 to, or between, which the plates 85 are secured. In the interest of facilitating the description of this invention, these sections are designated 87a, 87b, 87c, 87d, 87e, 87f (see FIG. 1 and particularly FIG. 11). The plates 85 are generally horizontal. A plurality of plates 85 are arranged vertically, together serving as guides for the WDRC's. In the interest of precluding the necessity of providing an excessively thick plate 85 and facilitating the handling of the plate guides during installation and maintenance, each plate 85 is formed of a plurality of plate sections nested together in the manner of a "jig-saw puzzle". The different sections are shown nested in FIG. 5. There are five basic plate sections: a square section 91 (FIG. 6); a rectangular section 93 (FIG. 7); i.e., with adjacent sides of unequal length; a lower trapezoidal section 95 (FIG. 8) as viewed in FIG. 5, the left trapezoidal section 97 (FIG. 9) as viewed in FIG. 5, and a triangular section 99. The plate sections are perforated to pass and guide the control-rod and gray-rod clusters 51 between their guide sections 87 and the WDRC's 55 and also to pass the coolant. Perforations 101 of cruciform configuration are distributed uniformly throughout each plate 85. Except at the periphery of the plate, each four of these perforations 101 define the corners of a square. The perforations 101 are included as a whole in each of the sections 91-99. These cruciform perforations 101 pass the control-rod or gray-rod clusters 51. The hole 103 in the center of each perforation 101 passes the sleeve 61 (FIG. 3). The plate 85 also includes perforations 105 and 107 in uniformly-distributed square patterns 109a and 109b. The corners of these square patterns extend into the quadrants defined between the arms of the cruciform perforations. Some of these square patterns 109a are included wholly in the plate sections 91-97, others 109b are formed by the nesting of adjacent sections. For example, the pattern designated by the number 109b in the lower part of FIG. 5 is formed by the nesting of the rectangular section 93 in the lower right-hand corner and the trapezoidal section 99. Each perforation 105 includes a central circular opening 111 dimensioned to pass the stem 71 (FIG. 4) of the WDRC. Perforations 113 and 115 radiate from the periphery of opening 111. Perforations 113 are shaped and dimensioned to pass the vanes 73 and 77 of a WDRC and its associated rods 57, 57a, 57b and perforations 115 are shaped and dimensioned to pass the vanes 75 and 79 of a WDRC and their associated rods 53, 53c, 53d. Holes 107 serve to pass the coolant. FIGS. 11 through 19 show the manner in which the columnar guide sections 87 and the separate plate sections 91 through 99 are assembled into integrated guide units. Specifically, FIGS. 11 through 19 disclose the assembly of square plate sections 91 with columnar section 87a through 87f into an integrated unit 110. The other plate sections 93 through 99 are also integrated with columnar guide sections. Their integration will be understood from the following description as to the integration of the square plate sections 91 with sections 87. In the unit 110, each section 87a through 87f includes a plurality (five in the specific case shown) of open-ended enclosures or cans 121a through 121f, designated generally 121 (FIG. 3). Each can has a cruciform transverse cross-section with openings dimensioned the same as the cruciform perforations 101 (FIGS. 5-8). At its lower end, each can 121a is welded to a frame-like open-end plate 123 whose boundaries are intersecting parallel strips 125 (FIG. 13). Extending through the strips symmetrically on both sides of each intersection are cruciform perforations 126 (FIGS. 13, 14) congruent with the perforations 101. Each of these cruciform perforations 126 is centered or symmetrical about an axis 128 through the center of the intersections of the strips 125. These axes 128 define the corners of a square. The cans 121a are welded so that each can envelops a cruciform perforation 126 at an intersection of strips 125 and is also centered on the corresponding axis 128. The end plate 123 is aligned by pins 127 which engage the upper core plate 49 (FIG. 1) thus holding the lower end of the columnar guide formed of the columnar sections 87a through 87f in lateral alignment. When the rods are inserted in the core 43, the vanes 63, 73, 75, 77 and 79 are advanced to the end plate 123. The control rods and gray rods 53 pass through holes 130 (FIG. 17) and the water-displacement rods through holes 132 (FIG. 16) in the core plate 49. Holes 134 pass the coolant. Only two rows of holes 130, 132 and 134 are shown in FIG. 14; these are representative of holes which extend throughout the core plate 49. The upper ends of the can 121a are welded to an end plate 131. An end plate 131 is shown enlarged in FIG. 18. This end plate 131 is also composed of intersecting strips 133, each strip having a flanges 135 at its ends. The strips 133 have perforations 136 which are congruent with the perforations 101. The upper end of each can 121a encompasses these perforations and is centered on the axis of the perforations defined by the center point 140 of intersection of strips 133. The lower and upper end of can 121b (FIGS. 1, 11) is similarly welded to a lower end plate 131a (FIG. 11) spaced a short distance from the end plate 131, to which can 121a is welded, and an upper end plate (not shown for 121b). The lower end plate 131a and the one not shown are identical to plate 131. Likewise, cans 121c, 121d, and 121e are welded between end plates 131 (not shown for 121c) at the bottom and end plate 131 at the top. A square plate section 91 is secured between each pair of end plates 131 and 131a to which respectively an upper end of a lower guide section 121, for example 121c, and a lower end of an adjacent upper guide section, for example 121d, are welded. The perforations 101 (FIG. 6) in each of the plate sections 91 are precisely aligned with the perforations 136 in the end plates 131 which abut each plate section. The strips 133 which form the end plates 131 and 131a extend only along the unperforated frames bounding the squares 109a and 109b so that they do not obstruct the movement of the control rod and gray rod clusters and the WDRC's. The plate section 91 and the end plates 131 and 131a are in each case secured together by two screws 137 (FIG. 18) through each flange 135, spaced symmetrically about the center line of the strip 133, and shoulder bolts 139 (FIG. 21) threaded into the centers of each of the strips. The upper ends of cans 121e are welded to end plates 131 and the lower and upper ends of cans 121f are welded to end plates 131a and 131. Between end plates 131 and 131a, at the upper end of each can 121e and the lower end of each can 121f, there is secured a plate section 91a substantially thicker than the other square plate sections 91, but having the same perforations. This thickened plate section 91a may be described as the parking plate section. It is the plate section in which the vanes (spiders) of the associated control-rod and gray-rod clusters 51 (FIGS. 3, 19) and WDRC's 55 (FIGS. 4, 19) are housed or parked when these clusters are in their uppermost position with the rods removed from the core 43. The upper ends of cans 121f are welded to top plate section 141 which is shown in plan view in FIG. 12. The top plate section 141 is perforated similarly to the plate sections 91 and 91a to permit insertion into each integrated unit of a columnar guide column and plate guides of the clusters 51 and 55 with their vanes 63, 73, 75, 77, 79 (FIG. 19). On each of its sides, each plate section 91 and 91a is provided with at least one pin 143 inserted in a semi-circular slot (FIGS. 19, 20) and at least one semicircular slot 145. The function of the pins 143 and the slots 145 is to align the integrated unit 110 with contiguous integrated units of columnar guides and plate sections. The pins and slots are staggered along the sides of the plate sections 91 and 91a. The pins 143 engage slots in the contiguous plate sections and the slots 145 engage pins in the contiguous plate sections. In the integrated units 110 including the columnar guide and the plate sections 91 and 91a, the guide sections 121a through 121f are precisely aligned and the perforations in the plate sections 91 and 91a are precisely aligned so that the rod clusters can be moved through the cruciform guide columns and through the perforations without binding. Integrated units of the other plate sections 93 through 99 integrated with columnar guide sections 121a through 121f similarly to the integrated unit 110 are also provided. These units are not shown, but they are similar to the units 110 and their structure can be understood from the above description of the unit 110. Like unit 110, each of the units formed of sections 93 through 97 includes a plurality of sets, typically of four each, of housings of cruciform transverse cross-section. The lowermost housing of each set is welded at its lower end to an end plate 123 and at its upper end to an end plate 131. The other housings of each set are each welded at its lower end to end plate 131a and at its upper end to end plate 131. In the case of the triangular plate section 99, there is only one housing of cruciform section. The plate sections are interposed and secured between the end plates 131 and 131a. The uppermost plate section in each case, which serves to house the spiders 63, 73, 75, 77, 79 (FIGS. 3, 4, 19) in the retracted position, is thicker than the other sections. The integrated unit 110 and the other integrated units are fabricated externally to the reactor. Each integrated unit is inserted in the reactor vessel 33. The pins 127 (FIGS. 11, 15) extending from the end plates 123 engage holes in the upper core support 49 thus aligning each unit laterally. Pins such as the pins 143 in each integrated unit engage slots such as slot 145 in the contiguous units and slots in each integrated unit engage pins in contiguous units. The integrated units are thus aligned into the complete guide assembly in upper internals 81. The upper internals include a flow-through screen 151 (FIGS. 1, 2) which is centered within the vessel 33 on the upper core support 49. For proper positioning, the flow-through screen 151 is provided with a skirt 149. After the integrated units are formed into an assembly, the assembly is centered within the flow-through screen 151. The upper-internals top plate 147 (FIG. 1) is then mounted on the top of the assembly. The flow-through screen 151 extends from the top plate 147. Formers 153 extend at each level of the plate 85. All cruciform cans have slots 155 (FIG. 11) to facilitate lateral flow of coolant. The flow-through screen 151 distributes the outflowing coolant over the whole volume encompassed by the screen and precludes concentrated flow directly through the nozzles 40. Such flow would be at a very high velocity because the space through which it is conducted is limited and failure of control and water-displacement rods would result. The plate guide assembly 85 is made up typically of nine square-shaped integrated units 130, four rectangular-shaped units, four right trapezoidal-shaped units, four left trapezoidal-shaped units and four triangular-shaped units. The total number of units required for the guide assembly is therefore only twenty-five. The integrated units are readily fabricated at relatively low cost. The integrated guide units are held in position by the pins 127, and 143 and by a vertical spring-loaded package (not shown) located in the upper-internals top plate 147. The control rod and gray-rod clusters 51 can be removed and inspected through openings (not known) in the top plate 147. The WDRC's can be removed and inspected by disassembly of the top plate 147. The guide plates are subdivided into areas. These areas define columnar volumes bounded by the arms of cans 121 except along the periphery of each plate where the columnar volumes are defined by the arms of less than 4 cans. These arms limit the flow of coolant through water-displacement rods when these rods are retracted. While a preferred embodiment of this invention has been disclosed herein, many modifications thereof are feasible. This invention is not to be restricted except insofar as is necessitated by the spirit of the prior art.
	description	This patent application claims the benefit of priority of U.S. Provisional Patent Application No. 62/370,164, filed on Aug. 2, 2016. The entire content of the before-mentioned patent application is incorporated by reference as part of the disclosure of this application. This invention was made with government support under Contract DE-AC52-07NA27344 awarded by U.S. Department of Energy. The government has certain rights in the invention. The present disclosure relates to energy production by nuclear fusion. The aim of fusion energy, such as inertial confinement fusion (ICF), is to contain fusion material at high temperatures and densities long enough to produce net energy gain. In ICF, a spherical shell of material (e.g. deuterium and tritium (DT) fuel) may be compressed several times in volume, using lasers, x-rays, or magnetic fields. In order to achieve high energy production, the compression may remain spherical without being disturbed by hydrodynamic instabilities in flight. However, achieving spherical and stable compression may be an obstacle to ignition. Prior attempts include creating a uniform capsule drive, driving the implosion with more energy, and imploding capsules with tight engineering tolerances on seed imperfections. Since the constraints are very tight, alternate schemes include relaxing the compression requirement, shock-igniting the capsule, and fast-ignition. However, these solutions have limitations. In particular, tight engineering and fielding constraints increase the cost of individual implosions, decreasing their attractiveness for energy production. Neither shock nor fast ignition at ignition-scale energies have been tested, and both require significant capital investment. One aspect of the disclosed technology relates to a method for asymmetric inertial confinement fusion. The method includes fixing in position a target capsule comprising an inertial confinement fusion fuel, wherein the target capsule is substantially spherical. The method further includes applying an oscillatory compression to the target capsule. The oscillatory compression includes compression at a first time in a radial direction orthogonal to a diametric axis of the target capsule, and compression at a second time along the diametric axis. The target capsule is driven into an ovoid shape and the ovoid shaped target capsule implodes at a third time upon application of an additional oscillatory compression. The following features may be included in any combination. The first time occurs before the second time and the second time occurs before the third time. The target capsule can include deuterium and/or tritium or helium-3. The target capsule is held inside a hohlraum. The target capsule may be cooled in the hohlraum and/or held in the hohlraum by support structures, such as membranes, foams, wires, etc. The compression can be performed by one or more of a laser beam or an ion beam. In another aspect an asymmetric inertial confinement fusion target capsule is disclosed. The target capsule includes an ovoid shaped shell and includes a fuel comprising deuterium and/or tritium. The target capsule is positioned inside a hohlraum. The target capsule can be cooled in the hohlraum and/or held in the hohlraum by one or more support structures. The compression is performed by one or more of a laser beam or an ion beam. The target capsule is positioned to receive an oscillatory energy to cause nuclear ignition of the target capsule. The oscillatory energy is provided by a driver that is configured to drive the target capsule with more energy along a first axis than a second axis orthogonal to the first axis, and subsequently drive the target capsule with more energy along the second axis than the first axis. The first axis may be an equatorial axis and the second axis may be a polar axis. In another aspect an apparatus for driving a target capsule in an inertial confinement fusion system is disclosed. The apparatus includes a first driver to deliver energy to the target capsule along a first axis, and a second driver to deliver energy to the target capsule along a second axis, where the second axis is orthogonal to the first axis. The apparatus also includes a controller configured to cause the first driver and the second driver to deliver energy to the target capsule in a sequence. The sequence includes delivering energy to the target capsule including a first energy along the first axis and a second energy along the second axis, wherein the first energy is greater than the second energy, and delivering energy to the target capsule including a third energy along the first axis and a fourth energy along the second axis, wherein the fourth energy is greater than the third energy. The following features may be included in any combination. The sequence may further include delivering a fifth energy to the target capsule, wherein the fifth energy includes energy along the first axis and energy along the second axis having approximately equal values. The first axis may be an equatorial axis and/or the second axis may be a polar axis. The target capsule may include deuterium and/or tritium. The target fuel capsule can be held inside a hohlraum; the target capsule may be cooled in the hohlraum, and/or held in the hohlraum using a support structure. The first energy, the second energy, the third energy, and the fourth energy may be provided by one or more of a laser beam or an ion beam. The first driver and the second driver are positioned to directly drive the target capsule. The first driver and the second driver are positioned to indirectly drive the target capsule by directing beams that impinge on an inside surface of a hohlraum before reaching the target capsule. Where possible, like reference numbers refer to the same or similar features in the drawings. Generally, the present disclosure is directed to an asymmetric ICF method and ovoid target capsule for achieving. The disclosed subject matter is directed to asymmetric inertial confinement fusion (ICF) technology and ovoid target capsule for achieving robust compression and ignition of thermonuclear fusion fuel. Both the disclosed methodology and the target capsule are configured to produce an asymmetric ovoid shaped implosion that is resistant to implosion imperfections. The ovoid geometry may create an internal flow field that protects the central hot region during stagnation. Disclosed embodiments solve both problems of hydrodynamic instabilities and capsule distortions via an asymmetric implosion. Unlike round implosions, ovoid (e.g. 3D oval or eggshell shaped) target fuel capsules can achieve high pressures in the presence of asymmetric drive and hydrodynamic instabilities. Such ovoid shells can be either pre-manufactured in the ideal shape or dynamically driven from an initially spherical shape to an ovoid shape. Due to enhanced implosion robustness, the engineering tolerances on the ovoid capsules can be relaxed. One example embodiment is an ovoid-shaped ICF target fuel capsule that includes an ICF fusion fuel, such as deuterium and/or tritium. These capsules may be pre-manufactured to have a predetermined ovoid shape. Another example embodiment is an asymmetric ICF method which dynamically compresses a substantially spherical target into an ovoid shape during implosion. The method for dynamic compression includes an oscillatory drive that first compresses a substantially spherical target in a radially inward direction orthogonal to a diametric axis, followed by compressing the target along the diametric axis. The diametric axis may be characterized as a polar axis in alignment with opposing poles. The foregoing first stage of the asymmetric compression may be characterized as directed along the equator and considered a “waist hot drive,” and the second stage may be characterized as being directed on the poles and considered a “pole hot drive.” This type of compression is counter to 40 years of ICF research, which asserts that compression must be nearly one-dimensional such as spherical to achieve high gain. In both the method and target capsule embodiments, the target fuel capsule may be contained in a hohlraum, such as a cylindrical hohlraum. For example, a cylindrical hohlraum may be used at the National Ignition Facility of the Lawrence Livermore National Laboratory. It is appreciated that references to a hohlraum axis are equivalent to the cylindrical/longitudinal axis in the case of a cylindrical hohlraum. When contained in a cylindrical hohlraum, the ovoid target capsule may be oriented with its longitudinal/major axis aligned with the hohlraum axis. Conventional ICF calls for ignition via round implosions with convergence at high velocity to achieve high pressure. Conventional ICF holds that deviations from sphericity rob the implosion of energy conversion efficiency and so a round driver is needed. High velocity and/or high convergence can lead to unstable linear growth of imperfections which leads to needing tight engineering tolerances to minimize the impact from instability seeds. Conventional thought holds that spherical implosions are required. Contrary to conventional thought, non-round shaped implosions may perform better than round implosions. As described below, the disclosed non-round implosions produce a greater yield amplification (higher ignition margin). For example, a calculated yield for an example round implosion may produce a burn-off energy of 0.05 megaJoules (MJ), a burn-on energy of 15.1 MJ, and a yield amplification of 302. A calculated yield for an example ovoid implosion may produce a burn-off energy of 0.04 MJ, a burn-on energy of 16.2 MJ, and a yield amplification of 371. As described below, ovoid shapes are stable because of large induced flow fields. Flow fields prevent instabilities and other drive asymmetries from distorting hot spots providing much higher pressures than round implosions. Robust implosions may be generated by using a time-varying drive. An early waist-hot drive may produce a sausage-like shell and a later pole-hot drive may cause the shell to be ovoid shaped. This is counter to conventional thought which holds that the drive must be very round (e.g., symmetric drive from all directions) for the entire implosion. In some example embodiments, an asymmetric inertial confinement fusion (ICF) method includes providing a substantially spherical target capsule comprising an ICF fuel. The method can further include dynamically compressing the target capsule so that the target capsule is initially compressed in a radial direction orthogonal to a diametric axis of the target capsule so as to elongate the target capsule and form opposing poles, and subsequently compressed along the diametric axis at the poles, whereby the target capsule is driven to implode with an ovoid geometry. In some example embodiments, an asymmetric inertial confinement fusion (ICF) target fuel capsule includes an ovoid shell comprising an ICF fuel material. The left hand side of FIG. 1 depicts at 100 a target capsule in a hohlraum 120 heated via indirect drive, and the right hand side of FIG. 1 depicts a target capsule at 150 not in a hohlraum heated via direct drive. At 150, direct drive beams 160 spherically impinge on target capsule 110. For example, target capsule 110 may be spherically shaped and direct drive beams 160 may impinge normal to the surface at each position on the surface of the spherical target capsule. When the direct drive beams are of equal intensity, the target capsule is uniformly heated on the surface. The direct drive beams may include laser beams, ion beams, or other particle or electromagnetic beams. At 100, capsule 110 is held inside hohlraum 120. In some example embodiments, the hohlraum may be made of gold, uranium, or other material. In some example embodiments, the target capsule 110 may be held in place in the hohlraum by one or more support structures, such as wires, foams and the like. The cooling can be effectuated using an external device; as a result of cooling ice can form inside the capsule. Drive beams 130 impinge on the inside surface of hohlraum 120. The drive beams impinging on the hohlraum cause the hohlraum material to radiate electromagnetic energy 140 such as x-rays. The radiated electromagnetic energy 140 impinges on the target capsule 110 causing the capsule to be heated. Because the hohlraum absorbs drive energy and then radiates electromagnetic energy, the configuration at 100 may be referred to as indirect drive. The right hand side of FIG. 2 depicts a hot-waist drive in a direct drive configuration at 250 and the left hand side of FIG. 2 depicts an indirect drive configuration at 200. Hot-waist drive may also be referred to herein as an equatorial drive. In some example embodiments, hot-waist drive may be accomplished by directly or indirectly driving the waist of target capsule 110 at a higher energy level than the polar axis which is orthogonal to the waist axis. For example, drive 132 may cause the hot-waist drive by being more energetic than drive 134. Drives 132 and 134 may be referred to as asymmetric because one axis drives at a higher energy that the orthogonal axis drive. The right hand side of FIG. 3 depicts a hot-polar drive in a direct drive configuration at 350 and the left hand side of FIG. 3 depicts an indirect drive configuration at 300. In some example embodiments, hot-polar drive may be accomplished by directly or indirectly driving the poles of target capsule 110 at a higher energy level than the waist axis which is orthogonal to the polar axis. For example, drive 334 may be more energetic than drive 332 thereby causing the hot-pole drive. In some example embodiments, the hot-polar drive may follow in time the hot-waist drive. In other example embodiments, the hot-polar drive of FIG. 3 may occur before the hot-waist drive of FIG. 2. FIG. 4A depicts a target capsule 410 after the waist drive depicted in FIG. 2 and the polar drive depicted in FIG. 3. Due to the waist and polar drives, the target capsule can be deformed from a spherical shape to a ovoid shape. The beams 430 may be of similar intensities or equal intensities to one another. Similarly, the beams 460 may be of similar intensities or equal intensities to one another. Beams 430 and 460 can cause heating and compression to produce ignition of target capsule 410. For example, target capsule 410 may be compressed to 1/30th of its original size, or another compressed size. FIG. 4B depicts a process 470, in accordance with some example embodiments. In particular, at 475, a target capsule is driven on a first axis with a first predetermined energy. A drive may be expressed as an energy and a direction from which the energy impinges on the target capsule. For example, as shown in FIG. 3, target capsule 110 may be driven with a higher energy along the polar axis at 334 than the equatorial axis at 332. The drive energy may correspond to a radiation temperature as described below. The stronger drive along one axis such as the polar axis compared to an orthogonal axis such as the equatorial axis may be referred to as a drive asymmetry. The drive asymmetry may be expressed as a percentage of stronger (or weaker) drive along one axis compared to the orthogonal axis. For example, in FIG. 3 the asymmetry may be expressed as a percentage, such as 2%, stronger drive along the polar axis than the equatorial axis. Any other percentage asymmetric drive may also be used. In some example embodiments, the first drive may be an axis different from a the polar axis. For example, the first drive axis may be an equatorial axis. At 480, the target capsule is driven on a second axis with a second predetermined energy. For example, as shown in FIG. 2, target capsule 110 may be driven with a higher energy along the equatorial axis at 132 than the polar axis at 134. In FIG. 2, the asymmetry may be expressed as a percentage, such as 3%, stronger drive along the equatorial axis that the polar axis. Any other percentage asymmetric drive may also be used. In some example embodiments, the second drive may be driven along an axis different from the equatorial axis. For example, the second drive axis may be a polar axis. At 485, the target capsule may be driven on one of more of the first axis and the second axis at with a third predetermined energy to cause the target capsule to ignite in a nuclear reaction. For example, the third predetermined energy may include driving the first axis and the driving the second axis with equal or nearly equal energies. For example, the equal or nearly equal energy drive on both axes may correspond to a spherical drive to the target capsule. In another example, the third predetermined energy may include driving the target capsule asymmetrically with greater energy on one axis such as the polar axis or the equatorial axis. In some example embodiments, the first drive at 475, the second drive at 480, and the third drive at 485 may be followed by other oscillatory drives such as the foregoing drive on one axis followed by drive along an orthogonal axis. In some example embodiments, the third drive may cause ignition, and in other embodiments, later drives may cause ignition. Ignition may follow an asymmetric drive or a spherical drive. FIG. 5 depicts two example ovoid shapes for target capsules, in accordance with some example embodiments. Shown at 520 is an ovoid shape found using a machine learning system as further described below. Shown at 510 is an ovoid shape determined using HYDRA simulation code based on the ovoid shape found using machine learning. The HYDRA-determined shape at 510 may have a much higher energy released and/or much improved metrics such as ignition factor compared to the spherical target capsule. The HYDRA-determined shape at 510 may have a higher energy released and/or improved metrics such as ignition factor compared to the target capsule shape found using machine learning. FIG. 6 depicts an example vector flow diagram produced by HYDRA for an ovoid target capsule at the time of ignition. The ignition time may also be referred to as “bang time,” and “burn on” may refer to nuclear reactions being turned-on in HYDRA. The length of each vector in the example vector field of FIG. 6 are proportional to a velocity in the direction of each vector. For example, vector 602 shows a velocity that is outward from the target capsule. Vector 604 also shows an outward velocity with a smaller magnitude than vector 602. Vectors 610 and 612 show a velocity that is inward rather than outward. Gas in these regions are directed outward again as shown by vectors 614 and 616. There is no vector flow, or nearly no vector flow, near the center 611 of the target capsule. The small flow at the center and the flows around the center such as flows 610 and 612 protect the center of the target capsule (e.g., the target capsule hotspot) from perturbations. The protective flow may result from the ovoid shape of the target capsule. Spherical target capsules do not have a similar protective flow. FIG. 7A depicts an example of a vector flow diagram for a round target capsule. The vector field in FIG. 7A corresponds to the round target capsule with nuclear reactions turned off (“burn off”) at the time of ignition of the capsule (“bang time”). The example of FIG. 7A includes a drive imperfection (“P4”), a tent imperfection (“tent”), and an imperfection in the cryogenic target capsule including cracks in the ice causing the surface to not be smooth (“ice”). The vector field shows gas leaking out of the tent scar at 710 and instabilities flowing toward the center of the target capsule at 720. As can be seen in the example of FIG. 7A, the flow due to instability at 720 is not pushed away from the center of the round target capsule. FIG. 7B depicts an example of a vector flow diagram for a ovoid target capsule. The vector field in FIG. 7B corresponds to the ovoid target capsule with nuclear reactions turned off (“burn off”) at the time of ignition of the capsule (“bang time”). The example of FIG. 7B includes a drive imperfection (“P4”), a tent imperfection (“tent”), and an imperfection in the cryogenic target capsule including cracks in the ice causing the surface to not be smooth (“ice”). The vector field shows swirling gas flow that keeps the gas from leaking out of the tent at 760 and instabilities pushed away from the center of the target capsule at 770. As can be seen in the example of FIG. 7B, the flow due to instability at 770 is pushed away from the center of the ovoid target capsule. Inertial confinement fusion (ICF) may compress a hollow shell of cryogenic deuterium-tritium (DT) fuel to thermonuclear conditions. With direct and indirect drive, a spherical ice shell may be encased in an ablator material. In some implementations of direct drive, the outer surface may be heated directly by impinging laser beams. In some implementations of indirect drive, a hohlraum encasing the target capsule (also referred to herein as a DT shell) may be heated to produce x-rays to heat the outer surface. As the capsule surface ablates, the spherical shell compresses to high temperature and density resulting in gas at the center of the shell igniting a fusion burn wave. The fusion burn wave may consume the DT shell and releases large amounts of fusion energy. The foregoing process may be sensitive to hydrodynamic instabilities, which can arise throughout the implosion. At early times, shocks can cause the Richtmyer-Meshkov growth of small-scale imperfections, which can be amplified by the Rayleigh-Taylor (RT) instability during the main capsule acceleration. As the capsule compresses, the back-pressure exerted by the central gas on the shell may cause the shell to decelerate, at which point the inner surface of the shell may become RT unstable. During this time, a perturbation of wave number k on the inner surface of a shell with characteristic density scale length Lm and local ablation velocity va (due to mass ablating into the forming hot spot) that experiences a deceleration of magnitude g can grow with a growth rate of the form, γ RT = α ⁢ kg 1 + k ⁢ L m - β ⁢ ⁢ kv a , Eq . ⁢ ( 1 ) In some implementations, α≈0.9 and β≈1.4. Under converging geometry, Bell-Plesset-like (BP) effects can enhance this growth, such that the total linear growth from any perturbation increases with the convergence ratio R(t=0)/R(t) of the shell. The foregoing effects and processes can lead to a distorted asymmetric shell at stagnation, which may adversely affect performance. For example, the Ignition Threshold Factor (ITF) formalism defines an ignition metric that depends on (among other things) the hotspot radius, Rhs, and the mode-weighted RMS deviation from that hotspot, ΔRhs ITF ∝ ( 1 - 1.2 ⁢ Δ ⁢ R h ⁢ s R h ⁢ s ) 4 . Eq . ⁢ ( 2 ) Larger values of ITF may be more robust ignitors, and ITF can be used to estimate a capsule's margin. For example, a hot-spot deviation ΔRhs/Rhs of 0.13 can reduce the margin for ignition by a factor of two. Other values of hot-spot deviation may also be used. The capsule performance may degrade when implosions are asymmetric. Because non-radial capsule stagnation may be detrimental to performance, skilled artisans in the field developed techniques to produce nearly spherical implosions. This may be accomplished by minimizing the seeds for asymmetric stagnation. For example, this may be accomplished by placing engineering tolerances on the capsule surfaces, reducing the effective footprints of the capsule support tent and DT gas fill tube, and/or ensuring a smooth capsule drive. Additionally, one may reduce the total integrated linear growth of hydrodynamic instabilities via techniques such as adiabat shaping and/or via implosions with low convergence ratios. In some embodiments consistent with the disclosed subject matter, a nonlinear stabilization process for ICF implosions may address drive and/or shell distortions. In particular, generated large scale coherent flows within the hotspot can stabilize other shell deformations that arise during stagnation. These flows, which are reminiscent of the zonal flows that appear in planetary atmospheres and magnetic fusion devices, can shear off both smaller scale instabilities and larger scale asymmetries, making these implosions more robust than those without zonal flows. Implosions with these flows may not be spherical, but rather asymmetric. The shell may be formed in the shape of an ovoid during capsule deceleration and/or stagnation. As explained herein, the disclosed technology provides for ovoid shaped shells that have improved ignition margins over the customary belief consistent with Eq. (2) that spherical shells have the largest ignition margins. The improvement by using an aspherical implosion may be shown using a machine learning algorithm trained on an ICF capsule simulation dataset. The dataset, a learning algorithm, and how to predict the improvement of using an ovoid shape is disclosed below along with a series of simulations and/or computations at a predicted robust point. The simulations and/or computations show that the ovoid implosions are more resilient to distortions driven by asymmetric drives and the capsule support tent than those driven symmetrically. Simulations and/or computations consistent with the disclosed subject matter further quantify the performance of implosions under adverse conditions. Data science techniques may be used to search for implosions in flat regions of parameter space, thus being explicitly robust to perturbations around a design point. In this way, robustness may be optimized. In some example embodiments, a baseline implosion design is chosen, a suite of multi-dimensional simulations of perturbations is generated around the baseline, and a quantity of interest Q (for instance, the total produced yield) is fit to a functional form of the i individual input variations x∈d={xi}, thereby creating a data-trained surrogate model {tilde over (Q)}≈Q that can be evaluated for any value of x, not necessarily included in the original simulation ensemble. One can then use {tilde over (Q)} to optimize x for robustness without performing additional simulations. In studies that produce rules like Eq. (2), the surrogate model may be a power-law of the form, Q ~ ⁡ ( x _ ) = ∏ i d ⁢ f i ⁡ ( x i ) a i , Eq . ⁢ ( 3 ) with fi(xi) being an analytic function of the input parameters, such as (1−bixi), for a constant bi. It is also possible as in Eq. (2) for the xi not to be simulation input parameters, but rather physical quantities extracted from the simulation database, like ΔR/R. Eq. (3) may be constructed as a linear regression on log fi and interpreted because the relative strengths and impacts of the various terms come directly from the constants ai and bi. They can be built with datasets of a few hundred, a few thousand, or any other number of individual simulations. However, Eq. (3) may preclude interactions between the xi and the ai which may be held constant over the entire space. This may make power laws less favored for design optimization, when the xi are simulation input parameters. To search for a robust implosion, we may look to a family of non-parametric {tilde over (Q)}, such as those generated by machine learning statistical algorithms. These methods may not necessarily make assumptions about the underlying structure or functional form of Q but may require many simulations to achieve a level of accuracy, especially when the input space is of high dimensionality (d>4). However, machine learning techniques can build accurate surrogate models from ICF simulation ensembles (without alpha-particle deposition) of d=7 with ˜3500 simulations. In some example embodiments, the application of machine learning to construct surrogate models for ICF implosions may be extended to search for a design that ignites under adverse conditions. A dataset disclosed herein consists of a nine-dimensional survey of drive amplitudes, asymmetries and convergence ratios, and uses machine learning techniques to build surrogate models for multiple quantities of interest. The surrogates may be combined to serve as fast and complex interpolators to the dataset, into a robustness cost function for optimization and find a set of input parameters that produce an optimally robust implosion. Additional surrogates may be queried, which may be trained on the individual Legendre moments of the DT fuel shape, and used to determine the optimal input parameters produce an ovoid shape at stagnation. The surrogate models may also predict that the optimal ovoid-shaped implosion is more resilient to a variety of perturbations than the baseline spherical implosion. Additional simulations at the optimal point (which are not in the original dataset) may be used to confirm the surrogate predictions and provide physical insight into the high performance of this implosion. In one approach, the first step towards searching for a robust design is to create a large dataset of input simulations, upon which to build a surrogate model. In some example embodiments, an ensemble of approximately 60,000 two-dimensional ICF capsule simulations may be used. Any other number of capsule simulations may also be used. The data-set may constitute a nine-dimensional parameter scan of time-varying drive magnitudes, drive asymmetries (described by Legendre modes 1, 2 and 4, Pn), and capsule gas fill densities. Due to its large size (e.g., 5 petabytes of raw data), the data may be processed using the disclosed in-transit data analysis technique which uses internal servers (built on the bean-stalked library) to direct dedicated compute nodes to analyze and erase raw data on-the-fly before overflowing the file system. The baseline simulation may be run using radiation hydrodynamics code such as HYDRA, which is an axisymmetric variant of a high-density-carbon (HDC) National Ignition Facility (NIF) implosion design, meant to ignite in 1D, with a 20 μm dopant layer of 3% Si embedded in the 75 μm-thick shell of 1108 μm outer radius. Both the DT and the HDC may use tabular equations of state (LEOS 1018 and 64, respectively) and opacities. The ice layer may be 55 μm thick and the central gas may have a baseline density of 5×10−4 g/cm3. The initially Arbitrary-Lagrangian-Eulerian mesh with 513 angular zones and 321 impedance matched radial zones may remap to an entirely Eulerian elliptic mesh near stagnation to trade accuracy for robustness (99.9% of the simulations completed without human intervention). The nine-dimensional study may constitute Latin Hypercube Sampling of a space around the baseline implosion, with linearly varying drive magnitude A and asymmetry perturbations between three time points (the end of the first shock “trough,” the end of the “rise” to peak laser power, and the end of “peak” radiation drive). Time-dependent perturbations may ramp-up from time zero and down from the end of peak power. P1 and P4 may have the same value at the three time points, but P2 and A can vary (see FIGS. 8A-8B for an example). The peg points may be sampled linearly between ±[2, 10, 5, 25]% for [P1, P2, P4, A]. The capsule gas fill density may be sampled logarithmically between 0.2× and 0.5×, for a total of 9 independent variables. With the dataset, a machine learning technique may be used to build surrogate models that infer the behavior of multiple quantities of interest at any point within a design space, even where no simulations exist. In some example embodiments, Random forest regression may be used which may consist of a series of bootstrapped decision trees trained on subsets of the whole dataset. As such, a random forest can handle large quantities of high-dimensional data and incorporate nonlinearities, which may be important near regions of high yield that are potentially surrounded by steep “cliffs.” A random forest regression surrogate for the logarithm of the total energy yield (log10 Y) may achieve an 80% cross-validation mean error of 8%. For example, the surrogate model may be trained on a random 80% of the data and the prediction tested against the other 20%: the mean error on the prediction for the 20% random hold-out points may be 8%. Being a statistical fit to the entire data set, the yield surrogate may not only allow for the continuous real-time interpolation of parameter space between the discrete samples but may also be used to define a metric for robustness. For this measure, a point in parameter space may be chosen and with the surrogate make 1000 random input variations within a hypercube centered at chosen location with side length Δ of 10% of the total sample space. The number of surrogate evaluations that achieve Y>1 MJ may serve as a local estimate of the probability of achieving high yield under variable conditions: (Y>1\|Δ=0.1). This function may serve as a smoothing operator on the yield, filtering out narrow “peaks” of high performance in favor of more broad “plateaus.” Surrogates can be built for any number of extracted physical quantities, such as DT fuel areal density (ρR), the first Legendre moments of the DT shell (P0-8), and/or an ignition threshold factor metric ITFX≐Y(ρR)2 (“ignition”≐ITFX{tilde under (>)}1) at the time of peak energy production (“bang time”). By combining the surrogate models, a cost function can be defined for multi-dimensional optimization that similarly weights robustness and yield:C=10+ITFX Eq. (4)The first term in Eq. (4) may include broad areas of parameter space that ignite, and the second term may include locations that are high up the ignition cliff. To speed up convergence toward an optimal point, the first term may be weighted higher (which being a probability maximizes at 1) to make it of similar order as the second term (which crosses the ignition threshold at 1, but can be 10 or higher). Furthermore, because the operational space in this example is nine-dimensional and a single evaluation of may use 1000 surrogate evaluations, a simplex based optimization algorithm may be chosen to avoid gradient evaluations in the search for a robust design. Optimization of Eq. (4) can produce the example drive shown in FIG. 8A. In some example embodiments, the optimal point has a 0.5× gas fill multiplier. The optimal point, which is predicted to achieve high yield, has a time-varying P2 drive asymmetry. FIG. 8B compares the DT bang-time fuel shapes as predicted by the Pn surrogate models for both the baseline and optimal drives. Due to the time-varying asymmetry, the stagnated shape of the optimal drive may be an ovoid rather than a sphere. In the example of FIG. 8A, radiation temperature and applied asymmetry are plotted against time. The radiation temperature may correspond to a drive energy. Shown are baseline energies compared to the optimal energy. The drive asymmetry is shown at 810, 815, and 820. Drives above line 805 may correspond to polar drives and drives below line 805 may correspond to equatorial drives. In the example of FIG. 8A, a polar drive with an asymmetry of approximately 2% is applied at 810. For example, the 2% asymmetric drive may correspond to 2% more energy along the polar axis compared to the equatorial axis. At 815, an asymmetric drive may be applied with approximately 5% more equatorial drive than the polar drive. At 820, polar drive may again be applied with approximately an asymmetric drive of 4% more polar drive. The asymmetric drives such as 810, 815, and 820 may be applied in succession over a time period creating an oscillatory drive to the target capsule. For example, the oscillatory drive may correspond to a polar drive such as 810 followed by an equatorial drive such as 815, followed by a polar drive again at 820. In the example of FIG. 8A, three asymmetric oscillatory drives are shown. Other numbers of oscillatory drives may also be used. In the example of FIG. 8A, drives 808 and 822 may be symmetric drives having no or little asymmetry between the polar and equatorial energies. The surrogates may also predict that the optimal ovoid-producing drive is more resistant to other perturbations than the symmetrically driven baseline. FIG. 9 shows surrogate outputs for yield under a changing peak drive multiplier, represented by the total drive fluence ∫Tr4dt (normalized to the baseline). This metric may serve as an estimate of the total laser energy to achieve a given radiation drive. To eliminate the effects of the remaining drive parameters, the optimal and symmetric implosions may be compared with the same gas fill, P1, and P4 perturbations such that the change in performance is due to the P2 drive alone. Both designs may fall off in yield as the drive is reduced. While adding a +2% P1 to the baseline may make ignition impossible for any drive in the example parameter space, the location of the optimal point's ignition cliff remains unchanged at 1.05 total fluence. FIG. 9 depicts the surrogate's estimate of yield under changing peak drive, as represented by the total normalized drive fluence ∫Tr4dt/(∫Tr4dt)baseline for the baseline (round) and optimal (ovoid) cases. Unlike the baseline, the optimal point can ignite for fluences >1.05, even with an applied P1. Movement of the ignition cliff may be understood via yield contours predicted by the surrogate shown in FIG. 10A for varying P2 drives on the rise to peak power (P2R) and at the end of the peak radiation drive (P2P). The contour lines 1010 correspond to the surrogate-predicted P2 moment of the fuel at bang time, and the point 1020 indicates the location of the optimal implosion. There may be a broad, high-yield ridge along a line of compensating P2R and P2P drives, with higher yields favoring a negative P2R and positive P2P. This compensating drive does not result in a round implosion, but rather an ovoid with positive P2 (≈20%) (see FIG. 10B for the reconstructed fuel shapes at the corresponding points shown in FIG. 10A). With a P1 perturbation added, the high-yield ridge contracts toward more extreme compensating drives. The optimal implosion may remain within the ridge boundaries, while the round implosion may fall to low yield. Additionally, the high-yield ridge may appear to extend beyond the boundaries of the design space, which may suggest that an implosion with higher performance than the optimal point may be present if more extreme compensating P2 drive perturbations are considered, outside the bounds of the considered dataset. The surrogate predictions in FIGS. 10A and 10B show that the optimal point may sit on a topological ridge in parameter space defined by compensating P2 drives of opposite signs. More extreme variations in drive P2 may be more robust to other perturbations. The high-yield ridge corresponds to ovoids rather than round implosions. Some ovoids may have bang-time fuel shapes given by P2/P0≈20%. FIG. 10A depicts contours of the surrogate predictions for the yield and bang time fuel P2/P0 for varying applied drive P2 on the rise to peak power (P2R) and at the end of the peak radiation drive (P2P), with and without an applied P1 perturbation. The reconstructed fuel shapes (μm) from the P0-P8 surrogates at the dots in FIG. 10A are plotted in FIG. 10B. In some example embodiments, FIG. 10A at 1020A depicts the optimal implosion. High-yield implosions lie along a ridge of compensating P2 drives (negative on the rise, positive at the peak), with the ridge shifting that may require more extreme drives to reduce the effects of P1. The P2 drives that correspond to the high-yield ridge that may produce ovoid implosions with positive P2/P0≈20%. Point 1020A in FIG. 10A corresponds to curve 1020 in FIG. 10B, Point 1020A in FIG. 10A corresponds to curve 1020 in FIG. 10B, point 1022A corresponds to curve 1022, point 1024A corresponds to curve 1024, and point 1026A corresponds to curve 1026. To confirm the surrogates' predictions of a robust ovoid at a location not in the original simulation database, a series of 2D HYDRA simulation may be performed for the optimal point and for a symmetric simulation with the same drive amplitude and gas fill, so that any differences are due to the time-varying drive asymmetry. To isolate alpha-particle bootstrapping from hydrodynamic effects, “burn-off” simulations with a reduced fusion cross-section may be performed (with their total reported yields re-scaled by the same factor for comparison to the full “burn-on” cases.) For example, with these conditions, the surrogate predicts, 17.4 MJ for the burn-on symmetric implosion and 15.2 MJ for the ovoid. HYDRA confirms the predicted high performance for the optimal point (Yburn-on=16.6 MJ, Yburn-on/Yburn-off=355 for the ovoid vs. 17.2 MJ and 262 for round). This corresponds to a burn-on yield-over-1D of 97%, but a burn-off yield-over-1D of only 71%. FIG. 11 depicts the trajectories of DT burn-rate averaged temperature and density for the round and ovoid implosions, in accordance with some example embodiments. The ovoid implosion compresses and burns at a higher density and lower temperature than a round implosion. Since the ICF hotspot self-heating condition can be written as f(ρR)>g(T) for some functions f and g, the higher density and lower temperature ovoid burn trajectories in FIG. 11 imply a more efficient burn process. This helps to explain the near 1-D full burn yield. FIG. 11 depicts trajectories in DT burn-rate weighted density and temperature for ovoid and round implosions (with and without alpha-particle heating). The ovoid may compress and burn at lower temperatures and higher densities than the round implosion. HYDRA may be used to show how an ovoid shape arises from the time-dependent implosion dynamics. An asymmetric shock bounce (that begins deceleration) seeds vorticity in the gas. The negative P2 drive on the rise to peak power may bias the capsule compression equatorially. As the compressed gas meets on axis, it may form axial jets. The late time positive P2 drive may prevent the jets from escaping and the flow may circle on itself forming two co-axial counter-propagating vortex rings, and the hotspot may organize during stagnation into a configuration such as the example shown in FIG. 12. As shown, the exterior shell may conform to the vortex rings (forming an ovoid) and the central gas between the vortexes may become trapped in a vorticity quadrupole. The hotspot may be elongated and may not align with the high-pressure central core. Strong coherent flows may exist throughout the hotspot, so that the cold dense shell on the equator may accrete into the central high-pressure region, burn, and exhaust via the poles. A single fluid element may be estimated to take ˜0.5 ns to make a complete revolution of the eddies shown in FIG. 12, longer than typical ({tilde under (<)}0.1 ns) ICF burn and disassembly times. FIG. 12 depicts velocity (arrows), density (gray), ion temperature (left side contours), and pressure (right side contours) at the time of peak energy production for a burn-off HYDRA simulation of the ovoid implosion. These flows may nonlinearly suppress the growth of hydrodynamic instabilities. FIG. 13 shows an example plot of the upper right section of a stagnating shell for a burn-off ovoid implosion with added roughness on the inner ice surface (at NIF ignition specifications) and a perturbation on the ablator surface due to the capsule support tent (calibrated to a 100 nm tent on an HDC capsule). A first distortion depicted in the example of FIG. 13 is a large on-axis jet, originally seeded by ice surface roughness, that is entrained in a large axial flow field, directed down toward the central hotspot. A second, distortion shown is the shell having a visible low-density “scar” due to the tent. However, the background flow may meet the axis jet head-on, form a high velocity (>300 km/s) shearing layer, and direct the jet away from the hotspot. The shear layer may direct the flow field tangentially to the tent scar, reducing convective loss through the tent hole. FIG. 13 depicts density and velocity fields in the upper part of the stagnating shell for an ovoid burn-off HYDRA simulation perturbed with ice layer roughness and capsule support tent membrane. The background flows may set up a high-velocity shear layer (thick arrows) that mitigate the effects of the perturbations during stagnation. The shearing in FIG. 13 may be strong enough to compete with shell distortions that occur during deceleration. The local shear rate can be estimated as the ratio of the local velocity to the eddy size. In the example of FIG. 13, the eddy is roughly 5 μm across with an average velocity of (150+300)/2=225 μm/ns, which gives a shearing rate of 45 ns−1. Perturbations on the shell may evolve at a characteristic rate that can be estimated as either the RT growth rate (given by Eq. (1)) or the inverse of the shell breakup time τ τ = 2 ⁢ π ⁢ R ⁡ ( ρ ⁢ R ) s ⁢ h ⁢ e ⁢ l ⁢ l l ⁢ P stag , Eq . ⁢ ( 5 ) for some mode number l=kR on a shell with areal density (ρR)shell stagnating against a hotspot with pressure Pstag. FIG. 14 shows that the eddy shear rate may be larger than both rates predicted by Eqs. (1) and (5), suggesting that the shear flows present in the ovoid may be strong enough to impact the growth of shell perturbations during capsule deceleration. FIGS. 7A-7B depict an example of the local shearing rate of the eddy in FIG. 13 which is larger than both the deceleration Rayleigh-Taylor growth rate (Eq. (1)) and the shell breakup rate (Eq. (5)). Ovoid implosions can tolerate larger distortions than round implosions. FIG. 15 shows of examples of contours of yield >1 MJ for the round and ovoid implosions with varying levels of applied P4 asymmetry and tent amplitude. In the example of FIG. 15, the performance of both implosions falls off with increasing perturbation strength, but the ovoid implosion may maintain high yield for a larger parameter range. For example, the ovoid may produce >9 MJ with a 300 nm tent and +3% P4, where the round implosion fails to ignite. FIG. 15 depicts contours of yield for the round and ovoid implosions under combination P4 drive asymmetry and tent perturbation amplitude. Simulations confirm the surrogates' predictions of an asymmetric ovoid implosion that is more resilient to perturbations than symmetrically driven one-dimensional designs. The subject matter disclosed herein includes a new class of ICF implosions. The new implosions are ovoid in shape which challenge the philosophy behind the traditional view including Eq. (2) that suggest spherical implosions uniformly outperform aspherical implosions. The new implosions may be driven by strong time-varying drive asymmetries which may serve to set up large-scale coherent convective flows at stagnation. Ovoid implosions may be more robust than spherical implosions to shell perturbations from either drive asymmetries or the hydrodynamic growth of smaller-scale shell imperfections. They may be more robust due to locally strong shearing rates induced by the flows. Although the ovoids challenge the notion that the ideal implosion is symmetric (that asymmetries always degrade performance), they are consistent with some studies and anecdotal observations of improved performance with positive P2 shapes. The ovoid shape may cause large-scale convective flows that circulate into the hotspot on the equator and out via the poles. In fluid dynamics, the upper and lower solutions are repelling circular vortexes (like smoke rings and mushroom clouds) joined by a Burger's vortex at the origin, all of which are axially symmetric 3D fluid solutions with increased stability at low Reynolds numbers, like those expected in ICF hotspots. This flow pattern is similar to that proposed to exist in the Jovian core that drives the large zonal flows on the planet's surface. Zonal flows also play a prominent role in the stabilization of magnetic fusion drift wave turbulence. These waves, like the Rossby waves in planetary atmospheres, experience an inverse cascade phenomenon, in which small scale perturbations can nonlinearly couple energy into larger scales. As such, larger scale flows can serve as energy sinks for smaller scales by shearing away perturbations as they grow. The ovoid implosions may exhibit a process. Because nonlinear Rayleigh-Taylor bubble merger can be thought of as an inverse cascade process (smaller bubbles merge into large bubbles), the stabilization of small-scale perturbations by larger scale flows may occur during the capsule stagnation phase. The flow fields induced in the ovoid implosion may protect the hotspot from shell imperfections during stagnation by serving as an energy sink for smaller scale shell instabilities, shearing away unstable shell imperfections. The central hotspot may not be in a hydrostatic equilibrium, but rather a nonlinear hydrodynamic one, dominated by coherent convective flow. Zonal flows induced in ICF implosions allows for a nonlinear mechanism for the stabilization of stagnating shells. Stabilization may come at a price, as the example of FIG. 9 shows, that the required energy to ignite an oval implosion is higher than a perfectly spherical implosion (e.g., roughly 8%). In this sense, the ovals contain a finite amount of residual hotspot kinetic energy at stagnation, but that energy may be coherently organized in a stabilizing flow pattern, trading one-dimensional margin for robustness in higher dimensions. Disclosed herein is a class of high-performing asymmetric implosions that are more robust to drive and shell perturbations than those driven symmetrically. A machine learning algorithm was trained on a large simulation dataset to show a stable implosion not shown before. From the foregoing, it will be appreciated that specific embodiments have been described herein for purposes of illustration, but that various modifications may be made without deviating from the scope of the disclosure. Therefore, it will be appreciated that the scope of the present disclosure fully encompasses other embodiments which may become obvious to those skilled in the art. In the claims, reference to an element in the singular is not intended to mean “one and only one” unless explicitly so stated, but rather “one or more.” All structural and functional equivalents to the elements of the above-described preferred embodiment that are known to those of ordinary skill in the art are expressly incorporated herein by reference and are intended to be encompassed by the present claims. Moreover, it is not necessary for a device to address each and every problem sought to be solved by the present disclosure, for it to be encompassed by the present claims. Furthermore, no element or component in the present disclosure is intended to be dedicated to the public regardless of whether the element or component is explicitly recited in the claims. No claim element herein is to be construed under the provisions of 35 U.S.C. 112, sixth paragraph, unless the element is expressly recited using the phrase “means for.” Although the description above contains many details and specifics, these should not be construed as limiting the scope of the disclosure but as merely providing illustrations of some of the presently preferred embodiments. Other implementations, enhancements and variations can be made based on what is described and illustrated in this patent document. The features of the embodiments described herein may be combined in all possible combinations of methods, apparatus, modules, systems, and computer program products. Certain features that are described in this patent document in the context of separate embodiments can also be implemented in combination in a single embodiment. Conversely, various features that are described in the context of a single embodiment can also be implemented in multiple embodiments separately or in any suitable subcombination. Moreover, although features may be described above as acting in certain combinations and even initially claimed as such, one or more features from a claimed combination can in some cases be excised from the combination, and the claimed combination may be directed to a subcombination or variation of a subcombination. Similarly, while operations are depicted in the drawings in a particular order, this should not be understood as requiring that such operations be performed in the particular order shown or in sequential order, or that all illustrated operations be performed, to achieve desirable results. Moreover, the separation of various system components in the embodiments described above should not be understood as requiring such separation in all embodiments.
	summary
	description	The contents of the following Japanese patent application are incorporated herein by reference: No. 2014-172299 filed on Aug. 27, 2014. 1. Technical Field The present invention relates to a radiation detecting element, a radiation detecting apparatus, and a manufacturing method of a radiation detecting element. 2. Related Art It has been known that a scintillator of a measuring apparatus to measure radiation is made thin while maintaining the strength of the scintillator by bonding the scintillator and a transparent reinforcing member with an adhesive to make them thin, as described for example in Japanese Patent Application Publication No. 2012-26821. However, when a scintillator and a reinforcing member are bonded by using an adhesive, scattering and reflection occur at interfaces between the scintillator and the adhesive and between the adhesive and the reinforcing member. Due to this, a blurred image is formed on a sensor, and the resolution deteriorates. According to a first aspect of the present invention, a radiation detecting element comprises: a substrate transparent to visible light; and a fluorescent screen that emits fluorescence in response to radiation by a dopant added to a material that is the same as a material of the substrate, wherein the fluorescent screen is thinner than the substrate, and the substrate and the fluorescent screen are bonded while maintaining continuity of a refractive index. According to a second aspect of the present invention, a radiation detecting apparatus comprises: a radiation detecting element having a substrate transparent to visible light, and a fluorescent screen that emits fluorescence in response to radiation by a dopant added to a material that is the same as a material of the substrate, wherein the fluorescent screen is thinner than the substrate, and the substrate and the fluorescent screen are bonded while maintaining continuity of a refractive index; an imaging optical system that forms an image of fluorescence emitted by the fluorescent screen; and a photoelectric conversion element on which photoelectric conversion pixels that perform photoelectric conversion on fluorescence an image of which has been formed are disposed two-dimensionally. According to a third aspect of the present invention, a manufacturing method of a radiation detecting element comprises: bonding, by solid state diffusion, a substrate transparent to visible light and a fluorescent screen that emits fluorescence in response to radiation by a dopant added to a material that is the same as a material of the substrate; and thinning the fluorescent screen. The summary clause does not necessarily describe all necessary features of the embodiments of the present invention. The present invention may also be a sub-combination of the features described above. Hereinafter, (some) embodiment(s) of the present invention will be described. The embodiment(s) do(es) not limit the invention according to the claims, and all the combinations of the features described in the embodiment(s) are not necessarily essential to means provided by aspects of the invention. FIG. 1 is a schematic cross-sectional view of a radiation detecting apparatus 10. The radiation detecting apparatus 10 is an indirect conversion-type two-dimensional radiation detector to measure a two-dimensional signal of a high energy radiation such as an X-ray by converting it into low energy fluorescence. One example of radiation is an X-ray free electron laser (hereinafter, called an XFEL in some cases). The radiation detecting apparatus 10 has a cylindrical housing 12, a CCD sensor 14, an objective lens 16 and a radiation detecting element 20. The radiation detecting element 20 is circular plate-like, and is disposed on one end surface side of the housing 12. The radiation detecting element 20 contains a fluorescent substance, and converts a high energy radiation that enters from the direction indicated with an arrow 18 into numerous low energy fluorescence lines. The radiation detecting element 20 makes detection by the CCD sensor 14 possible by converting radiation at a frequency that is outside the region of sensitivity and therefore cannot be detected by the CCD sensor 14 into fluorescence at a frequency that is within the region of sensitivity. It should be noted that the shape of the radiation detecting element 20 is not limited to a circle, but may be any shape such as a square. The CCD sensor 14 has photoelectric conversion pixels that are disposed two-dimensionally. The CCD sensor 14 converts light in the region of sensitivity including visible light the image of which has been formed by the objective lens 16 into an electrical signal and outputs it. The CCD sensor 14 is disposed on another end surface side of the housing 12 so that the surface on which the photoelectric conversion pixels are disposed faces the radiation detecting element 20. The CCD sensor 14 is one example of a photoelectric conversion element, and an EMCCD sensor or a CMOS sensor may be used in place of the CCD sensor 14. Because in an EMCCD sensor, a function of multiplying electrons is added to a readout unit of the CCD sensor, faint light can be measured highly sensitively by using the EMCCD sensor. Because a CMOS sensor adopts a high-speed readout system, light can be measured at a high frame rate by using the CMOS sensor. The objective lens 16 is disposed between the radiation detecting element 20 and the CCD sensor 14 so that the objective lens 16, the radiation detecting element 20 and the CCD sensor 14 are on a straight line. Furthermore, preferably, the CCD sensor 14, the radiation detecting element 20 and the objective lens 16 are disposed so that the center of the surface of the CCD sensor 14 on which the photoelectric conversion pixels are disposed and the center of the radiation detecting element 20 match the optical center of the objective lens 16. The objective lens 16 forms an image of fluorescence emitted from the radiation detecting element 20 on the surface of the CCD sensor 14 on which the photoelectric conversion pixels are disposed. The objective lens 16 may be configured with a single lens, or may be configured with a plurality of lenses including an objective lens and an imaging lens. It should be noted that the objective lens 16 is one example of an imaging optical system. FIG. 2 is a schematic perspective view of the radiation detecting element 20. The radiation detecting element 20 has a substrate 22, a fluorescent screen 24 and anti-reflection films 26, 28. The substrate 22 is disposed on the objective lens 16 side of the radiation detecting element 20. The transmittance of fluorescence which is visible light for the substrate 22 is higher and the fluorescence is more transparent than radiation whereas the transmittance of radiation for the substrate 22 is lower than visible light. The substrate 22 is configured by Y3Al5O12 (hereinafter, called YAG). It should be noted that YAG is one example of a material to configure a substrate, and any other materials including Lu2SiO5, LuYSiO5, LuYSiO5, Gd3Ga5O12, CdWO4, Bi4Ge3O12, Gd2SiO5, Gd2O2S, Y2SiO5, Yb2SiO5, LuAlO3, Lu3Al5O12, Gd3Al2Ga3O12, Lu0.7Y0.3AlO3, or a mixture thereof may be used. The fluorescent screen 24 is disposed on a side of the radiation detecting element 20 where radiation enters. The fluorescent screen 24 emits fluorescence corresponding to radiation by adding a dopant, Ce, to YAG which is a material same as the material of the substrate 22. That is, the fluorescent screen 24 functions as a scintillator that emits fluorescence corresponding to radiation that has entered there. It should be noted that, in the present embodiment, YAG to which Ce has been added emits fluorescence whose wavelength centers at 550 nm. Also, Ce is one example of a dopant to be added to the fluorescent screen 24, and as other dopants, Mn, Tl, Sn, Pb, Eu, Tb, La, Gd, Al, Ge, Yb, Nd, Sm, Er, Tm, Am or Pr may be used. The thickness of the substrate 22 needs to have the proportion so that while the handling strength is maintained, radiation that is transmitted through the fluorescent screen 24 gets attenuated and extinguished. Also, in order to suppress spherical aberration occurring in the substrate 22, it is preferred to make the thickness of the substrate 22 thin. In the present embodiment, the thickness of the substrate 22 is 3 mm for example. On the other hand, the thickness of the fluorescent screen 24 is smaller than the thickness of the substrate 22, and is preferably within the range of 1 μm to 2 mm. In the present embodiment, the thickness of the fluorescent screen 24 is 20 μm for example. By making the fluorescent screen 24 thin, scattering of light-emission points of fluorescence in the traveling direction of radiation can be prevented; as a result, the spatial resolution of the radiation detecting apparatus 10 improves. However, the spatial resolution of the radiation detecting apparatus 10 does not improve to be equal to or exceed the resolution that is determined by the diffraction limit determined by fluorescence wavelength and the numerical aperture of the objective lens 16 and the pixel size and optical magnification of the CCD sensor 14. For this reason, in a case where the highest spatial resolution of the radiation detecting apparatus 10 is to be obtained, preferably, the objective lens 16 having high optical magnification so that the pixel size of the CCD sensor 14 becomes smaller than the diffraction limit size of fluorescence is used, and the thickness of the fluorescent screen 24 is selected so that it becomes the depth of focus determined by the diffraction limit of fluorescence. Furthermore, the thickness of the fluorescent screen 24 is determined considering that the target viewing field, spatial resolution, and amount of emitted light of fluorescence can be ensured, and that a region to which a fluorescent substance, Ce, is added can be ensured on the entire surface of the fluorescent screen 24 even when the fluorescent screen 24 is bonded with the substrate 22 by solid state diffusion. The fluorescent screen 24 and the substrate 22 are bonded while maintaining continuity of the refractive index. Here, bonding while maintaining continuity of the refractive index means that the refractive index structure of an interface where the fluorescent screen 24 and the substrate 22 are bonded is substantially uniform. For example, when the difference between the refractive index of the substrate 22 configured by YAG and the refractive index of the fluorescent screen 24 in which Ce is added to YAG is on the order of 0.1%, and the substrate 22 and the fluorescent screen 24 are bonded, the refractive index structure of the bonding interface can be said to be uniform. By making the refractive index structure of the interface where the fluorescent screen 24 and the substrate 22 are bonded substantially uniform, scattering, refraction and reflection of fluorescence at the bonding interface can be prevented. The anti-reflection film 26 and the anti-reflection film 28 are disposed on both end faces of the bonded substrate 22 and fluorescent screen 24, respectively. The anti-reflection film 26 is disposed on a surface of the substrate 22 that is opposite to the surface of the substrate 22 that is bonded with the fluorescent screen 24. The anti-reflection film 26 prevents reflection of fluorescence emitted from the fluorescent screen 24. The anti-reflection film 26 is a thin film having the thickness equivalent to ¼ of the wavelength of fluorescence for example. In the present embodiment, because the wavelength of fluorescence of YAG to which Ce is added is 550 nm, the anti-reflection film 26 is a thin film of 137.5-nm thickness which is equivalent to ¼ of 550 nm. The anti-reflection film 26 prevents reflection of fluorescence by cancelling out light reflected on the interface between the substrate 22 and the anti-reflection film 26 by means of light reflected on the front surface of the anti-reflection film 26. Also, a multi-layered film on which two or more layers of dielectric films are coated may be used as the anti-reflection film 26. The anti-reflection film 28 is disposed on a surface of the fluorescent screen 24 that is opposite to the surface of the fluorescent screen 24 that is bonded with the substrate 22. The anti-reflection film 28 prevents reflection of fluorescence emitted from the fluorescent screen 24 and re-reflection of reflected light of fluorescence that occurs on the surface of the substrate 22 that is opposite to the surface of the substrate 22 that is bonded with the fluorescent screen 24. It should be noted that because the configuration of the anti-reflection film 28 is the same as that of the anti-reflection film 26, explanation of the configuration of the anti-reflection film 28 is omitted. FIGS. 3A to 3D are figures for explaining a manufacturing method of the radiation detecting element 20. FIG. 3A shows a step where the substrate 22 and the fluorescent screen 24 are prepared for respectively. In this state, for example, the thickness of the substrate 22 is 3 mm, and the thickness of the fluorescent screen 24 is 1 mm. It should be noted that the thicknesses of the substrate 22 and the fluorescent screen 24 before bonding may be determined as appropriate considering easiness of handling or the like. FIG. 3B shows a step where the substrate 22 and the fluorescent screen 24 are bonded by solid state diffusion. The bonding surface of the substrate 22 is superposed on the bonding surface of the fluorescent screen 24, and bonded by solid state diffusion. It should be noted that, before superposing the substrate 22 and the fluorescent screen 24, the surface of the substrate 22 on which it is bonded with the fluorescent screen 24 may be polished and smoothed. Similarly, the surface of the fluorescent screen 24 on which it is bonded with the substrate 22 may be polished and smoothed. By smoothing the bonding surfaces, the contact area of the substrate 22 and the fluorescent screen 24 can be increased. Thereby, the reliability of solid state diffusion bonding can be improved. Also, before superposing the substrate 22 and the fluorescent screen 24, the surface of the substrate 22 on which it is bonded with the fluorescent screen 24 may be washed. Similarly, the surface of the fluorescent screen 24 on which it is bonded with the substrate 22 may be washed. In solid state diffusion bonding, if the bonding surfaces of the substrate 22 and the fluorescent screen 24 are contaminated, diffusion of respective atoms does not proceed on the contaminated part, and the bonding strength of solid state diffusion bonding becomes low. By washing the bonding surfaces, contamination can be removed, thereby improving the reliability of solid state diffusion bonding. Also, in solid state diffusion bonding, pressure may be applied in the bonding directions of the substrate 22 and the fluorescent screen 24, respectively. By applying pressure in the bonding directions of the substrate 22 and the fluorescent screen 24, the interfaces of the substrate 22 and the fluorescent screen 24 can be closely adhered to each other. Thereby, the reliability of solid state diffusion bonding can be improved. Furthermore, in solid state diffusion bonding, the substrate 22 and the fluorescent screen 24 may be heated. By heating the substrate 22 and the fluorescent screen 24, diffusion of atoms of the substrate 22 and atoms of the fluorescent screen 24 at the bonding surfaces is enhanced. Thereby, the bonding strength of solid state diffusion bonding can be improved. FIG. 3C shows a state where the fluorescent screen 24 is polished and thinned. The fluorescent screen 24 is polished and thinned from a surface on a side on which the substrate 22 is not bonded. It should be noted that polishing is performed for example by chemical mechanical polishing (hereinafter, called CMP in some cases), and the fluorescent screen 24 of 1-mm thickness is thinned to 20 μm. It should be noted that the thickness of the fluorescent screen 24 may be 4 μm, 2 μm, or 1 μm. Also, polishing may be mechanical polishing such as machining. Because the fluorescent screen 24 is reinforced by being bonded with the substrate 22 of 3-mm thickness, the fluorescent screen 24 can be thinned to the thickness of 1 μm without damaging the fluorescent screen 24. Also, in this manner, by thinning the fluorescent screen 24 from the side where radiation enters by CMP, the surface of the fluorescent screen 24 on the side where radiation enters can be flattened highly accurately. FIG. 3D shows a step where the anti-reflection film 26 and the anti-reflection film 28 are provided to both end faces of the bonded fluorescent screen 24 and substrate 22. As described above, the anti-reflection film 26 is disposed on the side surface of the substrate 22 on the surface of the side on which the fluorescent screen 24 is not provided. Also, the anti-reflection film 28 is disposed on the side surface of the fluorescent screen 24 on the surface of the side on which the substrate 22 is not provided. The anti-reflection film 26 and the anti-reflection film 28 are disposed on the respective surfaces for example due to vacuum deposition. FIG. 4 is a schematic cross-sectional view of a spatial resolution measuring apparatus 40. The spatial resolution of the radiation detecting element 20 according to the present embodiment is explained by using FIG. 4. The spatial resolution measuring apparatus 40 has a pinhole plate 42 to which a φ10-μm pinhole is provided, the radiation detecting element 20, the objective lens 16 and the CCD sensor 14. It should be noted that in FIG. 4, elements that are the same as those shown in FIG. 1 are provided with the same reference numerals, and overlapping explanation is omitted. For measurement of the spatial resolution, first, an XFEL was allowed to pass through the φ10 μm pinhole provided to the pinhole plate 42. Thereby, the XFEL that entered the radiation detecting element 20 was cut into φ10 μm. Next, the XFEL cut into φ10 μm was entered into the radiation detecting element 20, and fluorescence was emitted. An image of the fluorescence emitted from the radiation detecting element 20 was formed in the CCD sensor 14 by using the objective lens 16. Thereby, the profile of a point spread function of the fluorescence the image of which was formed by the objective lens 16 was acquired from the CCD sensor 14. Next, the profile of a point spread function of the XFEL cut into φ10 μm at the position of the radiation detecting element 20 was acquired. Then, the profile of the point spread function of the XFEL cut into φ10 μm at the position of the radiation detecting element 20 was deconvoluted from the profile of the point spread function acquired from the CCD sensor 14, and thus the spatial resolution of the optical system configured by the radiation detecting element 20 and the objective lens 16 was calculated. FIG. 5 shows the profile of the point spread function of fluorescence acquired from the CCD sensor 14. The vertical axis in FIG. 5 indicates the numbers of counts of received fluorescence, and the horizontal axis indicates pixel positions where the received fluorescence was received. The half-value width calculated from the profile of the point spread function acquired from the CCD sensor 14 was 12 μm. The half-value width calculated from the profile of the point spread function the XFEL cut into φ10 μm at the position of the radiation detecting element 20 was 9 μm. Here, the profile of the point spread function acquired from the CCD sensor 14 is assumed to be A and its half-value width is assumed to be a. Also, the profile of the point spread function of the XFEL cut into φ10 μm at the position of the radiation detecting element 20 is assumed to be B, and its half-value width is assumed to be b. Also, the profile of the point spread function from the radiation detecting element 20 to the CCD sensor 14 is assumed to be C, and its half-value width is assumed to be c. Then, because the profile A is a profile obtained by convolution of the profile B and the profile C, the following relational expression (1) is established when these profiles are deconvoluted and they are respectively expressed with the half-value widths of the profiles.a=[(b)2+(c)2]1/2 (1) By using the relational expression (1) and substituting 12 for a and 9 for b to calculate the half-value width c, the half-value width c=7.9 is obtained. Thereby, it can be known that the spatial resolution from the radiation detecting element 20 to the CCD sensor 14 is about 8 μm in terms of a half-value width. Also, even when measurement of spatial resolution was performed for about six hours by using the spatial resolution measuring apparatus 40, there was no malfunction of the CCD sensor 14. This indicates that the XFEL was attenuated by the substrate 22 and the XFEL did not reach the CCD sensor 14. It should be noted that if an XFEL reaches the CCD sensor 14, malfunction that the XFEL destroys the photoelectric conversion elements and the operation of the CCD sensor 14 stops or dark current of the CCD sensor 14 increases and noise increases occurs. FIG. 6 is a schematic diagram of a part that detects radiation in the radiation detecting apparatus 10. Effects of the radiation detecting apparatus 10 and the radiation detecting element 20 according to the present embodiment are explained by using FIG. 6. It should be noted that in FIG. 6, elements that are the same as those shown in FIG. 1 are provided with the same reference numerals, and overlapping explanation is omitted. In the present embodiment, the substrate 22 and the fluorescent screen 24 in the radiation detecting element 20 are configured by the same material, YAG. Furthermore, because the substrate 22 and the fluorescent screen 24 are bonded while maintaining continuity of the refractive index, fluorescence emitted from the fluorescent screen 24 is not refracted at the interface between the substrate 22 and the fluorescent screen 24. If it is supposed that fluorescence is refracted at the interface between the substrate 22 and the fluorescent screen 24, the refracted fluorescence is not focused on the CCD sensor 14 by the objective lens 16 due to the influence of spherical aberration. Accordingly, by bonding the interface between the substrate 22 and the fluorescent screen 24 while maintaining continuity of the refractive index, refraction of fluorescence at the bonding interface can be prevented, thereby improving resolution of an image formed on the CCD sensor 14 by the objective lens 16. Also, by bonding the substrate 22 and the fluorescent screen 24 while maintaining continuity of the refractive index, fluorescence emitted from the fluorescent screen 24 can also be prevented from being reflected at the bonding interface between the fluorescent screen 24 and the substrate 22. If fluorescence is reflected at the interface, the light amount of an image formed on the CCD sensor 14 by the objective lens 16 decreases. Accordingly, the light amount of an image formed on the CCD sensor 14 can be increased by suppressing reflection at the interface, and the sensitivity of the radiation detecting apparatus 10 can be increased. Also, if fluorescence is reflected at the bonding interface between the fluorescent screen 24 and the substrate 22, the resolution of the radiation detecting apparatus deteriorates. Deterioration of the resolution of the radiation detecting apparatus due to the reflection is explained by using FIG. 7. FIG. 7 is a schematic diagram of a part that detects radiation in a radiation detecting apparatus 50 according to a comparative example. The radiation detecting apparatus 50 has a radiation detecting element 52, the objective lens 16 and the CCD sensor 14. Also, the radiation detecting element 52 is configured by bonding a fluorescent screen 54 and a substrate 56 by an adhesive 58. Differences between the radiation detecting apparatus 50 and the radiation detecting apparatus 10 shown in FIG. 1 are that the radiation detecting element 52 is configured by gluing the fluorescent screen 54 and the substrate 56 with the adhesive 58, the thickness of the fluorescent screen 54 is thick, and the anti-reflection films 26, 28 are not provided. Generally, the refractive index of the adhesive 58 is different from the refractive index of the fluorescent screen 54 and the refractive index of the substrate 56. When the refractive index of the adhesive 58 is different from the refractive index of the fluorescent screen 54, fluorescence is reflected due to the change in the refractive index at the interface between the fluorescent screen 54 and the adhesive 58. Similarly, when the refractive index of the adhesive 58 is different from the refractive index of the substrate 56, fluorescence is reflected due to the change in the refractive index at the interface between the adhesive 58 and the substrate 56. The reflected light is further reflected on the front surface of the fluorescent screen 54, and enters the objective lens 16. Because the reflected light that has entered the objective lens 16 is not focused on the CCD sensor 14, components that are not focused on the CCD sensor 14 are mixed in an image formed on the CCD sensor 14, and the resolution of the image formed on the CCD sensor 14 deteriorates. The reflected light repeats reflection at constant reflectance at each of the interface between the fluorescent screen 54 and the air, the interface between the fluorescent screen 54 and the adhesive 58, the interface between the adhesive 58 and the substrate 56, and the interface between the substrate 56 and the air, and spreads in directions that are vertical to the optical axis. Components that are transmitted through the interface between the substrate 56 and the air in this process enter the objective lens 16. Because fluorescent components that repeated the reflection and have entered the objective lens 16 are not focused on the CCD sensor 14, components that are not focused on the CCD sensor 14 are mixed in an image formed on the CCD sensor 14, and the resolution of the image formed on the CCD sensor 14 deteriorates. On the other hand, because in the radiation detecting element 20 according to the present embodiment, the substrate 22 and the fluorescent screen 24 are bonded by solid state diffusion, the refractive index does not change at the interface between the substrate 22 and the fluorescent screen 24. Thereby, mixing of components that are not focused on the CCD sensor 14 can be prevented, and the resolution of an image formed on the CCD sensor 14 by the objective lens 16 can be prevented from deteriorating. FIG. 8 is a schematic diagram for explaining a light-emission position of fluorescence in the radiation detecting apparatus 50 in FIG. 7. When radiation enters the fluorescent screen 54, fluorescence is emitted from a plurality of positions that are displaced in the thickness direction of the fluorescent screen 54. Images of the fluorescence emitted from the plurality of positions displaced in the thickness direction are formed by the objective lens 16 at different positions. Accordingly, by adjusting the position of the objective lens 16 so that an image of fluorescence emitted from a certain position from among the fluorescence emitted from the plurality of positions is formed on the CCD sensor 14, fluorescence emitted from other positions become components that are not focused on the CCD sensor 14, and the resolution of an image formed on the CCD sensor 14 deteriorates. In the radiation detecting apparatus 50 shown in FIG. 8, the position of the objective lens 16 is adjusted so that fluorescence emitted from a position close to a side surface where radiation enters is imaged on the CCD sensor 14. In this case, fluorescence emitted from a position that is far from the side surface where radiation enters is formed at a position that is farther from the objective lens 16 than the position of the CCD sensor 14. Thereby, the resolution of the radiation detecting apparatus 50 deteriorates. In the radiation detecting element 20 according to the present embodiment, scattering of light-emission points of fluorescence in the traveling direction of radiation can be prevented by reducing the thickness of the fluorescent screen 24. For example, by making the thickness of the fluorescent screen 24 1 μm, the spatial resolution of the radiation detecting element 20 can be made equal to the spatial resolution determined by the diffraction limit of fluorescence emitted by the fluorescent screen 24. The spatial resolution and sensitivity of the radiation detecting apparatus 10 are in a trade-off relationship with the thickness of the fluorescent screen 24. When the thickness of the fluorescent screen 24 is small, the spatial resolution increases whereas the amount of emitted light of fluorescence is reduced. For example, if the thickness of the fluorescent screen 24 is 300 μm, 99.4% of a signal of an X-ray with photon energy of 10 KeV can be detected by the fluorescent screen. If the thickness of the fluorescent screen 24 is 20 μm, 29% of a signal of an X-ray with photon energy of 10 KeV can be detected by the fluorescent screen. If the thickness of the fluorescent screen 24 is 1 μm, 1.7% of a signal of an X-ray with photon energy of 10 KeV can be detected by the fluorescent screen. The problem of the detection ratio of an X-ray signal becoming low can be remedied by changing the material used for the fluorescent screen 24 to a material having a greater atomic number or a denser material. Also, in the present embodiment, the surface of the fluorescent screen 24 where radiation enters is flattened highly accurately in the process of thinning the fluorescent screen 24 by CMP. Thereby, scattering of an entering radiation on the front surface of the fluorescent screen 24 can be suppressed, and the sensitivity and spatial resolution of the radiation detecting apparatus 10 can be increased. Also, in the present embodiment, the thickness of the substrate 22 is 3 mm, which thickness attenuates radiation that has been transmitted through the fluorescent screen 24 and does not allow passage of the radiation. In a conventional apparatus, fluorescence is reflected on a mirror and allowed to enter the CCD sensor 14 for the purpose of preventing exposure of the objective lens 16 and the CCD sensor 14 to radiation. However, because in the present embodiment, radiation is extinguished at the substrate 22 having high radiation resistance, it does not reach the objective lens 16 or the CCD sensor 14. For this reason, the radiation detecting apparatus 10 according to the present embodiment can form an image of fluorescence on the CCD sensor 14 by using the objective lens 16 without providing a mirror. Thereby, the radiation detecting apparatus 10 can be downsized, the design freedom of the optical system of the radiation detecting apparatus 10 can be improved, and furthermore, cost reduction can be realized. Furthermore, the objective lens 16 can be arranged close to the radiation detecting element 20 by not providing a mirror. Thereby, the working distance which is a distance between the objective lens 16 and the radiation detecting element 20 can be decreased, and the numerical aperture of the objective lens 16 can be increased. Thereby, the sensitivity and resolution of the radiation detecting apparatus 10 can be increased. It should be noted that even in a radiation detecting apparatus in which a mirror is installed between the objective lens 16 and the CCD sensor 14 and the incident direction of fluorescence on the CCD sensor 14 is set to be vertical to the radiation incident direction, the radiation detecting element 20 according to the present embodiment can be used. Thereby, the width of the radiation detecting apparatus can be reduced. Furthermore, even when the working distance is increased in order to attain low magnification and wide viewing field of the radiation detecting apparatus, this can be realized, without increasing the width of the radiation detecting apparatus, by installing a mirror between the objective lens 16 and the CCD sensor 14 and making the incident direction of fluorescence toward the CCD sensor 14 vertical to the radiation incident direction. Also, in the present embodiment, the thickness of the substrate 22 is decreased as much as possible while ensuring that radiation transmitted through the fluorescent screen 24 is not transmitted through the substrate 22. When the substrate 22 is made thin, because the optical path within the substrate 22 through which fluorescence emitted from the fluorescent screen 24 is transmitted becomes short, the spherical aberration due to the substrate 22 can be reduced. Because the spherical aberration causes blurring and distortion in an image formed, the resolution of the radiation detecting apparatus 10 can be increased by reducing the spherical aberration. Furthermore, for example, the spherical aberration can be corrected by providing a correction collar to the objective lens 16. It should be noted that the spherical aberration can be completely corrected by a commercially available objective lens with a correction collar by making the substrate 22 about 500 μm or smaller. Also, the substrate 22 can be thinned up to 300 μm only by means of normal mechanical processing/polishing. Furthermore, the objective lens 16 can be brought close to the radiation detecting element 20 by making the substrate 22 thin. By further decreasing the working distance, the numerical aperture of the objective lens 16 can be further increased. Thereby, the sensitivity and resolution of the radiation detecting apparatus 10 can be further increased. Also, because the anti-reflection film 26 is provided in the present embodiment, fluorescence emitted from the fluorescent screen 24 can be prevented from being reflected at the interface between the substrate and the air. Furthermore, fluorescence emitted from the fluorescent screen 24 in a direction in which radiation enters can be prevented from being reflected at the interface between the fluorescent screen 24 and the air by providing the anti-reflection film 28. The reflected light becomes components that are not focused on the CCD sensor 14, and the resolution deteriorates thereby. For this reason, by providing at least one of the anti-reflection film 26 and the anti-reflection film 28, occurrence of the reflected light can be suppressed, and deterioration of the resolution of the radiation detecting apparatus 10 can be suppressed. FIG. 9 shows an example of an image captured with the radiation detecting apparatus 10. The image shown in FIG. 9 is an X-ray photograph of an ant captured by using the radiation detecting apparatus 10 that comprises the radiation detecting element 20 in which the fluorescent screen 24 of 10-μm thickness and the substrate of 2.990-mm thickness are bonded by solid state diffusion. The radiation detecting apparatus 10 has the spatial resolution of 8 μm. The X-ray photograph of the ant is captured by using a one-time 2-mm square beam of SPring-8. Background image difference processing and standardization processing of X-ray intensity in the unit of pixel have been performed on the image that was captured and generated. It can be seen in the X-ray photograph of the ant shown in FIG. 9 that the radiation detecting apparatus 10 can capture an image of the appearance of the feelers, skeleton structure inside feet, and internal organs of the abdominal region of the ant at a high contrast. While the embodiments of the present invention have been described, the technical scope of the invention is not limited to the above described embodiments. It is apparent to persons skilled in the art that various alterations and improvements can be added to the above-described embodiments. It is also apparent from the scope of the claims that the embodiments added with such alterations or improvements can be included in the technical scope of the invention. The operations, procedures, steps, and stages of each process performed by an apparatus, system, program, and method shown in the claims, embodiments, or diagrams can be performed in any order as long as the order is not indicated by “prior to,” “before,” or the like and as long as the output from a previous process is not used in a later process. Even if the process flow is described using phrases such as “first” or “next” in the claims, embodiments, or diagrams, it does not necessarily mean that the process must be performed in this order.
	abstract	A device for producing electricity. In one embodiment the device comprises a germanium substrate doped a first dopant type and a plurality of stacked material layers above the substrate. These stacked material layers further comprise an InGaP base layer doped the first dopant type, an InGaP emitter layer doped the second dopant type, a window layer having a lattice structure matched to the lattice structure of the emitter layer and doped the second dopant type and a beta particle source for generating beta particles.
	abstract	Surfaces of the structural components of a nuclear power plant exposed to reactor water are wetted with an electroless plating solution containing an electrical insulating substance. The electrical insulating substance has a high resistivity about 105 (100000) times those of the structural components of the nuclear power plant or above. A metal film containing the electrical insulating substance is formed on the surfaces of the structural components exposed to the reactor water by wetting the surfaces of the structural components with the electroless plating solution. Thus, the electrochemical corrosion potential of the structural components is reduced regardless of whether hydrogen is injected into the reactor water.
	claims	1. A method of providing an irradiation target positioning system, the method comprising:determining an irradiation target and a daughter product produced from the irradiation target;determining physical characteristics of a radiation field to which the irradiation target will be exposed;configuring the irradiation target, an irradiation target plate, and a target plate holder to produce the daughter product when the irradiation target is loaded in the irradiation target plate and the target plate holder in the radiation field; anddetermining a quantity and a location of the irradiation target relative to a center of the irradiation target plate such that the quantity and the location of the irradiation target varies based on the physical characteristics of the radiation field. 2. The method of claim 1, further comprising:loading the irradiation target into the irradiation target plate and the target plate holder; andirradiating the irradiation target loaded in the irradiation target plate and the target plate holder in the radiation field so as to produce the daughter product. 3. The method of claim 2, wherein the radiation field is a neutron flux including thermal neutrons produced in a light-water reactor. 4. The method of claim 2, wherein the configuring includes providing at least one of,a shape, size, and known absorption cross-section for the irradiation target,a constant position of the irradiation target in the radiation field to be maintained by the irradiation target plate and the target plate holder, andmaterials for the irradiation target plate and the plate holder with known absorption cross-sections for the radiation field. 5. The method of claim 1, wherein the physical characteristics of the radiation field include at least one of radiation type and radiation energy distribution over position. 6. The method of claim 1, wherein the irradiation target is fabricated from a material including at least one of cobalt (Co), chromium (Cr), copper (Cu), erbium (Er), germanium (Ge), gold (Au), holmium (Ho), iridium (Ir), lutetium (Lu), molybdenum (Mo), palladium (Pd), samarium (Sm), thulium (Tm), ytterbium (Yb), and yttrium (Y). 7. The method of claim 1, wherein the configuring includes providing at least one loading position in the target plate for the irradiation target. 8. The method of claim 7, wherein the configuring further includes defining a hole in the target plate at each loading position, the hole configured to retain the irradiation target in the target plate. 9. The method of claim 8, wherein the configuring further includes placing at least one target spacing element in the hole so as to maintain the irradiation target at a constant position within the loading position. 10. The method of claim 8, wherein the configuring further includes placing at least one spacer plate in the target plate holder so as to maintain the target plate and at least one loading position at the constant position within the radiation field. 11. The method of claim 10, wherein the at least one spacer plate is placed adjacent to the target plate in the target plate holder so as to retain the irradiation target at the constant position. 12. The method of claim 1, wherein the configuring includes providing the irradiation target plate with a first planar surface and an opposing second planar surface, a distance between the first and second planar surfaces being less than a diameter or length of the irradiation target plate. 13. The method of claim 1, wherein the configuring includes providing the irradiation target plate with an opening extending through the center and a plurality of holes surrounding the opening, the opening being larger than each of the plurality of holes. 14. The method of claim 13, wherein the configuring includes inserting the target plate holder through the opening of the irradiation target plate.
050826204	claims	1. A recirculation system for a boiling water reactor having a cylindrical shroud surrounding a reactor core and spaced radially inwardly from a pressure vessel to define an annular downcomer for channeling downwardly a recirculation reactor coolant into an inlet of the core disposed at a lower plenum of the vessel comprising: an annular pump deck disposed in said downcomer and fixedly joined to said pressure vessel and said core shroud; a plurality of circumferentially spaced impeller-driven reactor internal pumps (RIPs) disposed in said downcomer and joined to said pump deck for pumping a first portion of said coolant in said downcomer downwardly through said pump deck and into said lower plenum as RIP discharge flow to said core inlet; and a plurality of circumferentially spaced fluid-driven jet pumps (JPs) disposed in said downcomer and joined to said pump deck for pumping a second portion of said coolant in said downcomer downwardly through said pump deck and into said lower plenum as JP discharge flow to said core inlet in parallel flow with said RIP discharge flow. a nozzle for ejecting a JP driving fluid as a jet; an inlet for receiving said JP driving fluid jet and said downcomer coolant second portion; a mixer for mixing said jet and said coolant second portion; a diffuser for diffusing said mixed jet and coolant second portion; an outlet for discharging said mixed and diffused jet and coolant second portion through said pump deck and into said lower plenum as said JP discharge flow; and means for supplying said JP driving fluid to said nozzle. a feedwater driving pump for pressurizing said feedwater; a flow control valve disposed in flow communication with said feedwater pump; a feedwater sparger disposed inside said pressure vessel and in flow communication with said flow control valve; and said flow control valve being disposed in flow communication with said JP nozzles in parallel flow with said feedwater sparger. 2. A recirculation system according to claim 1 wherein said RIPs and said JPs have diverse motive power sources for preventing simultaneous tripping thereof. 3. A recirculation system according to claim 2 wherein each of said RIPs has an RIP flow inertia, and each of said JPs has a JP flow inertia greater than said RIP flow inertia. 4. A recirculation system according to claim 3 wherein each of said JPs includes: 5. A recirculation system according to claim 4 wherein said JP nozzle and inlet are disposed above said reactor core. 6. A recirculation system according to claim 4 wherein said JP driving fluid is steam. 7. A recirculation system according to claim 6 wherein said steam supplying means are disposed in flow communication with a steam outlet nozzle of said boiling water reactor for receiving a portion of steam discharged from said boiling water reactor for flow to said JP nozzle as said JP driving fluid. 8. A recirculation system according to claim 7 wherein said steam supplying means include means for reheating said steam from said outlet nozzle for raising the temperature of said steam for flow to said JP nozzle. 9. A recirculation system according to claim 4 wherein said JP driving fluid is feedwater. 10. A recirculation system according to claim 9 wherein said feedwater supplying means include: 11. A recirculation system according to claim 10 wherein said feedwater driving pump is sized for effecting said JP flow inertia greater than said RIP flow inertia. 12. A recirculation system according to claim 10 wherein in a trip of said boiling water reactor, said JPs are effective for maintaining and increasing water level in said pressure vessel while said RIPs are effective for decreasing total flow of said reactor coolant into said core inlet for increasing steam void fraction in said core for decreasing reactivity of said core.
	summary
	claims	1. A system for monitoring a state of a reactor core included in a nuclear reactor, the system comprising:an internal monitoring device located inside the reactor core, the internal monitoring device including,one or more internal sensor arrays each configured to take measurements of conditions at each of a plurality of different vertical regions within the reactor core to generate internal measurement data, the internal monitoring device being configured to provide the internal measurement data to one or more data lines of the nuclear reactor, the one or more internal sensor arrays being configured to take measurements of a conductivity at each of the plurality of different vertical regions, the one or more internal sensor arrays including a conductivity sensor array, the conductivity sensor array including a plurality of conductivity sensors; anda first case, at least a portion of the one or more internal sensor arrays being located within the first case, the first case including a plurality of protrusions, the plurality of protrusions extending outwards above each of the plurality of conductivity sensors, and the plurality of protrusions configured to substantially mitigate coating of each of the plurality of conductivity sensors by molten materials;an external monitoring device located in the reactor structure outside the reactor core, the external monitoring device including one or more external sensor arrays each configured to take measurements of conditions of the reactor core at positions outside the reactor core corresponding to each of the plurality of different vertical regions within the reactor core to generate external measurement data, and a transmitter configured to wirelessly transmit the external measurement data outside the nuclear reactor; anda receiver station configured to receive the external measurement data and the internal measurement data, and to determine a state of the reactor core based on the received external and internal measurement data,wherein the one or more internal sensor arrays are further configured to take measurements of a temperature at each of the plurality of different vertical regions. 2. The system of claim 1, wherein the first case is made from a rigid material having a melting point higher than that of stainless steel. 3. The system of claim 2, wherein the first case is made of at least one of tungsten, molybdenum, niobium and silicon carbide. 4. The system of claim 2, wherein the one or more internal sensor arrays includes a temperature sensor array including a plurality of temperature sensors each corresponding to one of the plurality of different vertical positions within the reactor core. 5. The system of claim 2, whereineach of the plurality of conductivity sensors extends to an outer surface of the first case, and each of the plurality of conductivity sensors corresponds to one of the plurality of different vertical regions within the reactor core. 6. The system of claim 1, wherein the internal monitoring device is located inside a traversing in-core probe (TIP) tube of the reactor core. 7. The system of claim 1, wherein the internal monitoring device further includes a first data processor configured to generate the internal measurement data by digitizing the measurements taken by the one or more internal sensor arrays. 8. The system of claim 1, wherein the internal monitoring device includes a power line configured to receive power from a source external to the internal monitoring device. 9. The system of claim 1, wherein at least one of the one or more external sensor arrays is configured to take measurements of at least one of a gamma flux and a neutron flux of the reactor core. 10. The system of claim 9, wherein the external monitoring device further includes a second case made of at least one of stainless steel, tungsten, molybdenum, niobium and, silicon carbide, at least a portion of the one or more external sensor arrays being located within the second case so as to be enclosed by the second case. 11. The system of claim 9, wherein the one or more external sensor arrays includes a sensing unit array including a plurality of sensing units, each of the plurality of sensing units being configured to measure at least one of a gamma flux and a neutron flux, each of the plurality of sensing units corresponding to one of the plurality of different vertical regions within the reactor core. 12. The system of claim 9, wherein the external monitoring device further includes a power unit configured to power the external monitoring device independently of any external power source. 13. The system of claim 9, wherein the external monitoring device further includes a coupling unit configured to affix the external monitoring device to a surface inside the nuclear reactor, the coupling unit including at least one of a magnet, an adhesive, and a bolting mechanism. 14. The device of claim 9, wherein the external monitoring device further includes a controller configured to,control an operation mode of the external monitoring device to be one of an active mode and a stand-by mode based on the external measurement data,wherein the external monitoring device is configured to operate such that less power is used in the stand-by mode than in the active mode. 15. The device of claim 1 wherein the receiver station includes a data processing unit configured to determine conditions in the reactor core including at least one of cladding oxidation, water level, cladding melting, nuclear reactivity keff and temperature, based on at least one of the external measurement data and the internal measurement data. 16. A device for monitoring a reactor core or a nuclear reactor, the device comprising:a case made from a rigid material having a melting point higher than that of stainless steel, the case being configured to fit inside an internal tube of the reactor core;one or more sensor arrays inside the case, at least one of the one or more sensory arrays being configured to take measurements of at least one of a temperature and a conductivity of the reactor core at each of a plurality of different vertical regions within the reactor core to generate internal measurement data; andone or more internal data lines configured to transfer the internal measurement data to data lines of the nuclear reactor,the one or more sensor arrays includes a conductivity sensor array including a plurality of conductivity sensors, each of the plurality of conductivity sensors extending to an outer surface of the case, each of the plurality of conductivity sensors corresponding to one of the plurality of different vertical regions within the reactor core, andthe case includes a plurality of protrusions extending outwards above each of the plurality of conductivity sensors, the plurality of protrusions configured to substantially mitigate coating of each of the plurality of conductivity sensors by molten materials. 17. The system of claim 16, wherein the case is made of at least one of tungsten, molybdenum, niobium and silicon carbide. 18. The device of claim 16, wherein the case is configured to fit inside a traversing in-core probe (TIP) tube of the reactor core.
059189118	summary	BACKGROUND OF THE INVENTION The invention disclosed herein relates to nozzles, nozzle assemblies, and nozzle repair assemblies for vessels and piping that are installed either initially or as replacements without any welding at all to the vessel. (A "nozzle" may be, or include as part thereof, a sleeve and/or, piping. A "vessel" may also be large bore piping.) The invention more particularly relates to nozzles, nozzle assemblies, nozzle repair assemblies and procedures which replace or repair nozzles that are attached to the vessel on the inside diameter of the vessel with a J groove structural weld. (The term "nozzle assembly" is sometimes interchangeably used herein with the term "partial nozzle assembly" which encompasses an assembly that includes part of an existing nozzle and a partial replacement nozzle or the term "full nozzle assembly" which encompasses an assembly that includes a full replacement nozzle. Both partial and full nozzle assemblies in addition may include components such as gaskets, packing glands, a spring, and anti-rotation and leak path devices etc. These assemblies provide a mechanical attachment of a nozzle to a vessel, a mechanical seal of a nozzle, or both. The term "nozzle repair assembly" encompasses an existing nozzle and a structure which mechanically seals the existing nozzle or attaches the existing nozzle to the vessel without welding to the vessel, or both.) The invention has particular application to nozzles, nozzle assemblies, nozzle repair assemblies and procedures which replace, repair or initially install nozzles in ASME pressure vessels and large bore piping of pressurized water reactor (PWR) nuclear power facilities, either because a nozzle has failed, or to prevent future failure or leakage due to a phenomenon known as Primary Water Stress Corrosion Cracking, PWSCC. Repair of a nozzle also encompasses plugging the bore, for example, replacing a heater with a plug. Further, the invention relates to replacement and repair of nozzles and assemblies that were previously replaced or repaired. A typical PWR nuclear power generating facility includes in part a reactor vessel, steam generator, pressurizer vessel, and a reactor coolant piping system, all of which operate under high pressure. Nozzles are attached to the vessels and/or piping for a number of purposes, e.g., for connecting piping and instrumentation, vents, and to secure control element drive mechanisms and heater elements. A typical pressurizer vessel 20 is shown in FIG. 1 with nozzles 22 for vents, nozzles 24 for sample or liquid level, nozzles 25 (one of which is shown behind nozzle 24) for pressure sensing, a nozzle 26 for temperature measuring, and a number of nozzles 27 for heating elements. All of those nozzles were heretofore welded to the pressurizer vessel at the time of original manufacture. As shown in FIG. 2, inconel or stainless cladding 29 is welded to the interior of the pressurizer vessel which is made of carbon steel. The nozzle 26 shown in 5 cross section in FIG. 2, which is exemplary of the welded nozzles 22 and 24-27, passes through a hole or bore 30 in the pressurizer vessel 20 and is structurally welded at its interior end 32 to the vessel 20 with a J-groove weld 34 along the interior opening to the bore 30. The diameter of nozzle 32 is slightly less than the diameter of bore 30, so that there is a small annular space 36 between the nozzle exterior and the wall of bore 30. In some applications the nozzles are fit tight to the bore, and in a control rod drive mechanism, they are installed with a shrink fit process. The J-groove weld 34 also functions as a seal weld to seal the annular space 36. A reactor vessel (not shown) similarly has nozzles represented by nozzle 26 in FIG. 2 welded thereto. The piping of the reactor coolant system (not shown) also includes similar nozzles welded thereto. Further details of pressurizer vessels, reactor vessels, and coolant system piping, in particular, and nuclear power facilities, in general, are known to those of skill in the art. As mentioned, the invention has particular application to the replacement of nozzles which have failed, and to the prevention of nozzle failures and leakage in nuclear power facilities due to the PWSCC phenomenon, which occurs on components having a susceptible material, high tensile stresses, high temperature and which are in a corrosive environment, conditions which primarily exist on nozzle penetrations in the pressurizer vessel, reactor coolant piping, and the reactor vessel. Such failures are manifested by cracking, which the applicant recognized resulted from several contributing factors. Such cracking occurs at the grain boundaries on the inside diameter of the nozzle material (Alloy 600) at or near the heat affected zone of the weld and propagates radially outward through the thickness of the nozzle which eventually leads to small leakage of the reactor coolant supply. Recently, failures have also occurred on stainless steel pressurizer nozzles though the failure mechanism is not known. As indicated, nozzles of these types have failed over time and have had to be replaced or repaired, either because of a failure in the nozzle or the weld attaching and sealing the nozzle to the vessel. A typical replacement procedure in a nuclear power plant environment requires shutting down the nuclear power plant, removing the nozzle, which typically requires machining operations, and welding a replacement nozzle to the vessel or piping. The welded replacement nozzles currently in use closely duplicate the original welded nozzle they replace, except that they may be made of a different alloy, e.g., Alloy 690 (less susceptible to PWSCC) instead of Alloy 600, and may also be represented by the nozzle shown in FIG. 2. Other weld repair methods involve installing a thick weld pad on the outside of the vessel and structurally welding the nozzle to the pad, and seal welding the interior end of the nozzle to the vessel. Other refinements to this method involve inserting a thin sleeve between the nozzle and vessel, seal welding the thin sleeve on the inside diameter of the vessel, and structurally welding the thin sleeve and nozzle on the outside of the vessel. This methodology promotes less weld residual stress and eliminates the axial thermal stresses imposed since the nozzle is now free ended and allowed to grow thermally when the vessel heats up. Still another derivative of the weld pad designs is the use of a partial nozzle instead of a full nozzle. With this approach, only a portion of the nozzle is removed leaving the cracked portion of the nozzle remaining in place. A partial nozzle is structurally welded to the weld pad leaving approximately a 1/16" to 1/8" gap between the existing and new nozzle. A final derivative of the weld pad is replacing a portion of the existing heater nozzle, otherwise known as a heater sleeve, with a plug instead of a partial nozzle or sleeve, which is used due to the inherent difficulties of aligning the replacement nozzle such that the heaters can be reinstalled. Therefore, the failed heater location is abandoned in place with the plug preventing further leakage. Also, a heater sleeve may be plugged because a replacement for the failed heater is not available during the repair period, typically a scheduled outage. The heater will then be replaced during next scheduled outage. The above partial nozzle replacement and plugs are normally used as last resorts since the carbon steel vessel is exposed to reactor coolant containing boric acid which can corrode the vessel and since the failed nozzle can continue to crack with the crack propagating into the base material of the vessel. Boric acid corrosion, though a concern, has not proven to corrode the vessel bore to any sufficient depth for at least one nozzle installed in the steam space in the pressurizer, which is in a low oxygen, stagnant environment. However, corrosion could be a concern for those areas with high fluid flow such as nozzles on reactor coolant piping. Another method of plugging a heater location, though currently not used in the industry, involves removing a portion of the existing nozzle, inserting a new nozzle with split flared ends that compress during insertion and expand once the nozzle is fully inserted such that the flared ends overlap the inboard end of the existing nozzle relying on the integrity of the existing nozzle and weld to prevent ejection of the replaced nozzle. Once the split flared nozzle is installed, a portion of the nozzle is then rolled with high compressive load expanding the rolled area into the vessel bore, thereby forming a seal therebetween. After the rolled joint is formed a steel rod is inserted through the nozzle past the split flared ends, which prevents them from compressing inward, and a seal weld is applied at the outside diameter of the vessel at the nozzle and rod interface. The inherent problem with this plug is the rolled joint. The rolling process causes a compressive stress layer in the joint itself, but at the interface between the rolled and non rolled area, high tensile stresses are imposed which promote cracking. Cracking has been observed in the rolled transition region on similar repairs used for steam generator A600 tubes. Though this method uses A690 as the replacement material, A690 is not believed to be immune from PWSCC. Furthermore, the rolled joint may not be reliable to form a seal for extended periods because of relaxation over time in the rolled area. Any leakage here would cause severe corrosion to the outside of the vessel because as the reactor coolant leaks to the atmosphere, it flashes to steam and is in a highly oxygenated environment. Replacements employing the above-described procedures in a nuclear power plant currently require a minimum of approximately fourteen days for some types of nozzles (although some claim a minimum of seven to ten days), and are extremely expensive. Including the lost revenue resulting from plant shut-down, which may be as high as $750,000 per day, the total cost of each repair can result in several millions dollars if the repair has to be performed in a non-scheduled facility shut down. The above-described nozzle replacement and repair procedures and any other replacement and repair procedure that requires welding to the vessel not only is time consuming and therefore expensive, but also exposes repair personnel to more radiation than the replacement and repair procedures disclosed herein, particularly where the nozzle replacement or repair procedure involves personnel entering inside the vessel to perform the procedure. Also, both the original welded nozzle and the known welded replacement nozzles and methods subject the welded nozzle to high residual stresses imposed by weld shrinkage. These high residual stresses increase the susceptibility to PWSCC. Thus, the welded replacement nozzle offers no improvement over the original nozzle in terms of expected life and reduction of failures, other than any improvement that may result from use of a superior nozzle material. Although, Alloy 690 material is less susceptible to PWSCC than Alloy 600, it is not known that the change in nozzle material alone will not eliminate the possibility of nozzle failures. At least three utilities have replaced nozzles using the original design criteria, Alloy 690 material for the nozzle, and alloy 600 weld material, and have experienced failures in the weld material itself Based on this information, improved nozzle replacement and repair methods are needed. There now appears to be some recognition in the field that replacement and/or repair of nozzles in pressure vessels of nuclear power plants may proceed without structurally welding a replacement nozzle to the vessel. For example, U.S. Pat. Nos. 5,149,490 and 5,202,082 (both of Brown et al.) and U.S. Pat. No. 5,404,382 (Russ et al.) (see FIGS. 8 and 9) describe methods and apparatus for replacing a nozzle for a pressurizer vessel without structurally welding the nozzle to the vessel. Although the replacement nozzles of the '490 and '382 patents are mechanically attached to the pressurizer vessel, according to these patents welding is still required to provide the seal between the nozzle and the pressurizer vessel. Therefore, the residual stresses discussed above are imposed on the nozzle by the weld whether it be a structural weld or a seal weld, which imposes less stress than a structural weld. The '382 patent discloses with respect to FIGS. 8 and 9 that the weld seams are free from loading because the nozzle is prestressed in a such a way that axial and horizontal forces acting on the nozzle are compensated for. Here, as understood by the applicant, the purpose of the prestressing is to compress the tapered nozzle against the vessel bore to eliminate lateral movement of the nozzle and bending at the weld location. It appears that the replacement procedure disclosed in the '382 patent can only be used in those locations where the inside of the vessel can be accessed such as the reactor head. This method cannot be used in the heater locations and most likely could not be used in the reactor coolant piping (typically 28" piping located in a confined area, with a high dose rate that can not be easily shielded). In the replacement procedure and nozzle described in the '082 patent, the original welded nozzle is not fully removed, and a mechanical seal is made between the remaining cracked nozzle portion and the end of the replacement nozzle. Leaving part of the existing nozzle at the interior welded may lead to future cracking in the existing failed portion of the Alloy 600 nozzle that was not removed from the vessel. However, further cracking in the base material of the vessel would not likely result in reactor coolant leakage. Experience and analysis indicate that the base material of the vessel will only crack some minimum distance into the base material which will not jeopardize the integrity of the vessel. Boric acid corrosion for this design is less of a concern than the partial nozzle welded design discussed above due to the fact that there is not an exposed gap between the new and existing nozzle thereby limiting or eliminating boric acid from contacting the vessel. Partial nozzle designs not having sealing capabilities between the new and existing nozzle can be justified readily where the nozzles are located in stagnant region of the pressurizer vessel, e.g. the liquid space or the upper steam space because there is a limited amount of bore cavity available for the stagnant boric acid to collect and oxidize between fuel cycles and because the reactant coolant has a low oxygen content. Further, if corrosion does occur, a protective oxidation layer will form which resists further corrosion in the stagnant environment. Therefore, partial nozzle replacements without a seal between the existing nozzle and the partial replacement nozzle appear to be justified for the life of the plant. For those locations in which the partial nozzle replacement without the seal is not in a stagnant environment, i.e., reactor coolant piping, additional evaluations are required to justify extended usage. Partial nozzle designs with the seal between the existing nozzle and the partial replacement nozzle can be used in stagnant environment locations and at high fluid flow locations; however, a better design practice would be to remove the cracked nozzle to eliminate further degradation of the vessel. The procedure described in the '082 patent thus has the drawback that a portion of the failed nozzle remains structurally welded to the vessel and therefore continues to subject the vessel to the same stresses as the original nozzle, although, as mentioned above, there now appears to be justification for employing partial nozzle replacements. In any event, the remaining nozzle portion and the vessel portion surrounding the bore opening are subject to further degradation. As far as the applicant is aware, the replacement nozzles and replacement procedures disclosed in the '490 and '082 patents have not been used in a nuclear power facility anywhere, and the replacement nozzles and nozzle replacement procedures disclosed in FIGS. 8 and 9 of the '382 patent have not been used in a nuclear power facility in the United States. A repair procedure has also been developed which does not use welded replacements and plugs. This repair procedure involves leaving the existing nozzle installed, installing a split flange exterior to the vessel, applying a split seal at the outside diameter of the vessel and nozzle interface, and providing an unloaded anti-ejection device. In the event the existing nozzle were to crack due to PWSCC, the seal at the outside of the vessel is intended to prevent reactor coolant leakage. This repair procedure is intended as a preventative leakage measure as opposed to an ASME code section XI type of repair or replacement. Nuclear Regulatory approval may be required before this repair procedure could be used. However, it has already been installed at one utility as a preventive measure. This methodology does not prevent PWSCC from occurring. However, without further evaluation it is unknown by the applicant if the repair is acceptable for the life of the plant. It is possible that if the nozzle began to crack at numerous locations, the crack could propagate axially from the J-groove weld to beyond the seal incorporated at the flange. However, if this occurred it would not be a significant safety concern but would require additional repairs. The following U.S. patents relate to installing, replacing, repairing or sealing nozzles, sleeves, tubes or plugs in pressure vessels and piping: U.S. Pat. No. 2,711,841 (Schaefer et al.); U.S. Pat. No. 4,255,840 (Loch et al.); U.S. Pat. No. 4,440,339 (Tamai et al.); U.S. Pat. No. 4,480,841 (Schukei et al.); U.S. Pat. No. 4,615,477 (Spada et al.); U.S. Pat. No. 4,653,483 (Margotta); U.S. Pat. No. 4,723,795 (Shenoy); U.S. Pat. No. 4,826,217 (Guerrero); U.S. Pat. No. 5,091,140 (Dixon et al.); U.S. Pat. No. 5,094,801 (Dixon et al.); U.S. Pat. No. 5,196,160 (Porowski); U.S. Pat. No. 5,209,895 (Wivagg); U.S. Pat. No. 5,271,048 (Behake et al.); U.S. Pat. No. 5,274,683 (Broda et al.); U.S. Pat. No. 5,278,878 (Porowski); U.S. Pat. No. 5,297,187 (Sodergren et al.) U.S. Pat. No. 5,323,428 (Porter et al.); and U.S. Pat. No. 5,367,768 (Weems). Also a method similar to that used in steam generator tube repairs has been proposed with certain modifications to the Nuclear Regulatory Committee for repairing a leaking nozzle. According to the proposal, the existing nozzle is rolled and deformed against the ID of the vessel such that a seal is created between the nozzle and vessel. (See U.S. Pat. No. 5,367,768 (Weems), cited above). Also, an optional A690 sleeve could be rolled into the existing nozzle forming a seal therebetween. A similar design was also proposed for a plug. However, the Nuclear Regulatory Committee declined the proposals because that rolling technique causes high tensile stresses at the rolled transition region which promotes PWSCC, and because that repair method may be only leak limiting which could allow the boric acid in the reactor coolant to erode a portion of the carbon steel vessel. Nozzles are currently being replaced or repaired in PWR nuclear power facilities both because they have failed and as a preventive measure where a statistical analysis has indicated a high probability of a future failure. Nozzle failures and such statistically indicated failures have been occurring frequently enough to be a major concern for nuclear power plant operators (and owners) for a number of reasons including the high cost of repairs and the millions of dollars in lost revenue due to plant shut down. Therefore, there is a need for procedures for replacing and repairing nozzles that have failed or may fail or leak in the future, that (a) reduce the time and expense required to make the replacement or repair and (b) do not require confined entry into a pressure vessel, which reduce radiation exposure to the personnel performing the replacement or repair, and (c) reduce the susceptibility to PWSCC and do not result in further degradation of the vessel, and accordingly reduce the risk of future failures and leaks. A similar need also exists for a nozzle for initial installation applications and a method of initially installing such a nozzle in a vessel. The invention disclosed herein addresses the above-described needs and avoids the problems discussed above, and provides original nozzles and nozzle assemblies, replacement nozzles and nozzle assemblies, nozzle repair assemblies and procedures for installing and repairing nozzles and nozzle assemblies mechanically in pressure vessels in nuclear power facilities (and in other fields) that avoid some or all of the problems discussed above, and which do not employ a structural weld or a weld of any kind to the vessel. OBJECTS AND SUMMARY OF THE INVENTION It is an object of the invention disclosed herein to reduce the susceptibility to PWSCC of replacement nozzles and initial installation nozzles in nuclear power facilities as much as reasonably achievable, and thereby reduce the possibility of future nozzle failures. It is another object of the invention to reduce the time and expense involved with installing a replacement or initial installation nozzle, or repairing an existing nozzle in a vessel, particularly in a nuclear power facility. It is another object of the invention to not require confined entry into a vessel in order to install a replacement or initial installation nozzle in the vessel, or to repair an existing nozzle, particularly in a nuclear power facility. It is another object of the invention to reduce the exposure to radiation of repair personnel in a nuclear power facility while installing a replacement or initial nozzle installation, or repairing an existing nozzle in a vessel. It is another object of the invention to install a replacement or initial installation nozzle in a vessel, or repair an existing nozzle without structurally welding the nozzle to the vessel, particularly in a nuclear power facility. It is another object of the invention to install a replacement or initial installation nozzle in a vessel without any welding at all, particularly in a nuclear power facility. It is another object of the invention to repair an existing nozzle without any welding at all, or at least without welding the existing nozzle to the vessel, particularly in a nuclear power facility. It is another object of the invention when replacing a nozzle in a vessel to remove the entire existing nozzle and to install a complete (whole) mechanical nozzle replacement, thereby removing the defective (or possibly defective) portion of the existing nozzle and avoiding further degradation to the vessel. It is another object of the invention when partially replacing a nozzle in a vessel to not remove the part of the existing nozzle adjacent the interior of the vessel, i.e., to only remove the part of the existing nozzle from a point beyond the J-groove weld to the exterior of the vessel, and to install a partial mechanical nozzle replacement which is sealed to the bore adjacent the remaining nozzle portion, thereby simplifying removal of the existing nozzle. It is another object of the invention when repairing an existing nozzle to not remove the nozzle at all, but to either provide a mechanical seal for the existing nozzle, or attach the existing nozzle to the vessel without welding the existing nozzle to the vessel, or both. It is another object of the invention provide a mechanical attachment and/or a mechanical seal in the repair an existing nozzle without employing a weld of any kind, or optionally, a weld that is not made to the vessel, e.g., a weld of the existing nozzle to a clamp or flange device located outside of the vessel. It is another object of the invention to provide nozzles, nozzle assemblies and nozzle repair assemblies which may be installed in vessels while achieving the objects set forth above. It is another object of the invention to apply the techniques described herein and in the prior application for full nozzle replacement to partial nozzle replacement and nozzle repair, and to apply the techniques described herein for partial nozzle replacement to full nozzle replacement and nozzle repair. It is another object of the invention to provide a primary and a secondary seal for applicable replacement and repair assemblies to prevent leakage if the primary seal leaks. It is another object of the invention to provide a primary corrosion resistant leak path in replacement and repair assemblies so that in the event the primary and secondary seals leak, corrosion of the vessel would not occur. It is another object of the invention to provide a secondary corrosion resistant leak path in replacement and repair assemblies in the event the primary and secondary seals leak and the primary leak path seals failed, so that corrosion of the vessel would not occur. It is another object of the invention to provide a leak path or paths which provide a visual indication of any leakage in the leak path. It is another object of the invention to provide a means to prevent nozzle rotation in replacement and repair assemblies for those parts which may have high torsional loads. It is another object of the invention to live load the mechanical seals provided by the invention such that seal shrinkage over time will not cause leakage. The applicant recognized that the effect of high tensile stresses in nozzles introduced by welds that structurally attach and/or seal the nozzle to the vessel and the corrosive effect of coolant within the vessel can be limited by mechanical designs, which are disclosed herein and in the prior application. The invention(s) disclosed herein and in the prior application achieve the above and other objects by providing full and partial nozzle replacements and full and partial nozzle replacement assemblies for vessels, and repairs for existing nozzles, all of which are mechanically attached to the vessel and mechanically sealed to the vessel, i.e., without any welding at all to the vessel. The invention(s) also provide nozzle replacement and repair procedures for mechanically attaching and mechanically sealing existing nozzles, the full and partial nozzles and nozzle assemblies to the vessel, and for repairing nozzles, also without any welding to the vessel. The replacement and repair techniques, features and arrangements disclosed herein are also applicable to initial nozzle installation. Therefore, "replacement" should be understood to encompass initial installation where applicable. Most embodiments of the invention do not use any weld at all, although some may weld the nozzle to a sleeve or flange that is mechanically attached to the vessel. The full and partial replacement nozzles and replacement and repair assemblies are flanged, clamped, or bolted or compressively loaded to the vessel, and in some embodiments a part of the replacement nozzle is attached to the vessel with an interference fit. In all embodiments a mechanical seal is obtained using the interference fit between metal surfaces of a part of the nozzle and the vessel (which surfaces may be polished), and/or by use of sealing materials. In most embodiments, the seals are live-loaded to self-compensate for changes in the sealing material after initial installation. Also, in most embodiments, the parts of a nozzle assembly adjustably compressively load the sealing material, so that a part or parts may be tightened after initial installation to compensate for shrinkage of the sealing material. In the case of replacement of an existing nozzle, a part or the entire existing nozzle is removed and a partial or full nozzle or nozzle assembly is mechanically attached and mechanically sealed to the vessel. In the case of repair, the existing nozzle within the bore is not removed, and a mechanical seal is provided for the existing nozzle, and the existing nozzle may also be attached to the vessel without welding it to the vessel. For use in booth replacement and repair, means are also provided to attach the nozzle to the vessel without any welding at all via a clamping arrangement engaged at the nozzle and a drive sleeve. The invention departs significantly from the prior art of nozzle replacement and initial installation by not utilizing a weld of any kind, which eliminates the stresses imposed by welding and significantly reduces the risk of a PWSCC type failure, or by not utilizing any weld at all. The invention(s) disclosed herein further provide a corrosion resistant leak path or paths past the nozzle or nozzle assembly to the exterior of the vessel should the mechanical seals leak for any reason. The leak path or paths channel any reactor coolant leakage which may occur through a corrosion resistant path out past the exterior surface of the vessel and beyond the insulation surrounding the vessel to the oxygenated environment where the coolant can flash steam without eroding the vessel. Preferably, any leakage in a leak path or leak paths is visible. Further, the invention(s) disclosed herein provide an anti-rotation device or devices which prevent a mechanically attached nozzle part from rotating relative to the vessel. Still Further, the invention(s) disclosed herein provide a repair assembly which rotatably couples a sleeve threaded to the bore and a nozzle or nozzle body so that the sleeve may be tightened without breaking a weld between the nozzle and the assembly which attaches the nozzle to the sleeve. This technique allows the sleeve to be torqued to tighten a mechanical seal without having to break the weld. According to one embodiment of the invention, an existing nozzle in a bore of an ASME pressure vessel is partially replaced by a replacement nozzle or replacement nozzle assembly, which comprises a nozzle stub opening to the interior of the vessel and extending part way into the bore and a partial replacement nozzle or nozzle assembly extending therefrom or spaced therefrom to the exterior of the vessel. The nozzle stub is part of a previously installed nozzle which the nozzle or nozzle assembly is replacing, and is welded to the vessel where the bore opens to the interior of the vessel. The replacement nozzle or nozzle assembly also includes the partial replacement nozzle or assembly extending from adjacent the nozzle stub in the bore, through the bore and exiting the bore to the exterior of the vessel, which is mechanically attached to the vessel, and one or more mechanical seals of the partial replacement nozzle or assembly to the vessel. In some embodiments, a mechanical seal is independent of the existing nozzle stub, and in others cooperates with the nozzle stub to form the mechanical seal. According to another embodiment of the invention, an existing nozzle in a bore of an ASME pressure vessel is entirely removed and replaced by a full replacement nozzle or replacement nozzle assembly extending from the interior of the vessel, through the bore and exiting the bore to the exterior of the vessel. The full replacement nozzle or assembly is mechanically attached to vessel, and a mechanical seal is made of the full nozzle or assembly to the vessel when the full nozzle or assembly is mechanically attached to the vessel. In the repair embodiments, the existing nozzle remains welded to the vessel at the opening of the bore extending out of the vessel. A repair assembly is provided comprising a sleeve mechanically attached to the vessel, e.g., by threading the sleeve to the bore or by a flange bolted to the vessel, with the sleeve extending into the bore surrounding a substantial portion of the nozzle in the bore. A mechanical seal of the sleeve to the vessel within the bore is made surrounding the existing nozzle. If PWSCC were to occur after making a repair as described herein, a portion or all of the existing nozzle may be removed and a partial or full nozzle replacement made as described herein. The adaptation of the bore is the same as in the applicable partial or full nozzle replacement, except that part or all the existing nozzle is removed and a full or partial nozzle is installed. In the partial and full replacement nozzles and assemblies, and also in the repair assemblies, the mechanical seal or seals may be entirely within the bore or only at the vessel exterior surrounding the bore, or both. Various mechanical arrangements are disclosed for mechanically attaching partial and full replacement nozzles and assemblies and repair assemblies to the vessel. The nozzle may be one-piece or an assembly including a nozzle body and another part such as a sleeve or flange for mechanically attaching the nozzle body to the vessel. In one embodiment, the nozzle or nozzle assembly or repair assembly may include a sleeve which has exterior threads that engage interior threads in the bore to bolt the nozzle or assembly to the vessel, or the nozzle or assembly may be configured to engage structure in the bore when the nozzle or assembly is tightened in the bore to mechanically compressively load the nozzle or assembly to the vessel. The engaging structure may be interfering sections of the bore and a nozzle part, e.g., tapered sections of a nozzle part and the bore or a tapered sleeve inserted in the bore, or a flanged nozzle part and a shoulder in the bore, or a flange engaged with or attached to a nozzle body exteriorly of the vessel and bolted to the vessel. In another embodiment, a clamp arrangement frictionally engages the nozzle and a drive sleeve to mechanically attach the nozzle to the vessel without welding. The mechanical seal may be provided by sealing material, e.g., gasket material or packing material, glands or rings compressed between interfering surfaces of the nozzle or nozzle assembly and the bore or vessel exterior. For example, sealing material may be compressed between tapered surfaces, flanges and shoulders in the bore and an external flange and the vessel exterior, to name a few. Alternatively, a mechanical seal may be provided between contacting surfaces such as are compressively loaded together when the nozzle or nozzle assembly is attached to the vessel or via a shrink fit. Again, the seals may be live-loaded and also the compression thereof may be adjusted. In partial nozzle replacement embodiments where the nozzle stub remains in the bore and repair embodiments where the entire nozzle remains in the bore, the bore may be enlarged around the existing nozzle and a nozzle, nozzle body or sleeve overlapping the existing nozzle, with sealing material positioned around the existing nozzle compressed by the end of the nozzle, nozzle body or sleeve. This embodiment does not axially load the existing nozzle. This application discloses additional embodiments of full nozzle replacements, and also discloses embodiments of partial nozzle replacements and existing nozzle repairs. The features and principles disclosed herein and in the prior application may be utilized for full nozzle replacement, partial nozzle replacement and nozzle repair. Thus, it will be apparent to those of skill in the technical field to which this application relates that such features and principles disclosed in connection with one embodiment may be applied to other embodiments.
050287966	abstract	An improved support belt for radiation shield garments has a padded, slip-resistant rear support element, one or more elasticized elements, one or more connecting elements, and adjustable fastening means for securing the ends of the belt to one another. The padded, slip-resistant rear support element comprises a relatively thin, flexible foam core that is covered with a foam-backed vinyl, the foam facing outwardly for frictionally engaging and supporting the radiation shield garment at the waist of a user.
	claims	1. A nuclearized hot-isostatic press (HIP) system comprising:a high temperature HIP furnace;a multi-wall vessel surrounding the furnace, wherein the multi-walled vessel comprises an inner wall and an outer wall, at least one detector contained between the inner wall and the outer wall to detect a gas leak, a crack in a vessel wall, or both;multiple heads located on top and underneath the furnace;a yoke and a yoke frame; anda lift mechanism configured to load and unload a HIP can to the high temperature HIP furnace. 2. The nuclearized HIP system of claim 1, wherein the at least one detector comprises a pressure detector, a gas flow detector, a gas analyzer, a radiation detector, or an acoustic detector. 3. The nuclearized HIP system of claim 1, wherein multi-wall vessel comprises two concentric vessels. 4. The nuclearized HIP system of claim 3, wherein two concentric vessels contain at least one groove between the vessels, wherein said groove is contained in the outside of the inner vessel or on the inside of the outer vessel, or both, and forms one or more pathways for gas located between the vessel walls to travel. 5. The nuclearized HIP system of claim 1, wherein the yoke comprises multiple elements and is configured to allow the yoke frame to remain operational upon the failure of one element of the yoke. 6. The nuclearized HIP system of claim 1, further comprising at least one strain gauge on the yoke configured to collect and provide real time stress data during operation of the nuclearized HIP system. 7. The nuclearized HIP system of claim 1, wherein the multiple heads comprise a top head, an outer lower head, and an inner lower head. 8. The nuclearized HIP system of claim 7, wherein the outer lower head is configured to allow the furnace to sit on it. 9. The nuclearized HIP system of claim 8, wherein the outer lower head can be locked to the vessel while the inner lower head can be lowered to accept a part to be pressed in the nuclearized HIP system. 10. The nuclearized HIP system of claim 7, wherein the inner lower head is configured to hold a stand on which a part to be pressed in the nuclearized HIP system is placed, and is configured to allow the inner lower head to fit within the inner diameter of the outer lower head. 11. The nuclearized HIP system of claim 7, wherein the inner lower head contains at least one seal to form a seal with the outer head, and/or to keep the furnace and thermal barrier in place when the component to be pressed is loaded and unloaded. 12. The nuclearized HIP system of claim 7, wherein the inner lower head comprises at least one pneumatic pin, cylinder or clamp that couples it to the outer lower head. 13. The nuclearized HIP system of claim 7, wherein the top head is located on top of the furnace and sits in the bore of the vessel. 14. The nuclearized HIP system of claim 1, wherein the system comprises a loading element comprising an electric lift, hydraulic cylinders, pneumatic cylinders, machine screws, or a combination thereof, to load and unload a HIP can from outside the HIP system to the HIP furnace. 15. The nuclearized HIP system of claim 14, wherein the loading element comprises a bottom loading design. 16. The nuclearized HIP system of claim 15, wherein the system further comprises a dual bottom closure design to allow the furnace and thermal barrier to stay in place inside the vessel while a part that was pressed in the nuclearized HIP system is removed from the system. 17. The nuclearized HIP system of claim 1, wherein the furnace is locked in place for normal operation with spring loaded catches. 18. The nuclearized HIP system of claim 17, wherein the latches can either be manually or automatically actuated. 19. The nuclearized HIP system of claim 1, further comprising at least one thermal barrier layer located between the furnace and the multi-walled vessel.
	abstract	A beam manipulating arrangement for a multi beam application using charged particles comprises a multi-aperture plate having plural apertures traversed by beams of charged particles. A frame portion of the multi-aperture plate is heated to reduce temperature gradients within the multi-aperture plate. Further, a heat emissivity of a surface of the multi-aperture plate may be higher in some regions as compared to other regions in view of also reducing temperature gradients.
054992834	abstract	A method and a apparatus for helical scan imaging in an X-ray CT, in which the initial set up operation can be achieved easily and accurately. In this X-ray CT apparatus, relative motion of a bed plate and an X-ray source is automatically controlled according to a desired imaging region specified by the operator such that the bed plate and the X-ray source are relatively moved through a distance covered by a scanning region appropriate for collecting the data required to reconstruct tomographic images for the desired imaging region. The scanning region includes the desired imaging region and supplementary interpolation data regions, located at the ends of the desired imaging region and utilized in carrying out interpolations for deriving data in the desired imaging region, and initial acceleration region accounting an initial acceleration of the bed plate, and a final deceleration region accounting a final deceleration of the bed plate.
051075300	abstract	A reliable shutter position indicator for an x-ray diffractometer utilizes an optical shutter position sensor for detecting the position of a shutter of the x-ray diffractometer. In one embodiment, a pair of optical isolators are utilized with an arm coupled to the shutter shifting to a position to interrupt a light beam of the optical isolators when the shutter is in either the open or closed position. The controller of the x-ray diffractometer then compares the detected shutter position with the expected shutter position to determine whether a shutter error is present.
052232085	abstract	Groundwater running below the surface is intercepted with a cutoff wall of a predetermined height and length to form an underground dam, and a nuclear power generation system is built below said underground dam. The water pooled in the underground dam is utilized as the secondary cooling water for the reactor of the power system.
052934110	claims	1. A boiling water type nuclear reactor power control method, comprising the steps of: feeding water with a plurality of first pumps to the nuclear reactor; generating a trip signal when one of the first pumps supplying feed water to the nuclear reactor is tripped; and controlling the nuclear reactor power to be at a first value of at least 80 percent of rated output power prior to any of said first pumps being tripped and reducing the power in the nuclear reactor in response to the trip signal by first decreasing the speed of a recirculation pump providing core flow to a core of the nuclear reactor to provide a core flow corresponding to a nuclear reactor power of a second value less than said first value. a plurality of first pump means for feeding feed water to the nuclear reactor; trip detection means for outputting a trip signal when one of the first pump means for supplying feed water to the nuclear reactor is tripped, said nuclear reactor operating at a predetermined first level of nuclear reactor power when said one of the first pump means is tripped; first control means responsive to said trip signal; reduction means responsive to said first control means for reducing circulation of core flow in the core of the nuclear reactor sufficiently to reduce power; said first control means controlling said reduction means to first reduce the circulation of core flow in response to said trip signal; second reduction means including a variation means for determining and generating a mismatched amount corresponding to a difference between the flow of steam discharged from said nuclear reactor and the flow of the feed water after the trip signal is generated, and second control means for controlling and reducing the nuclear reactor power to a predetermined second level from said first level based upon the mismatched amount. means for determining and generating a mismatched amount corresponding to a difference between the flow of steam discharged from said nuclear reactor and the flow of the feed water after the trip signal is generated, and means for controlling the nuclear reactor power based upon the mismatched amount. a plurality of first pumps for supplying feed water to the nuclear reactor; a second pump circulating cooling water in the core in said nuclear reactor; trip detection means for generating a trip signal when one of said first pumps is tripped; first output control means to output a first control signal for operating the nuclear reactor at a first level of power when none of said first pumps is tripped; second output control means to output a second control signal for operating the nuclear reactor at a second level of power, less than said first level, when one of said first pumps is tripped; selection control means for switching control from said first output control means to said second output control means; and reactor power reduction means for first reducing the speed of said second pump in response to generating said trip signal for reducing the nuclear power and then switching said selection control means to further reduce the nuclear reactor power to said second level from said first level. a plurality of first pumps supplying feed water to the nuclear reactor; a second pump circulating cooling water in the core in said nuclear reactor; trip detection means for generating a trip signal when one of said first pumps is tripped; core flow control means for controlling the speed of said second pump; control rods mounted for insertion into the core; control rod control means for controlling operation of said control rods; first output control means to output a first control signal for operating the nuclear reactor at a first level of power when none of said first pumps is tripped; second output control means to output a second control signal for operating the nuclear reactor at a second level of power; said core flow control means controlling said second pump based upon the output signal from said second output control means and said first output control means; said control rod control means controlling selected control rods based on the output signal of said second output control means and said first output control means; and reactor power reduction means for first reducing the speed of said second pump in response to generating said trip signal and then switching operation to said second control means to further reduce the nuclear reactor power to said second level from said first level. controlling nuclear reactor power in response to a first nuclear reactor power level set at a first predetermined value corresponding to a required output of a generator driven by the nuclear reactor under normal operation of a feed water supplying system; detecting an abnormality in the feed water supplying system; reducing the nuclear reactor power by first decreasing core flow in response to said detecting of the abnormality; and then switching the nuclear reactor power level to a predetermined second level lower than said first level and corresponding to the specific abnormality detected, and continuing said controlling at the lower level. operating said nuclear reactor at a predetermined first level of power; detecting an abnormality of the feed water supplying system for the nuclear reactor; first reducing speed of a pump circulating cooling liquid in the core in the nuclear reactor when the abnormality is detected; thereafter, detecting a mismatched amount between flow of steam discharged from the nuclear reactor and flow of the feed water; and then reducing the nuclear reactor power to a predetermined second level of power less than said first level after said reducing speed step based upon the mismatched amount. operating said nuclear reactor at a predetermined first level of power detecting an abnormality by only measuring characteristics of the feed water supplying system, which cause a flow of feed water to decrease abnormally; reducing the nuclear reactor power in response to said detected abnormality by first reducing a speed of a recirculation pump providing core flow to a core in the nuclear reactor to a speed corresponding to a nuclear reactor power lower than said first nuclear reactor power; and thereafter, further reducing said nuclear reactor power to a predetermined second level lower than said first level of nuclear reactor power. operating said nuclear reactor at a predetermined first level of power; detecting an abnormality of the feed water supplying system; first reducing recirculation cooling flow within a core of the nuclear reactor, when the abnormality is detected, to a core flow corresponding to a nuclear reactor power lower than said first level by decreasing a speed of a pump providing recirculating cooling water in the core of the nuclear reactor; and thereafter further reducing nuclear reactor power to a predetermined second level less than said first level that prevents over speed of the others of the pumps that remain operating for supplying feed water. 2. A nuclear reactor power control method according to claim 1, wherein said reducing reduces the nuclear reactor power in accordance with a mismatched amount between the flow of steam discharged from the nuclear reactor and the flow of the feed water. 3. The nuclear reactor power control method of claim 1, wherein said reducing further includes adjusting selected control rods based upon a mismatched amount between a flow of the steam discharged from the nuclear reactor and the flow of the feed water. 4. The nuclear reactor power control method of claim 1, wherein said reducing reduces the nuclear reactor power to a level that prevents overspeed of the other first pumps under operation to supply feed water. 5. A nuclear reactor power control device for a boiling water type nuclear reactor, comprising: 6. The nuclear reactor power control device of claim 5, including: 7. The nuclear reactor power control device of claim 5, wherein said reduction means includes means to output a control signal to reduce the nuclear reactor power to a level that prevents overspeed of other ones of said first pump means still in operation and means for controlling the nuclear reactor power based upon the control signal. 8. The nuclear reactor power control device of claim 5, wherein said trip detection means sets the trip signal to a value to reduce the nuclear reactor power to a fixed value having a lower level than the nuclear reactor power at the moment of the output of the trip signal. 9. A nuclear reactor power control device according to claim 5, wherein said means for controlling the nuclear reactor power is means for controlling insertion of selected control rods into the core. 10. A nuclear reactor power control device for a boiling water type nuclear reactor, comprising: 11. A nuclear reactor power control device according to claim 10, wherein a part of said first output control means is commonly used by said second output control means. 12. A nuclear reactor power control device according to claim 10, including core flow control flow for controlling the speed of said second pump in response to the second control signal, control rods, control rod control means for controlling operation of said control rods. third output control means to output a third control signal for controlling selected control rods with said control rod control means. 13. A nuclear reactor power control device for a boiling water type nuclear reactor, comprising: 14. A method for controlling reactor power of a boiling water type nuclear reactor, comprising: 15. A method for controlling reactor power of a boiling water type nuclear reactor, comprising: 16. A method for controlling reactor power of a boiling water type nuclear reactor in response to an abnormality of a feed water supplying system, comprising: 17. A method for controlling reactor power of a boiling water type nuclear reactor in response to an abnormality of a feed water supplying system, said feed water being supplied by a plurality of pumps, comprising:
059094753	summary	FIELD OF THE INVENTION The invention relates to spent nuclear fuel storage and transport containment systems and more specifically to basket structures and thermal heat sinking systems that are compatible with low cost, concrete, spent nuclear fuel containment. BACKGROUND OF THE INVENTION As the nuclear power industry matures, the issue of safe storage, transport and disposition of spent nuclear fuel becomes increasingly critical. While the original intent in the industry was to reprocess spent fuel, that option has been delayed by a variety of cost and political concerns. Central storage of spent nuclear fuel in containers pending resolution of those issues, and in anticipation of future reprocessing, has been an industry goal, and storage containers for spent fuel have been designed in many cases in anticipation of hauling quantities of spent fuel over long distances to central storage facilities. The hauling of radioactively and thermally hot nuclear materials over public thoroughfares is highly regulated, and the containers for such materials must meet rigid safety and engineering standards. For this reason, the design and construction of containers for these materials is complex and expensive. However central storage of spent nuclear fuel has also been met with difficult regulatory and political issues, and to date has not been implemented on a large scale. Instead, nuclear power plants have resorted to on-site storage of spent fuel. Several plants have constructed expensive storage facilities for spent fuel utilizing water pools in which containers of spent rods are immersed. The surrounding water provides thermal cooling of the thermally "hot" materials, and affords radiation protection and isolation from the environment. However, pool capacity is limited and in some cases exhausted, and nuclear power plants are pursuing new and more economic strategies for handling spent fuel, including on-site dry storage. Previously, thin walled containers were developed and utilized for ease in handling. These containers could be lowered directly into the fuel storage pool, loaded with spent fuel rods, and once sealed, removed from the fuel pool and transferred to an outdoor storage area. This technology was attractive since it could be handled easily by utilities and due to its structural advantages, transported directly from the utilities to a central storage facility or to an underground repository. While thin walled metal transport and storage casks proved technically feasible, they also proved expensive. New technology, primarily storage only concrete containers were developed at a cost 6-8 times less expensive when compared to the metal containers. These concrete containers were approved for storage only and involved complex handling equipment and procedures since the large concrete containers could not be lowered into the fuel storage pool. In order to effect transfer, a thin walled metal transfer cask was lowered into the fuel pool, loaded with fuel and sealed. Its contents were transferred outside the fuel pool. Engineering issues for on-site dry storage containers are primarily thermal cooling and radiation shielding. Radiation shielding can effectively be incorporated in storage containers using thick steel walls or thick walls of other high radiation cross section materials, but this approach to radiation shielding results in a very heavy and very expensive container. The preferred solution from a cost standpoint is to use a concrete container. Concrete typically contains a significant amount of hydrated water which acts as an effective neutron radiation shield, and also contains iron aggregates which are effective in shielding gamma radiation. However, with anticipated storage times of months and years, or even decades, thermal issues are paramount. Storage of radioactive materials involves the use of sealed containers, and it is difficult to effectively cool materials that are generating heat inside of a sealed container. It is especially difficult if the container is concrete, or contains a concrete layer, as concrete is a poor heat transfer medium. Thus a satisfactory concrete storage container for thermally hot nuclear materials has not been available in the industry. BRIEF SUMMARY OF THE INVENTION With these and other industry realities and engineering constraints as a necessary framework, we have designed an improved container or "cask" for containing spent nuclear materials. The new design incorporates a novel thermodynamic system for efficient heat transfer from the sealed interior of the cask to the exterior. It also provides a high efficiency mechanical system for high container strength. In addition it provides double containment for sealing the cask against the atmosphere, and effective radiation shielding. Moreover, it utilizes concrete for radiation shielding, thus reducing the cost of storage containers and meeting a primary need in the industry. Added together these advantages establish a new level of safety and cost efficiency for spent fuel rod containers. Contributing to the achievement of these goals according to the invention is a multi-unit storage array, frequently termed a basket, in which several storage elements are formed from stacked units of an interlocking grid array. The interlocking array is mechanically coupled to an integral polygonal, e.g. octagonal, intermediate metal shell. The intermediate metal shell is embedded inside a rigid cylinder of filler material, e.g. concrete, and this assembly is sealed in an outer casing of high strength metal, e.g. steel. A key to the design is that the interior integral array of tubular supports comprises radial members that extend through the intermediate shell, through the rigid cylinder of filler material, and efficiently couple both mechanically and thermally to the outer metal casing. These radial members both impart strength to the mechanical system, and act as heat sinks for effectively conducting potentially dangerous levels of heat generated by the nuclear waste material to the exterior of the cask. As will be evident, the cask in this design has a first containment barrier at the intermediate shell and a second containment barrier at the outer steel casing. Thus the structure affords dual containment of the atmosphere sealed within the container.
056028873	claims	1. A tool for pushing a mass suspended inside a reactor pressure vessel, comprising: a pole adaptor having means for coupling to an end of a service pole, said pole coupling means being at one end of said pole adaptor; an extension member having one end connected to another end of said pole adaptor; a mounting member having hole means, said hole means having an axis; a support post fixed in and coaxial with said hole means; a pivot pin for coupling an end of said support post to an end of said extension member such that said support post and said extension member are relatively swivable about said pivot pin; a pivoting arm pivotably mounted on said mounting member for pivoting about a first axis which is generally parallel to said support post; a fluid-actuated drive mechanism supported by said mounting member and coupled to said pivoting arm so that said pivoting arm pivots about said first axis in a direction away from said mounting member in response to the input of pressurized fluid into said fluid-actuated drive mechanism; and a rocker plate pivotably mounted on a distal end of said pivoting arm, said rocker plate being pivotable about a second axis which is generally parallel to said first axis. a mounting member; a fluid-actuated spreader mounted on said mounting member and having open and closed states, said hydraulic spreader comprising a stationary member fixed to said mounting member, a fluid-actuated cylinder supported by said stationary member and a pivoting member pivotably mounted to said stationary member and coupled to said fluid-actuated cylinder, said pivoting member being pivotable about a first axis between a first angular position in said closed state and a second angular position in said open state in response to actuation of said fluid-actuated cylinder; an adaptor bracket attached to said pivoting member by first and second fasteners and extending further away from said first axis than does said pivoting member, said adaptor bracket comprising coaxial first and second pivot pin supports; a pivot pin supported by said first and second pivot pin supports and having a second axis which is generally parallel to said first axis; a rocker plate pivotably mounted only on said pivot pin; and a pressurized fluid line for connecting said fluid-actuated spreader to a supply of pressurized fluid, wherein said fluid-actuated spreader transitions from said closed state to said open state in response to the supply of pressurized fluid to said pressurized fluid line. a support structure; a fluid-actuated spreader mounted on said support structure and having open and closed states, said fluid-actuated spreader comprising a stationary member fixed to said support structure, a fluid-actuated cylinder supported by said stationary member and a pivoting member which is pivotable about a first axis between a first angular position in said closed state and a second angular position in said open state; an adaptor bracket attached to said pivoting member; a rocker plate pivotably mounted on a distal end of said adaptor bracket, said rocker plate being pivotable about a second axis which is generally parallel to said first axis; a roller pivotably mounted on said rocker plate, said roller being rotatable about a third axis which is generally perpendicular to said second axis; and a pressurized fluid line for connecting said fluid-actuated spreader to a supply of pressurized fluid, wherein said fluid-actuated spreader transitions from said closed state to said open state in response to the supply of pressurized fluid to said pressurized fluid line. 2. The tool as defined in claim 1, further comprises a saddle attached to said rocker plate. 3. The tool as defined in claim 2, wherein said saddle has a surface made of plastic material. 4. The tool as defined in claim 1, further comprises a roller pivotably mounted on said rocker plate, said roller being rotatable about a third axis which is generally perpendicular to said second axis. 5. The tool as defined in claim 4, wherein said roller has a surface of made plastic material. 6. The tool as defined in claim 1, wherein said fluid-actuated drive mechanism comprises a fluid-actuated cylinder, and said pivoting arm comprises a pivoting member which is pivotable about said first axis when said fluid-actuated cylinder is actuated and an adaptor bracket attached to said pivoting member. 7. The tool as defined in claim 4, wherein said roller has a central circumferential depression. 8. The tool as defined in claim 6, further comprising a restrictor plate attached to said adaptor bracket, wherein said restrictor plate is positioned to stop pivoting of said rocker plate about said second axis beyond a predetermined angle. 9. The tool as defined in claim 2, wherein said rocker plate comprises a base pivotably mounted on a pivot pin and an extension on one side which extends generally perpendicular to said base and beyond the portion of said saddle furthermost from said base. 10. A tool for pushing a mass suspended inside a reactor pressure vessel, comprising: 11. The tool as defined in claim 10, further comprises a saddle attached to said rocker plate. 12. The tool as defined in claim 11, wherein said saddle has a surface made of plastic material. 13. A tool for pushing a mass suspended inside a reactor pressure vessel, comprising: 14. The tool as defined in claim 13, wherein said roller has a surface made of plastic material. 15. The tool as defined in claim 13, wherein said roller has a central circumferential depression. 16. The tool as defined in claim 10, further comprising a restrictor plate attached to said adaptor bracket, wherein said restrictor plate is positioned to stop pivoting of said rocker plate about said second axis beyond a predetermined angle. 17. The tool as defined in claim 10, further comprising a support post connected to an extension by a clevis pin and connected to said mounting member. 18. The tool as defined in claim 17, further comprising a pole adaptor having means for coupling to an end of a service pole, said, said pole adaptor being connected to said extension. 19. The tool as defined in claim 11, wherein said rocker plate comprises a base pivotably mounted on a pivot pin and an extension on one side which extends generally perpendicular to said base and beyond the portion of said saddle furthermost from said base. 20. The tool as defined in claim 10, wherein said mounting member comprises contact surface means which are contoured to conform to the shape of the internal surface of a circular cylindrical structure.
046631151	claims	1. An apparatus for protecing organisms and the environment from harmful emissions from a source of heterogenous radioactive waste by shielding said organisms and the environment from at least a portion of such emissions, while drawing off the power of such emissions as electricity, which comprises a plurality, from 2 to 100, of pairs of concentric metal spherical shells surrounding a spherical mass of heterogeneous radioactive waste, the first of each pair of shells, located nearer to the heterogeneous radioactive waste, being of a lower atomic number than the other and being selected from the group consisting of aluminum, magnesium and titanium, and the other of each pair of shells being of a higher atomic number and being selected from the group consisting of copper, silver, iron, cobalt, nickel, manganese, chromium, vanadium and zinc, said pairs of spherical metal shells being separated by a dielectric selected from the group consisting of alumina, mica, silica, glass, synthetic organic polymeric plastic, and air, and said firsts of each pair of metal shells being electrically connected, and said others of each pair of metal shells being separately electrically connected, each to a separate conductor, and said conductors being connected to an electrical load located remote from the heterogeneous radioactive waste, so that electricity generated by the radioactive emissions from the radioactive waste is consumed remote from said waste, without heating the waste or the shielding and without raising the voltages thereof to dangerous levels. 2. An apparatus according to claim 1 wherein a source of electrical potential difference is present and is connected to the pairs of metal shells, with a positive connection being to the metal shells that are more positive, and a negative connection being to the metal shells that are more negative, which source of electrical potential difference, connected as described, stimulates flow of electricity generated by the radioactive waste, at least one of the connections being disconnectable, so that when such radioactivity induced flow of electricity begins the flow of stimulating current from the source of electrical potential difference may be halted. 3. An apparatus according to claim 1 wherein the difference in atomic number between the metal of the first of each pair of metal shells and the metal of the other of such pairs is at least 10. 4. An apparatus according to claim 3 wherein the metal of the first of each pair of shells is aluminum, the metal of the other of the pairs of shells is copper, and the dielectric is air. 5. An apparatus according to claim 1 wherein the metal of the first of each pair of shells is aluminum, the metal of the other of each pair of shells is copper, and the dielectric is air. 6. An apparatus according to claim 1 wherein the electrical load is a variable resistance, which automatically maintains the flow of current and keeps the voltage low. 7. An apparatus according to claim 2 wherein the electrical load is a variable resistance, which automatically maintains the flow of current and keeps the voltage low. 8. An apparatus according to claim 7 wherein the metal of the first of each pair of shells is aluminum, the metal of the other of each pair of shells is copper, and dielectric between each of the pairs of shells is air. 9. An apparatus according to claim 1 wherein the pairs of metal shells are in flexible sheet form, with alternating layers of different metals and layers of dielectric between them.
	claims	1. A focused ion beam apparatus, comprising:a sample stage for mounting a specimen;an ion beam irradiating optical system for irradiating a focused ion beam to the specimen;a deposition-gas supplying source for supplying a deposition gas used for forming a deposition film at an area to be irradiated with the focused ion beam;a sample holder provided with a fixing surface for fixing, via the deposition film, a micro-specimen separated from the specimen using a method for fabrication by the focused ion beam, in which a width of the fixing surface is smaller than 50 microns;a specimen transferring unit having a probe to which the specimen can be joined through the deposition film, and transferring the micro-specimen separated from the specimen by the focused ion-beam fabrication method, to said sample holder; anda sample chamber in which said sample, said sample holder and at least a portion of said probe are laid out,wherein, in said sample chamber, the micro-specimen separated from the specimen by the focused ion-beam fabrication method is fixed to the fixing surface of said sample holder through the deposition film, and the micro-specimen fixed to the fixing surface is fabricated by irradiating the focused ion beam. 2. A focused ion beam apparatus, comprising:a sample stage for mounting a specimen;an ion beam irradiating optical system for irradiating a focused ion beam to the specimen;a deposition-gas supplying source for supplying a deposition gas used for forming a deposition film at an area to be irradiated with the focused ion beam;a sample holder provided with a fixing surface for fixing, via the deposition film, a micro-specimen separated from the specimen using a method for fabrication by the focused ion beam, in which a dent with depth of at least 20 microns is not provided on the fixing surface of said sample holder;a specimen transferring unit having a probe to which the specimen can be joined through the deposition film, and transferring the micro-specimen separated from the specimen by the focused ion-beam fabrication method, to said sample holder; anda sample chamber in which said sample, said sample holder and at least a portion of said probe are laid out,wherein, in said sample chamber, the micro-specimen separated from the specimen by the focused ion-beam fabrication method is fixed to the fixing surface of said sample holder through the deposition film, and the micro-specimen fixed to the fixing surface is fabricated by irradiating the focused ion beam. 3. A focused ion beam apparatus, comprising:a sample stage for mounting a specimen;an ion beam irradiating optical system for irradiating a focused ion beam to the specimen;a deposition-gas supplying source for supplying a deposition gas used for forming a deposition film at an area to be irradiated with the focused ion beam;a sample holder provided with a fixing surface for fixing, via the deposition film, a micro-specimen separated from the specimen using a method for fabrication by the focused ion beam, in which a width of the fixing surface is smaller than 50 microns, and the fixing surface is flat so that the separated micro-specimen fixed to the fixing surface is not embedded in a dent on the fixing surface;a specimen transferring unit having a probe to which the specimen can be joined through the deposition film, and transferring the micro-specimen separated from the specimen by the focused ion-beam fabrication method, to said sample holder; anda sample chamber in which said sample, said sample holder and at least a portion of said probe are laid out,wherein, in said sample chamber, the micro-specimen separated from the specimen by the focused ion-beam fabrication method is fixed to the fixing surface of said sample holder through the deposition film, and the micro-specimen fixed to the fixing surface is fabricated by irradiating the focused ion beam. 4. The focused ion beam apparatus according to claim 1, wherein the sample holder is provided with a metallic wire. 5. The focused ion beam apparatus according to claim 1, wherein the fixing surface has a metallic surface. 6. The focused ion beam apparatus according to claim 1, wherein the sample holder is provided with a sliver of silicon. 7. The focused ion beam apparatus according to claim 1, wherein the fixing surface has a ground surface of silicon. 8. The focused ion beam apparatus according to claim 1, wherein the cross section of the sample holder is a convex shape. 9. The focused ion beam apparatus according to claim 2, wherein the sample holder is provided with a metallic wire. 10. The focused ion beam apparatus according to claim 2, wherein the fixing surface has a metallic surface. 11. The focused ion beam apparatus according to claim 2, wherein the sample holder is provided with a sliver of silicon. 12. The focused ion beam apparatus according to claim 2, wherein the fixing surface has a ground surface of silicon. 13. The focused ion beam apparatus according to claim 2, wherein the cross section of the sample holder is a convex shape. 14. The focused ion beam apparatus according to claim 3, wherein the sample holder is provided with a metallic wire. 15. The focused ion beam apparatus according to claim 3, wherein the fixing surface has a metallic surface. 16. The focused ion beam apparatus according to claim 3, wherein the sample holder is provided with a sliver of silicon. 17. The focused ion beam apparatus according to claim 3, wherein the fixing surface has a ground surface of silicon. 18. The focused ion beam apparatus according to claim 3, wherein the cross section of the sample holder is a convex shape.
	summary
061817619	claims	1. An apparatus for monitoring reactor power of a reactor at least at the time of startup thereof, comprising: a radiation sensor for detecting radiation doses associated with the reactor power as electrical signals having pulse components according to the radiation doses; analog filter means for filtering electrical signals having frequency components in a certain frequency band from the electrical signals detected by the radiation sensor; digital conversion means for converting the electrical signals having the frequency components filtered by the analog filter means to first digital data with an accuracy of certain bits on sampling intervals shorter than a pulse width of the pulse components; pulse measurement means for counting the number of pulses of the pulse components in a predetermined frequency band based on the first digital data converted by the digital conversion means; sum operating means for adding a plurality of sampling values forming the first digital data based on preset sum operation conditions so as to obtain second digital data having accuracy having more bits than those of the plurality of sampling values forming the first digital data converted by the digital conversion means; Campbell measurement means for calculating mean square values corresponding to power of fluctuation components caused by overlap of the pulse components in a predetermined frequency band based on the second digital data obtained by the sum operating means; and reactor power monitoring means for continuously monitoring the reactor power based on the number of pulses counted by the pulse measurement means and the mean square values calculated by the Campbell measurement means. wherein the digital conversion means is means for converting the electrical signals having the frequency components from the analog filter means to the first digital data at sampling intervals corresponding to 1/n of the pulse width of the pulse components, where n is a positive number; and wherein the pulse measurement means includes means for obtaining an operation value Out(k) at the sampling intervals, selecting the pulse components based on the Out(k), and counting the number of pulses of the pulse components selected, the Out(k) being operated based on the following formula; EQU Out(k)=C(0).times.S(k)+C(1).times.S(k-1)+C(2).times.S(k-2)+ . . . +C(k-n+1).times.S(k-n+1), where S(k) is a k-th sampling value among a plurality of sampling values forming the first digital data from the digital conversion means, S(k-1), S(k-2), . . . , S(k-n+1) are sequential sampling values before the S(k), the number of the sampling values is determined by said positive number n, and C(0), C(1), C(2), . . . , C(k-n+1) are constants individually allotted to the sampling values S(k), S(k-1), S(k-2), . . . , S(k-n+1). wherein the digital conversion means is means for converting the electrical signals having the frequency components from the analog filter means to the first digital data at sampling intervals, the sampling intervals being not less than one fourth and not more than one third of the pulse width of the pulse components, and wherein the pulse measuring means includes means for obtaining an operation value Out(k) at the sampling intervals, selecting the pulse components based on the Out(k), and counting the number of pulses of the pulse components selected, the Out(k) being operated based on the following formula; EQU Out(k)=-a.times.S(k-3)+b.times.S(k-2)+c.times.S(k-1)-d.times.S(k), where S(k) is a k-th sampling value among a plurality of sampling values forming the first digital data from the digital conversion means, S(k-1), S(k-2) and S(k-3) are sequential three sampling values before the S(k), and a, b, c and d are constants individually allotted to the four sampling values S(k), S(k-1), S(k-2) and S(k-3). wherein the digital conversion means is means for converting the electrical signals having the frequency components from the analog filter means to the first digital data at sampling intervals, the sampling intervals being not less than one third of and not more than a half of the pulse width of the pulse components and wherein the pulse measuring means includes means for obtaining an operation of value Out(k) at the sampling intervals, selecting the pulse components based on the Out(k), and counting the number of pulses of the pulse components selected, the Out(k) being operated based on the following formula; EQU Out(k)=-a.times.S(k-2)+2.times.b.times.S(k-1)-c.times.S(k), where S(k) is a k-th sampling value among a plurality of sampling values forming the first digital data from the digital conversion means, S(k-1) and S(k-2) are sequentially two sampling values before the S(k), and a, b and c:are constants individually allotted to the three sampling values S(k), S(k-1) and S(k-2). wherein the pulse measurement means comprises: analog type pulse comparison means used if a wave height value is not less than or not more than a predetermined value; and digital type pulse determination means for counting the number of pulses of the pulse components detected by the radiation sensor based on the first digital data converted by the digital conversion means; and means for counting only the number of pulses removed noise components from the pulse components detected by the radiation sensor based on both the analog type pulse comparison means and the digital type pulse determination means. wherein the digital conversion means is means for converting the electrical signals having the frequency components from the analog filter means to the first digital data at sampling intervals, the sampling intervals being 1/n of the pulse width of the pulse components, where n is a positive number; and wherein the pulse measuring means includes means for obtaining a first and second operation values Out 1(k) and Out 2(k) at the sampling intervals, selecting the pulse components based on the Out 1(k) and Out 2(k), and counting the number of pulses of the pulse components selected, the Out 1(k) and Out 2(k) being operated based on the following formula; EQU Out 1(k)=C(0).times.S(k)+C(1).times.S(k-1)+C(2).times.S(k-2)+ . . . +C(k-n+1).times.S(k-n+1) EQU Out 2(k)=D(0).times.S(k)+D(1).times.S(k-1)+D(2).times.S(k-2)+ . . . +D(k-n+1).times.S(k-n+1), where S(k) is a k-th sampling value among a plurality of sampling values forming the first digital data from the digital conversion means, S(k-1), S(k-2), . . . , S(k-n+1) are sequential sampling values before the sampling value S(k), the number of the sampling values is determined by said positive number n, C(0), C(1), C(2), . . . , C(k-n+1) are first constants individually allotted to the sampling values S(k), S(k-1), S(k-2), . . . ,S(k-n+1), and D(0), D(1), D(2), D(3), . . . , D(k-n+1) are second constants individually allotted to the sampling values S(k), S(k-1), S(k-2), . . . , S(k-n+1). wherein the Campbell measurement means comprises: cycle data removal means for removing data of a predetermined time interval from the second digital data acquired by the sum operating means; and Campbell evaluating means for operating the mean square value from residue data not removed by the cycle data removal means in a frequency band higher than a frequency band of the data removal cycle and correcting and outputting the mean square value operated in accordance with a ratio of data removed from the second digital data. wherein the cycle data removal means comprises: noise data addition means for discriminating data of not less than or not more than a predetermined value by the digital conversion means, determining the data as noise data and adding noise information on the noise data; and Campbell evaluating means for removing the data if the noise information is included in the second digital data and correcting and outputting a mean square value of the noise data in accordance with the ratio of the removed data to the second digital data. wherein the cycle data removal means comprises: noise monitoring means for evaluating the presence or absence of noise induced by an earth or the like and a cycle of the noise; and Campbell evaluating means for removing the noise data from the second digital data if the noise information is included in the second digital data in accordance with the noise cycle evaluated by the noise monitoring means and noise recognition timing, and correcting and outputting a mean square value of the removed data in accordance with the ratio of the removed data to the second digital data. 2. The apparatus of claim 1, wherein the sum operating means includes means for outputting the second digital data to the Campbell measurement means with output accuracy of more bits than those of the first digital data. 3. The apparatus of claim 1, wherein the sum operating means includes means for outputting the second digital data to the Campbell measurement means at longer output time intervals than sampling intervals for the digital conversion means. 4. The apparatus of claim 1, wherein the sum operating means is means for adding a plurality of sampling values forming the first digital data under a plurality of sum operating conditions having different numbers of added values as the sum operation conditions. 5. The apparatus of claim 1, wherein the sum operating means is means for adding a plurality of sampling values forming the first digital data under a plurality of sum operating conditions within a certain time as the sum operation conditions. 6. The apparatus of claim 1, wherein the Campbell measurement means includes Campbell output evaluating means for evaluating the mean square values in a plurality of frequency bands based on the second digital data from the sum operating means, and Campbell signal discriminating means for selecting normal values from the mean square values in the plurality of frequency bands by the Campbell output evaluating means and supplying the normal values to the reactor power monitoring means as information on the reactor power. 7. The apparatus of claim 1, 8. The apparatus of claim 1, 9. The apparatus of claim 1, 10. The apparatus of claim 1, 11. The apparatus of claim 1, 12. The apparatus of claim 11, wherein the obtaining means is means for obtaining a plurality of operation values including third or following operation values in addition to the first and the second operation values, based on the same formula as in the case of obtaining the first and second operation values at the sampling intervals while changing the constants. 13. The apparatus of claim 1, 14. The apparatus of claim 13, 15. The apparatus of claim 13,
	claims	1. An apparatus for conducting high resolution imaging of radioactive emissions from a desired source, the apparatus comprising:a collimator comprised of a plurality of stacked plates, each plate having a plurality of holes through each plate defining a plurality of passages through the collimator permitting the passage of radiation therethrough;a scintillation crystal permitting the detection of photons coming from the source and passing through the collimator to produce light received by a light pulse sensor;a data processor connected to the light pulse sensor for digitally computing the count of photons and visually constructing a digital image on a display device of the relative location of the emitted radiation;an optical integrating cavity for directing the photons emitted by the scintillation crystal to the light pulse sensor; andwherein the plurality of stacked plates comprises certain plates made from a different material than a remainder of the stacked plates. 2. The apparatus for conducting high resolution imaging of radioactive emissions from a desired source as set forth in claim 1 wherein the optical integrating cavity comprises a reflective, semi-hemispherical intermediate surface for directing the photons emitted from the scintillation crystal towards the light pulse sensor. 3. The apparatus for conducting high resolution imaging as set forth in claim 1 further comprising the about ten-thousand holes in each plate. 4. The apparatus for conducting high resolution imaging as set forth in claim 1 further comprising chemically etched holes in a size ranging from about 565 microns on a top plate to about 206 microns in a bottom plate of the collimator. 5. The apparatus for conducting high resolution imaging as set forth in claim 4 further comprising the plates being about 101 microns thick.
	abstract	The present disclosure may disclose a multi stage safety injection device, including a safety injection tank formed to contain coolant to be injected into a reactor vessel by a gravitational head of water when an accident occurs in which the pressure or water level of the reactor vessel is decreased, a pressure balance line connected to the reactor vessel and safety injection tank to form a pressure balance between the reactor vessel and the safety injection tank, and a set of safety injection lines connected to the safety injection tank and the reactor vessel to inject coolant to the reactor vessel in a pressure balance state between the reactor vessel and the safety injection tank, and connected to the safety injection tank with different heights to reduce a flow rate of coolant injected into the reactor vessel step by step according to the water level reduction of the safety injection tank in order to inject coolant to the reactor vessel at multi stages.
	claims	1. A lithography system in which an electronic image pattern is delivered to a charged particle exposure tool for projecting an image on a target surface, said exposure tool including a frame for elements constituting a charged particle projection column, and said exposure tool comprising a vacuum housing for housing said frame and particle column, and said lithography system comprising a control unit external to said vacuum housing for controlling exposure projections by means of control data, said control data being provided to a modulator for modulating charged particle beams generated in said exposure tool, said control data being provided via an optical interconnect conveying modulated light signals, said beam modulator thereto comprising an array of deflectors for modulating projection beams and light sensitive areas for receiving said control data, and said beam modulator at least partly extending in a charged particle projection space of said exposure tool, and being included in said element frame via a modulator holder, said control data finally being coupled in to said modulator using a free space optical interconnect for emitting modulated light beams to said light sensitive parts of said beam modulator, said optical interconnect further comprising an array of fibers coupled into said tool via a part of said vacuum housing, wherein an inner vacuum housing end part of said fibers is at its end mechanically secured to a light optics housing which in turn is secured to said modulator holder. 2. A system according to claim 1, wherein said feed through housing part is demountable. 3. A system according to claim 1, wherein said inner vacuum housing end part of said fibers is secured from outside mechanical impulses. 4. A system according to claim 1, wherein a feed through of an array of fibers in said vacuum housing realizes an air tight sealing by using a sealing material. 5. A system according to claim 1, wherein said light optics housing secures the positions of fiber ends of said array of fibers relative to light sensitive areas of the modulator. 6. A system according to claim 1, wherein the optics housing houses a lens part of the light optics interconnect. 7. A system according to claim 1, wherein the optics housing houses a mirror part of the light optics interconnect. 8. A system according to claim 7, wherein said mirror is holed. 9. A lithography system according to claim 1, in which an electronic image pattern is delivered to an exposure tool formed by a writing tool by means of light projection, utilizing an exposure projection, in particular a multi beam system for mask-less projection of a pattern on to an exposure surface, comprising a vacuum housing within which such writing tool is incorporated,a multi beam projection source being present, creating a plurality of writing beams for writing said pattern,which writing beams are directed to an blanker array comprising a control unit with individual controllers such as electrostatic blanker deflectors for individually controlling a writing beam in accordance received pattern information, in particular deflecting a writing beam or not, to a beam stopping part,a light optical system comprising light transmitting parts being present for transmitting pattern information signals to such controllers, in particular blanker deflectors,which controllers comprise light sensitive elements for receiving such modulated light beams, such light sensitive elements preferably being accommodated within near vicinity of the deflectors,said light optical system comprising a free space optical interconnect, forming a light optical data carrier system, transmitting pattern data carrying, modulated light beams towards said controllers,which free space optical interconnect comprises an emitter part emitting free space interconnect, pattern data carrying light beams to said light sensitive elements,whereinsaid free space optical interconnect comprises a holey mirror, incorporated in the projection trajectory of said plurality of writing beams,said mirror being arranged relative to said emitter part and said light sensitive elements to realize an on-axis incidence of said light beams on said light sensitive elements,and said mirror being provided with at least one hole allowing passage of one or more of said writing beams. 10. A system according to claim 1, in which an emitting part is incorporated to emit said light beams at least virtually perpendicular relative to a direction of said exposure projection. 11. A system according to claim 10, wherein said writing tool preferably utilizing a multi beam system for preferably mask-less projection of a pattern on to an exposure surface, comprising a vacuum housing within which such lithographic system is incorporated. 12. A system according to claim 1, in which said free space interconnect is included at a down stream side of the control unit. 13. A system according to claim 1, in which said free space interconnect, including an emitter and said holey mirror is included in a housing that is mechanically connected to the control unit, in particular a blanker array, more in particular via a holder therefore. 14. A system according to claim 1, in which a focusing lens is incorporated in said free space optical interconnect in near proximity to said holey mirror, in particular significantly nearer to said mirror than to said emitter, said lens in particular being common to all of the light beams emitted by said emitter. 15. A system according to claim 1, in which a micro lens is incorporated in said free space optical interconnect, in particular thereby forming said emitter, in close proximity to a light carrier end, in particular significantly closer to said carrier end than to said holey mirror, in particular each light carrier composing part such as an optic fiber end being provided with a micro lens. 16. A system according to claim 1, wherein the free space optical interconnect comprises a micro lens and a focusing lens common to a possible plurality of micro-lens completed fibers, the micro lens therein magnifying a fiber transmitted light signal, and the focusing lens de-magnifying the entirety of light signals transmitted by said possible plurality of fibers. 17. A system according to claim 1, wherein said free space optical interconnect is included in between a control unit formed by a blanker array for blanking writing beams, and a stopping plate for stopping writing beams deflected by said blanker array. 18. A system according to claim 1, wherein one or more light signal carriers are fed through a vacuum wall for the exposure tool using vacuum compatible sealing material, and are with end parts thereof subsequently mechanically coupled to a free space optical connect housing located in a vacuum space for the charged particle beam column. 19. A lithography system in which an electronic image pattern is delivered to a charged particle exposure tool for projecting an image on a target surface, said exposure tool including a frame for elements constituting a charged particle projection column, and said exposure tool comprising a vacuum housing for housing said frame and particle column, and said lithography system comprising a control unit external to said vacuum housing for controlling exposure projections by means of control data, said control data being provided to a modulator for modulating charged particle beams generated in said exposure tool, said control data being provided via an optical interconnect conveying modulated light signals, said beam modulator thereto comprising an array of deflectors for modulating projection beams and light sensitive areas for receiving said control data, and said beam modulator at least partly extending in a charged particle projection space of said exposure tool, and being included in said element frame via a modulator holder, said control data finally being coupled in to said modulator using an optical interconnect for transmitting light beams to said light sensitive parts of said beam modulator, said optical interconnect further comprising an array of fibers coupled into said tool via a part of said vacuum housing, wherein the array of fibers is fed through an opening in a part of said housing by using an amount of vacuum compatible sealing material for realizing an air tight sealing of the fibers in the opening. 20. Lithography system according to claim 19, wherein the array of fibers is fed through the opening in a demountable part of said housing.
	claims	1. A process for encapsulating a radioactive object to render the object suitable for shipment and/or storage, and including the steps of:(a) preparing a plastic material;(b) causing the plastic material to react with a foaming agent;(c) generating a foaming plastic;(d) encapsulating the radioactive object in the foaming plastic; and(e) allowing the foaming plastic to solidify around the radioactive object to form an impervious coating. 2. The Method according to claim 1, wherein the step of encapsulating the radioactive object includes the steps of filling a void in the object with the foaming plastic and encasing the object in an outer layer of foaming plastic. 3. The Method according to claim 1, wherein the step of encapsulating the radioactive object includes the step of placing the object in a bag before encasing the object in an outer layer of foaming plastic. 4. The Method according to claim 1, wherein the step of encapsulating the radioactive object includes the step of applying an outer layer of an elastomeric coating to the object. 5. The Method according to claim 1, wherein the foaming plastic comprises a rigid polyurethane foam with the composition comprising: Polyol blend, Crosslinkers, Water Fire retardant, Viscosity suppressant, Surfactants, Catalysts, Blowing agent, and/or Polymeric Isocyanate. 6. The Method according to claim 1, wherein the encapsulated object is coated with a polyurea elastomeric coating with the following composition: Polyetheramine blend, Amine Crosslinker, Moisture Scavenger, and/or Isocyanate Prepolymer. 7. A process for encapsulating a radioactive object to render the object suitable for shipment and/or storage, and including the steps of:(a) preparing a plastic material;(b) causing the plastic material to react with a foaming agent;(c) generating a foaming plastic;(d) placing a radioactive object in a container;(d) encapsulating the container in the foaming plastic; and(e) allowing the foaming plastic to solidify around the container to form an impervious coating. 8. The Method according to claim 7, and including the steps of evacuating displaced air from the container as the container is encapsulated and transferring the air to another treatment location. 9. The Method according to claim 7, wherein the foaming plastic comprises a rigid polyurethane foam with the composition comprising: Polyol blend, Crosslinkers, Water Fire retardant, Viscosity suppressant, Surfactants, Catalysts, Blowing agent, and/or Polymeric Isocyanate. 10. The Method according to claim 7, wherein the encapsulated object is coated with a polyurea elastomeric coating with the following composition: Polyetheramine blend, Amine Crosslinker, Moisture Scavenger, and/or Isocyanate Prepolymer. 11. A method of encapsulating a radioactive object to render the object suitable for shipment and/or storage, and including the steps of:(a) preparing a plastic material;(b) causing the plastic material to react with a foaming agent;(c) generating a foaming plastic; and(d) encapsulating the object in the foaming plastic, wherein the step of encapsulating the object in the foaming plastic includes the steps selected from the group consisting of:(i) placing a radioactive object in a container, encapsulating the container in the foaming plastic, and allowing the foaming plastic to solidify around the container to form an impervious coating; and(ii) encapsulating the radioactive object in the foaming plastic, allowing the foaming plastic to solidify around the radioactive object to form an impervious coating. 12. The Method according to claim 11, wherein the step of encapsulating the radioactive object includes the steps of filling a void in the object with the foaming plastic and encasing the object in an outer layer of foaming plastic. 13. The Method according to claim 11, wherein the foaming plastic comprises a rigid polyurethane foam with the composition comprising: Polyol blend, Crosslinkers, Water Fire retardant, Viscosity suppressant, Surfactants, Catalysts, Blowing agent, and/or Polymeric Isocyanate. 14. The Method according to claim 11, wherein the encapsulated object is coated with a polyurea elastomeric coating with the following composition: Polyetheramine blend, Amine Crosslinker, Moisture Scavenger, and/or Isocyanate Prepolymer.
047818824	description	DETAILED DESCRIPTION The present invention provides an improved refueling machine for a nuclear reactor and an improved guide means for aligning the inner and outer mast of a refueling machine. An apparatus 101 for loading or unloading a fuel assembly in a nuclear reactor and improved guide means 103, as illustrated in FIGS. 5 to 8, has a means 105 for positioning a fuel assembly 107 above a nuclear reactor (not shown). The apparatus includes a housing 109 which carries concentric inner mast 111 and outer mast 113, the inner mast having at the lower portion thereof a means 115 for gripping the upper nozzle 117 of a fuel assembly 107, and means for releasing or engaging the fuel assembly. The means 105 for positioning a fuel assembly comprises a movable bridge 119 that is mounted on opposite walls 121 of a reactor containment, the bridge 119 having structural members 123 for strength and rigidity. The bridge 119 is adapted to travel horizontally on wheels 125 and rails 127 fixed on the top of containment walls 121. Wheels 125 are driven by a motor 129 and shafts 131. A motor driven trolley 133 is mounted on wheels 135 which run on rails 137 mounted on the bridge 119 for horizontal travel transverse to the travel of the bridge 119, such that the housing 109 and inner and outer masts 111, 113 are positionable over any one of the fuel assemblies in the reactor or a refueling pool to either insert or remove a fuel assembly 107 from the reactor core. A platform 139 is provided on the trolley 133 as well as a control console 141 which contains equipment to control drive motors, hoists and air compressors. The housing 109 for concentric masts 111, 113 is attached to the platform 139 such as by welding, with one end of the housing 109 extending below the platform 139 while the other end projects thereabove. A window 143 may be provided to observe the position of the upper end of the masts, and an electrical control 145 may be welded to the mast housing 109. A stationary plate 147 is mounted on the upper end of housing 109 and has an opening through which hoist, electrical and air lines extend. This opening also serves to permit removal or insertion of the inner mast 111 for maintenance. Cable reels, such as reel 151 enclose reels which support air lines or electrical cables arranged to be played out or taken in as the mast is raised or lowered. Plate 147 also supports hoist 153, which hoist carries hoist cable 155 that supports the inner mast 111 which, in turn, carries the fuel assembly 107. In operation of the fuel assembly loading or unloading apparatus 101, to remove a fuel assembly 107 from a reactor core, means 105 for positioning the fuel assembly 107 is operated and the bridge 119 and trolley 133 positioned directly above the fuel assembly that is to be removed. The inner mast 111, which is in an upper stored position within outer mast 113, is then lowered and the gripping means 115 engages the upper nozzle 117 of the fuel assembly. The inner mast 111 is reciprocally mounted in the outer mast 113 and movable as indicated by the double arrow in FIG. 6. Upon gripping of the fuel assembly 107, the hoist is engaged to lift the fuel assembly from the reactor core. The hoist 153 raises the inner mast 111 and engaged fuel assembly 107 into the confines of the outer mast 113 to a housed portion. The positioning means, trolley and bridge are then moved to another position and the fuel assembly lowered to a storage chamber and released from the inner mast. Loading of a fuel assembly in a reactor core involves reverse operation of this procedure. During the coaxial movement of the inner mast 111 relative to the outer mast 113, guide means are provided to maintain a concentric relationship between these two masts. The guide means 103 of the present invention is illustrated in detail in FIGS. 7 to 10 in association with a rectangular-shaped inner mast and circular-shaped outer mast, although the cross-section of the masts could be of another shape. The guide means 103 has a plurality of inner members 161 on the inner mast 111 and a plurality of complementary shaped cooperating outer members 163 on the outer mast 113, with means 165 for supporting the outer members 163 on the outer mast 113 and means 167 for pivoting the outer members 163 relative to the support means 165 towards and away from the inner members 161. The inner members 161 are illustrated as a corner 169 of the rectangularly-shaped inner mast 111, and a pair of vertically extending rods 171, having a convex surface or rounded face 173, secured to the wall 175 of the inner mast by means of bolts 177. The convex surface or corner 169, acting as an inner member also has a rounded outer face 179. Preferably three such inner members 161 are provided equally spaced about the inner mast on an angle of about 120 degrees apart to provide a balanced support of the inner mast 111 in the outer mast 113. The complementary outer members 163 on the outer mast 113 which cooperate with the inner members 161 on the inner mast 111 are illustrated as rollers 181 that are pivotally attached to the means 165 for supporting the same, the rollers 181 having a groove 183 therein in which the rounded faces 173 and 179 will ride. The support means 165 comprises a framework 185 comprised of top wall 187, above an opening 189 in the wall 191 of the outer mast 113 and a bottom wall 193, below the opening 189, with a rear wall, or plate, 195 extending between top and bottom walls 187 and 193. The plate 195 has two spaced apertures, upper and lower apertures, 197, 199, therethrough, through which upper and lower bolts 201, 203 are inserted, the bolts 201, 203 threadedly engaging in threaded bores 205 in the wall 191 of the outer mast 113. The bottom wall 193 is in the form of a block 207 with an aperture 209 therethrough coaxial with lower aperture 199, through which the lower bolt 203 also extends, to provide additional rigidity, while the top wall 187 is of a lesser thickness than the block 207 and may have a concave groove 211 in the lower face 213 thereof in which upper bolt 201 will seat. A bore 215 is formed through the block 207 transverse to aperture 209 and a pivot pin 217 is positioned in the bore 215. The pivot pin has opposite ends 219 and 221 to which upwardly extending side bars 223 and 225 respectively are secured. The side bars 223 and 225 each carry an inwardly directed arm 227 between which a shaft 229 extends, with the roller 181 mounted on the shaft 229 between the arms 227. Adjacent the upper end 231 of the side bars and extending between the upper ends of the side bars 223 and 225 is an alignment means 232 such as an alignment beam 233 which is fixed at opposite ends 235 and 237 to the side bars 223 and 225. The side bars 223 and 225, being secured to the pivot pin 217 are movable towards and away from the inner mast 111, such that the roller 181 on the shaft 229 extending between arms 227 of the side bars are also movable towards and away from the inner mast 111. Precise alignment of the position of the rollers 181 is effected by use of an adjustable device 241 which contacts the alignment beam 233. Adjustable device 241, as illustrated, is in the form of a threaded shaft 243 which is movable towards or away from alignment beam 233 and can be locked in various positions relative to the same, the threaded shaft secured, such as by welding, to a sleeve 245 that encloses the alignment beam 233, but is adapted to permit rotational movement of the alignment beam relative to the sleeve. The threaded shaft 243 extends from the sleeve 245 outwardly through an opening 247 in the plate 195 and has threads 249 thereon. A hollow stud 251 having a shoulder 253, engageably with the inner wall of plate 195, has threads 255 on the internal wall 257 thereof and threads 259 on the external wall 261 thereof. The threads 255 on the internal wall 257 are engageable with the threads 249 on the threaded shaft 243. A lock nut 263 having threads 265, is threadedly engageable with the threads 259 on the external wall 261 of the hollow stud on the opposite side of the plate 195 from threaded shoulder 253. In operation of the adjustable device to move the roller 181 on shaft 229 inwardly towards the inner mast 111, the lock nut 263 is loosened and the threaded shaft 243 and hollow stud are moved inwardly which moves sleeve 245 and alignment beam 233. Side bars 223 and 225 secured to the alignment beam 233 also move inwardly, pivoting on pivot pin 217. When the desired placement is achieved, the hollow stud 251 is threaded along threaded shaft 243 until shoulder 253 abuts the inside surface 267 of plate 195. The threaded lock nut 263 is then threaded along the hollow stud 251 until the nut abuts the outside surface 269 of the plate 195 to lock the shaft 243, alignment beam 233, and the roller 181 in position. An alignment pin or knob 271 extends from an end 273 of alignment beam 233, which provides an external reference surface. The surface or pin facilitates field alignment of all elevations of guide rollers by using a plumb line or an optical sighting device to align the pins vertically. By so aligning the pins 271, the vertically spaced plurality of rollers 181 are aligned along a common vertical axis for easy field alignment of all of the rollers. The present guide means can be provided on new refueling machines or can be used as an alteration to existing refueling machines as a retrofit system.
	summary
039492311	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows an exemplary embodiment of an infrared radiating unit according to the present invention. This unit is to be used in an infrared analyzer of which the other components have not, however, been illustrated because they do not themselves form part of the invention. Such units in combination with infrared analyzers are of course well known. The novel unit illustrated in FIG. 1 has a mount 1 another portion of which is formed with a central bore and is split along this bore; a carrier 2 of ceramic material has a larger-diameter portion which is located in this bore, and a nut 3 is threaded onto the exterior of the split upper portion of the mount 1 and pushes this portion, or rather the sections of this portion which result from the split, against the larger-diameter part of the carrier 2, thus clamping the same in such a manner that when the nut 3 is loosened, it can be shifted axially or turned. However, when the nut 3 is tightened, then the carrier 2 will be centrally and immovably clamped and held in place. The carrier 2 has a portion 4 of reduced cross-sectional area which extends out of the mount 1 and extends through an opening in a curved portion 14 of a reflector, from an open side of which curved portion there extends a cylindrical portion 13 of the reflector. The portion 4 is formed or provided with two axially spaced flanges 5 and 6 and carries intermediate these flanges a metal-jacketed resistance wire 7 formed as a coil which in the illustrated embodiment is configurated as a double-layer coil. The use of a double-layer coil produces a more homogeneous distribution of the infrared radiation. Within the metal jacket the resistance wire 7 is imbedded in a fire clay or the like, for example in magnesium oxide. Terminals 8 and 9 extend through bores in the carrier 2 to the reduced diameter portion 4 where they are hard-soldered to the wire 7a of the coil 7 and at the locations 10 and 11 where the wire 7a enters into the metal jacket of the coil 7 the open ends of the metal jacket are closed by pouring a ceramic substance into them which hardens and forms plugs. The mount 1 is formed with a recess 12 in which the reflector 13 is secured, for example by a shrink fit. The reflector 13 has a central bore through which the portion 4 extends, and it is composed of the two parts 14 and 15 already mentioned. The inner reflector surface 14a of the part 14 is configurated as a rotationally symmetrical paraboloid which is axially bisected and pushed apart in transverse direction, so that the focal points 16 for the individual paraboloid sectors are located on a circle surrounding the axis of the arrangement, and which circle coincides with the circumference of the coil 7. This arrangement assures that the infrared radiation emitted from the middle of the coil 7 will, after reflection on the surface 14, be converted in axially parallel ways which increases the effectiveness of the unit since a larger proportion of the radiation than would otherwise be possible will now be passing through the measuring receptacle (not shown) to the radiation detector (not shown) of the infrared analyzer. The reflector 13 is also of a material having relatively poor thermal conductivity, for example a rust-resistant steel having a high nickel content, or the like. The inner surface 14a of the portion 14, and if desired also the inner surface of the portion 15, can be polished and gold plated to obtain a particularly good reflection. The unit of FIG. 1 has only small thermal losses due to its construction. The carrier 2 is of ceramic material having poor thermal conductivity, and in addition its portion 4 has a smaller cross-sectional area than the remainder of the carrier 2, so that the flow of thermal energy through the carrier 2 is further reduced thereby. The reflector 13 is heated by the heat of the coil 7, and due to the use of a material having poor thermal conductivity, for instance a rust-resistant steel having a small coefficient of thermal conductivity, the loss of heat by conduction to the mount 1 is maintained low. This effect is reinforced in that the contact area 12 between the mount 1 and the reflector 13 is relatively small. The clamping of the carrier 2 between the arms formed in the mount 1 by the slot provided in the latter, and the clamping action exerted by the nut 3, mounts the carrier 2 centrically and in a very stable manner, so that it will not change its position relative to the optical axis of the arrangement even if the arrangement is subjected to exceptionately strong vibrations or the like. Moreover, this arrangement makes it possible for the carrier 2 and the coil 7 to be readily removed and replaced or inspected, and to be shifted axially or turned in order to select a position of the coil 7 in which a maximum yield and focussing of the radiation is obtained. The embodiment of FIG. 2 is in almost all respects identical with that of FIG. 1, so that the identical portions have not been illustrated. It differs from FIG. 1 in the manner in which the coil 7 is mounted on the carrier 2. For manufacturing reasons it is advisable not to mount the coil 7 directly upon the carrier 2, or rather the portion 4 thereof, since the carrier 2 would have to undergo mechanical stresses during the application of the coil and might break somewhere in its reduced cross-section portion 4. This problem is overcome in FIG. 2 by providing a metallic sleeve 17, again preferably of rust-resistant steel or the like, which is provided with axially spaced flanges or the like, and on which the coil 7 is mounted intermediate these flanges. The portion 4 then of course does not have similar flanges, and the dimension of the sleeve 17 is such that it can be pushed onto the portion 4 where it is secured in place, for example by means of an adhesive. This eliminates any mechanical stresses upon the carrier 2 and maintains all of the advantages outlined with respect to the description of FIG. 1. The ceramic material for the carrier 2 consists of 99.9% pure Al.sub.2 O.sub.3. Instead of a rust-resistant steel as material for the mount 1 and the reflector 13 other materials having a relatively poor thermal conductivity may be used, for example Al.sub.2 O.sub.3 ceramique or glass. A suitable adhesive for securing the sleeve 17 in place on the portion 4 of the carrier is ceramic adhesive sold under the tradename "Thermoguss 2000". An infrared analyzer in which the invention may be used is of the type as described in Luft U.S. Pat. No. 3,162,761. It will be understood that each of the elements described above, or two or more together, may also find a useful application in other types of constructions differing from the types described above. While the invention has been illustrated and described as embodied in an infrared radiator of unit for infrared analyzers, it is not intended to be limited to the details shown, since various modifications and structural changes may be made without departing in any way from the spirit of the present invention. Without further analysis, the foregoing will so fully reveal the gist of the present invention that others can by applying current knowledge readily adapt it for various applications without omitting features that, from the standpoint of prior art, fairly constitute essential characteristics of the generic or specific aspects of this invention.
	summary
	description	1. Field of the Invention The present invention relates to a specimen observation method, an image processing device, and a charged-particle beam device. More particularly, it relates to a method and a device which are preferable for observing the same field-of-view as that of a specimen observed using an optical microscope. 2. Description of the Related Art Concerning a specimen to be observed using an optical microscope, in the case of, e.g., a living-creature specimen, the specimen can be colored by an appropriate staining method which depends on differences in the specimen components and its physiological state. As a result, the specimen can be observed in a state which is close to that of a living body. By the way, in the analysis of living-body reactions like this, fine structure analysis at cell level or macromolecular level is important. An optical microscope, however, finds it difficult to observe the fine structure. This is because the optical microscope has a limitation to its resolving power, and thus can exhibit only an insufficient magnification. Accordingly, for this fine structure analysis to be implemented, the observation made by an electron microscope is desirable. In JP-A-6-13011, the following explanation has been given: A specimen is observed using an optical image capture device provided independently of an electron microscope. Moreover, its coordinate position and the position of specimen stage of the electron microscope are made to correspond to each other, thereby searching for field-of-view in the electron microscope. In JP-A-8-129986 (corresponding to U.S. Pat. No. 5,646,403), the following explanation has been given: Regarding field-of-view displacement of a scanning electron microscope, the specimen stage is displaced in correspondence with observation field-of-view width of the scanning electron microscope. An optical microscope performs formation of a specimen image mainly by detecting reflection light from the specimen. In contrast thereto, however, an electron microscope performs formation of a specimen image by detecting secondary electrons and the like. In this way, since the specimen images are formed based on the different signals, the image qualities are exceedingly different from each other. Consequently, it has been difficult to retrieve the observation field-of-view of the electron microscope on the basis of the specimen image acquired using the optical microscope. It is an object of the present invention to provide a specimen observation method, an image processing device, and a charged-particle beam device which are preferable for selecting, based on an image acquired by an optical microscope, an image area that should be acquired in a charged-particle beam device the representative of which is an electron microscope. In the present invention, in order to accomplish the above-described object, there are provided a method and a device for determining the position for detection of charged particles by making the comparison between a stained optical microscope image and an elemental mapping image formed based on X-rays detected by irradiation with the charged-particle beam. The configuration like this makes it possible to easily implement that, e.g., a reaction region within an optical microscope image to which a staining corresponding to a living-body reaction has been applied is set and employed as the observation target of a charged-particle beam device such as an electron microscope. According to the present invention, it becomes possible to easily implement a search for field-of-view for the observation by a charged-particle beam device with respect to a specimen observed using an optical microscope. Other objects, features and advantages of the invention will become apparent from the following description of the embodiments of the invention taken in conjunction with the accompanying drawings. Hereinafter, referring to the drawings, the detailed explanation will be given below concerning an example of a charged-particle beam device in order to explain an embodiment of the present invention. As a charged-particle beam device, there exists a scanning electron microscope, transmission electron microscope, scanning transmission electron microscope, or ion beam irradiation device. In the present example, the explanation will be given below regarding an embodiment of the present invention, referring to FIG. 1 and selecting a scanning electron microscope as the example. As illustrated in the drawing, an electron beam 3 emitted from an electron gun 2 of main body of an electron microscope 1 is converged by an irradiation lens 4. Next, the electron beam 3 is deflected by a scanning coil 6 including an X-direction deflection coil and a Y-direction deflection coil. Moreover, the electron beam 3 deflected into the two directions is focused on a specimen 7 held by a specimen stage 8 by an objective lens 5, then being scanned on the specimen 7. The electron gun 2 is controlled by an electron gun control device 12. The irradiation lens 4 and the objective lens 5 are controlled by an irradiation lens control device 13 and an objective lens control device 14, respectively. Also, a not-illustrated specimen stage driving mechanism is controlled by a specimen stage control device 16. The scanning coil 6 is controlled by a scanning coil control device 15. These devices, i.e., the electron gun control device 12, the irradiation lens control device 13, the objective lens control device 14, the scanning coil control device 15, and the specimen stage control device 16, configure an observation condition setting device. Characteristic X-rays are generated from surface of the specimen 7 by the electron beam 3 scanned on the specimen 7. Next, the characteristic X-rays generated are detected by an X-ray detector 11, and are supplied to an image processing device 18, then being recorded and stored as image data. Simultaneously, secondary electrons generated from the specimen 7 are detected by a secondary-electron detector 9, and are supplied to the image processing device 18 as image data, then being recorded and stored. Simultaneously, reflection electrons generated from the specimen 7 are detected by a reflection-electron detector 10, and are supplied to the image processing device 18 as image data, then being recorded and stored. An optical image, which is acquired by an optical device 17 independently of the electron microscope, is supplied to the image processing device 18, then being recorded and stored as image data. In this embodiment, the observation condition setting device of the electron microscope, i.e., each of the electron gun control device 12, the irradiation lens control device 13, the objective lens control device 14, the scanning coil control device 15, and the specimen stage control device 16, is so configured as to be connected to the image processing device 18 via a predetermined transmission path so that the mutual data transmission/reception is executable. Also, the image processing device 18 is so configured as to allow implementation of the driving control over the specimen stage and setting of the observation conditions for each lens. On account of this, the image processing device 18 is so configured as to include, e.g., a computer onto which a predetermined program is installed. This program makes it possible to create control data which is necessary for the above-described specimen-stage driving mechanism control device and observation condition setting device, and which should be supplied thereto. Hereinafter, referring to a flowchart in FIG. 2, the explanation will be given below concerning an example of the operation when the same field-of-view as that of a specimen image captured using an optical device is observed using a charged-particle beam device. In the present embodiment, “image judgment processing”, “image record processing”, “image retrieval processing”, and “condition registration processing” are carried out by the image processing device 18. First, using the optical microscope, the operator observes the specimen 7 which is stained for the optical microscope. Although not illustrated in FIG. 1, the specimen image observed is photographed by a record device such as CCD camera. After that, the photographed image is supplied to the image processing device 18. Next, the specimen 7 for the optical microscope is set on the specimen stage 8 inside a specimen chamber of the scanning electron microscope. Moreover, the electron beam 3 is scanned on the surface of the specimen 7 in accordance with a predetermined procedure, and resultant characteristic X-rays generated are detected by the X-ray detector 11 (step: 2-1). In addition, elemental mapping images where intensity of the characteristic X-rays detected by the X-ray detector 11 is displayed in a two-dimensional manner is supplied to the image processing device 18, then being recorded and stored as image data (step: 2-2). Subsequently, in the image processing device 18, the comparison is made between the image data on the specimen image by the optical microscope and the elemental mapping images. At this time, it is advisable to use an image correlation method. Result of the image correlation, which is the degree of coincidence between both of the images, may be displayed as “coincidence degree”. If an elemental mapping image exhibiting the highest coincidence degree has been successfully selected out of the result by the image correlation (step: 2-3), a secondary-electron image and a reflection-electron image within the field-of-view are observed (step: 2-4). Incidentally, in the present embodiment, a predetermined threshold value may be set in advance regarding the coincidence degree of the images, and an image which is found to exceed this threshold value may be selectively acquired, observed, and recorded. Simultaneously with the acquisition of the elemental mapping image of the specimen image acquired by the image optical device, the secondary-electron image and the reflection-electron image within the field-of-view may be observed, and be recorded into the image processing device 18 as the image data. These series of image data, i.e., the specimen image acquired by the optical device, its elemental mapping image, and its secondary-electron image and reflection-electron image, may be recorded into the image processing device 18 as a single data group. At this time, the observation conditions of the optical microscope have been supplied to the observation condition setting device of the electron microscope via the image processing device 18. As a result, the observation conditions in the electron microscope coincide with the observation conditions of the optical microscope. In the image data groups, in the case of the image data stored in, e.g., TIF image format (step: 2-5), all the observation conditions of the observation condition setting device of the electron microscope may be written into tag areas, thereby being recorded in a manner of being made to correspond to the images (step: 2-6, step: 2-7). Moreover, the specimen stage is displaced to the next observation position (step: 2-8), then returning to observation of the specimen depending on the requirements. Hereinafter, basically the same steps will be executed. Next, in this embodiment, the same field-of-view is automatically searched for by performing displacement of the specimen stage in such a manner that the observation area by the optical device is selected and defined as the unit of this displacement. Referring to a flowchart in FIG. 3, the explanation will be given below concerning the operation at this time. With respect to the specimen image by the optical microscope, the observation field-of-view area of the optical microscope is determined from the image data supplied to the image processing device 18 (step: 3-1). The specimen stage 8 is controlled by the specimen stage control device 16 of the electron microscope in such a manner that the specimen stage 8 displaces with the observation field-of-view area as the unit of the displacement amount of the specimen stage (step: 3-2). Acquisition positions of the elemental mapping of the optical microscope specimen by the X-ray detector 11 are read by the specimen stage control device 16, then being recorded in a manner of being made to correspond to the elemental mapping images (step: 3-3). The results of the image correlation between the elemental mapping of the optical microscope specimen by the X-ray detector 11 and the optical microscope specimen images are displayed as, e.g., the coincidence degrees (step: 3-4). Depending on rank of the coincidence degrees by the image correlation between the elemental mapping of the optical microscope specimen by the X-ray detector 11 and the optical microscope specimen images, the elemental mapping images, secondary-electron images, and reflection-electron images of the specimen are ranked. The optical microscope specimen images, the elemental mapping images of the optical microscope specimen by the X-ray detector, the secondary-electron images, and the reflection-electron images are recorded and stored into the image processing device 18 as the data group (step: 3-5). Also, in this embodiment, the specimen stage control device 16 may be controlled so that the specimen stage 8 displaces in such a manner that the acquisition area of the elemental mapping of the optical microscope specimen by the X-ray detector 11 is selected and defined as the displacement amount of the specimen stage. As the result of the image correlation between the elemental mapping images of the optical microscope specimen by the X-ray detector 11 and the optical microscope specimen images, if the highest coincidence degree is found, or if a coincidence degree is found to exceed a predetermined threshold value, the displacement of the specimen stage 8 is halted. Then, its position coordinate at this time is recorded and stored (step: 3-6). Consequently, according to this embodiment, it becomes possible to directly observe the specimen for the optical microscope in the scanning electron microscope without preparing a specific specimen manufacturing intended for the scanning electron microscope. On account of the image correlation between the elemental mapping images of the specimen for the optical microscope and the specimen images acquired by the optical device, the same field-of-view as that of the observation region of the specimen for the optical microscope can be observed by the scanning electron microscope. The reaction region of the specimen for the optical microscope to which a staining corresponding to a living-body reaction has been applied can be directly observed by the electron microscope. This makes it possible to directly observe the relationship between the living-body reaction and the fine structure. Moreover, the specimen stage is displaced in such a manner that the acquisition area of the elemental mapping images of the optical microscope specimen is selected and defined as the unit of the displacement amount. Then, these pieces of coordinate information are stored in a manner of being made to correspond to the elemental mapping images. Furthermore, the elemental mapping images of the optical microscope specimen, the secondary-electron images, and the reflection-electron images are stored as the data group, then being ranked depending on the coincidence degrees. Then, the displacement of the specimen stage can be automatically halted at the position at which the highest coincidence degree is found. It should be further understood by those skilled in the art that although the foregoing description has been made on embodiments of the invention, the invention is not limited thereto and various changes and modifications may be made without departing from the spirit of the invention and the scope of the appended claims.

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